Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors, 86918-87128 [2024-23434]
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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
NUCLEAR REGULATORY
COMMISSION
10 CFR Parts 1, 2, 10, 11, 19, 20, 21,
25, 26, 30, 40, 50, 51, 53, 70, 72, 73, 74,
75, 95, 140, 150, 170, and 171
[NRC–2019–0062]
RIN 3150–AK31
Risk-Informed, Technology-Inclusive
Regulatory Framework for Advanced
Reactors
Nuclear Regulatory
Commission.
ACTION: Proposed rule.
AGENCY:
The U.S. Nuclear Regulatory
Commission (NRC) is proposing to
revise the NRC’s regulations by adding
a risk-informed, performance-based, and
technology-inclusive regulatory
framework for commercial nuclear
plants in response to the Nuclear Energy
Innovation and Modernization Act
(NEIMA). The NRC plans to hold a
public meeting to promote full
understanding of the proposed rule and
facilitate public comments.
DATES: Submit comments by December
30, 2024. Comments received after this
date will be considered if it is practical
to do so, but the NRC is able to ensure
consideration only for comments
received before this date.
ADDRESSES: You may submit comments
by any of the following methods
however, the NRC encourages electronic
comment submission through the
Federal rulemaking website:
• Federal Rulemaking website: Go to
https://www.regulations.gov and search
for Docket ID NRC–2019–0062. Address
questions about NRC dockets to Helen
Chang; telephone: 301–415–3228; email:
Helen.Chang@nrc.gov. For technical
questions contact the individuals listed
in the FOR FURTHER INFORMATION
CONTACT section of this document.
• Email comments to:
Rulemaking.Comments@nrc.gov. If you
do not receive an automatic email reply
confirming receipt, then contact us at
301–415–1677.
• Fax comments to: Secretary, U.S.
Nuclear Regulatory Commission at 301–
415–1101.
• Mail comments to: Secretary, U.S.
Nuclear Regulatory Commission,
Washington, DC 20555–0001, ATTN:
Rulemakings and Adjudications Staff.
• Hand deliver comments to: 11555
Rockville Pike, Rockville, Maryland
20852, between 7:30 a.m. and 4:15 p.m.
eastern time, Federal workdays;
telephone: 301–415–1677.
You can read a plain language
description of this proposed rule at
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SUMMARY:
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https://www.regulations.gov/docket/
NRC-2019-0062. For additional
direction on obtaining information and
submitting comments, see ‘‘Obtaining
Information and Submitting Comments’’
in the SUPPLEMENTARY INFORMATION
section of this document.
FOR FURTHER INFORMATION CONTACT:
Robert Beall, Office of Nuclear Material
Safety and Safeguards, telephone: 301–
415–3874; email: Robert.Beall@nrc.gov;
or Anders Gilbertson, Office of Nuclear
Reactor Regulation, telephone: 301–
415–1541; email: Anders.Gilbertson@
nrc.gov. Both are staff of the U.S. NRC,
Washington, DC 20555–0001.
SUPPLEMENTARY INFORMATION:
Executive Summary
A. Need for the Regulatory Action
On January 14, 2019, the President
signed the Nuclear Energy Innovation
and Modernization Act (NEIMA) into
law (Pub. L. 115–439). NEIMA section
103(a)(4) directs the NRC to ‘‘complete
a rulemaking to establish a technologyinclusive, regulatory framework for
optional use by commercial advanced
nuclear reactor applicants for new
reactor license applications.’’ NEIMA
defines a ‘‘technology-inclusive
regulatory framework’’ as one that is
‘‘developed using methods of evaluation
that are flexible and practicable for
application to a variety of reactor
technologies, including, where
appropriate, the use of risk-informed
and performance-based techniques.’’
NEIMA, as further amended by the
Accelerating Deployment of Versatile,
Advanced Nuclear for Clean Energy Act
of 2024 (ADVANCE Act), defines the
term ‘‘advanced nuclear reactor’’ as ‘‘a
nuclear fission reactor or fusion
machine, including a prototype plant (as
defined in sections 50.2 and 52.1 of title
10, Code of Federal Regulations (as in
effect on the date of enactment of
[NEIMA])), with significant
improvements compared to commercial
nuclear reactors under construction as
of the date of enactment of [NEIMA].’’
The NRC initially considered
establishing the scope of proposed part
53, ‘‘Risk-Informed, TechnologyInclusive Regulatory Framework for
Commercial Nuclear Plants,’’ of title 10
of the Code of Federal Regulations (10
CFR) as being for ‘‘advanced nuclear
plants’’ consisting of one or more
‘‘advanced nuclear reactors’’ as defined
in NEIMA. Based on public discussions
on the use of the term, the NRC
determined that the NEIMA definition,
although broad, did not define
‘‘significant improvements’’ with
enough specificity to implement in NRC
regulations. Additionally, a number of
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stakeholders suggested that the
descriptor, ‘‘advanced,’’ implied
enhanced safety, while the NEIMA
definition includes ‘‘significant
improvements’’ in areas other than
safety enhancements. In response to this
feedback, and to be technology
inclusive, the NRC determined that the
broader term ‘‘commercial nuclear
plant’’ would be preferable.
The current application and licensing
requirements in 10 CFR part 50,
‘‘Domestic Licensing of Production and
Utilization Facilities,’’ and 10 CFR part
52, ‘‘Licenses, Certifications, and
Approvals for Nuclear Power Plants,’’
were primarily developed to address
license requests concerning watercooled reactors, and to address
operational requirements for those types
of reactors. This proposed rule responds
to NEIMA by creating an alternative
regulatory framework for licensing
future commercial nuclear plants. The
new alternative requirements and
implementing guidance would adopt
technology-inclusive approaches and
use risk-informed and performancebased techniques to ensure an
equivalent level of safety to that of
operating commercial nuclear plants
while providing flexibility for licensing
and regulating a variety of technologies
and designs for commercial nuclear
reactors.
B. Major Provisions
Major provisions of this proposed
rule, supported by accompanying
guidance, include the following:
• A new alternative technologyinclusive, risk-informed, performancebased framework that includes
requirements for licensing and
regulating nuclear plants during the
various stages of their life cycles.
• A new alternative technologyinclusive, risk-informed, and
performance-based framework in 10
CFR part 26, ‘‘Fitness for Duty
Programs,’’ developed from existing
requirements in subpart K, ‘‘FFD
Programs for Construction,’’ of part 26.
• A new alternative technologyinclusive and performance-based
security framework in 10 CFR part 73,
‘‘Physical Protection of Plants and
Materials,’’ that includes requirements
for protection of licensed activities at
commercial nuclear plants.
C. Costs and Benefits
The NRC prepared a draft regulatory
analysis to determine the expected
quantitative costs and benefits of this
proposed rule and associated guidance
as well as qualitative factors to be
considered in the NRC’s rulemaking
decision. The conclusion from the
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analysis is that this proposed rule and
associated guidance would result in net
averted costs to the industry and the
NRC ranging from $53.6 million using a
7-percent discount rate to $68.2 million
using a 3-percent discount rate, using an
assumption of one applicant under 10
CFR part 53. As the number of
applicants increases, so do the
estimated averted costs.
The draft regulatory analysis also
considers qualitative factors, such as
greater regulatory stability,
predictability, and clarity to the
licensing process. These benefits would
result from incorporating advances in
probabilistic risk assessment (PRA) and
other risk-informed analyses and
codifying regulatory enhancements that
currently exist in regulatory guides
(RGs). Another qualitative factor is
promoting a performance-based
regulatory framework that specifies
requirements to be met and provides
flexibility to an applicant or licensee
regarding the information or approach
needed to satisfy those requirements.
For more information, please see the
draft regulatory analysis (available in
the NRC’s Agencywide Documents
Access and Management System
(ADAMS) Accession No.
ML21165A112).
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Table of Contents
I. Obtaining Information and Submitting
Comments
A. Obtaining Information
B. Submitting Comments
II. Background
A. NRC Advanced Reactor Readiness
B. Stakeholder Views on Part 53
Preliminary Proposed Rule Language
III. Discussion
A. Objective and Applicability
B. Need for Changes to the Existing
Regulatory Framework
C. 10 CFR Part 53: Framework
IV. Part 53: Framework
Subpart A—General Provisions
A. Discussion of Definitions in Proposed
Part 53
B. Other General Provisions
Subpart B—Technology-Inclusive Safety
Requirements
Subpart C—Design and Analysis
Requirements
Subpart D—Siting Requirements
Subpart E—Construction and
Manufacturing Requirements
Subpart F—Requirements for Operation
Subpart G—Decommissioning
Requirements
Subpart H—Licenses, Certifications, and
Approvals
Subpart I—Maintaining and Revising
Licensing Basis Information
Subpart J—Reporting and Other
Administrative Requirements
Subpart M—Enforcement
V. Changes to Other Parts of 10 CFR Chapter
I
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10 CFR Part 26
A. Introduction
B. Proposed Changes to Part 26, Subparts
A Through E and I
C. Proposed Requirements for Part 26,
Subpart M
D. Proposed Changes to Part 26, Subpart N
E. Proposed Changes to Part 26, Subpart O
10 CFR Part 50
A. Section 50.160: Emergency
Preparedness for Small Modular
Reactors, Non-Light-Water Reactors, and
Non-Power Production or Utilization
Facilities
B. Appendix B to Part 50: Quality
Assurance Criteria for Nuclear Power
Plants and Fuel Reprocessing Plants
10 CFR Part 73
A. Section 73.100: Technology-Inclusive
Requirements for Physical Protection of
Licensed Activities at Commercial
Nuclear Plants Against Radiological
Sabotage
B. Section 73.110: Technology-Inclusive
Requirements for Protection of Digital
Computer and Communication Systems
and Networks
C. Section 73.120: Access Authorization
Program for Commercial Nuclear Plants
VI. Specific Requests for Comments
VII. Section-by-Section Analysis
VIII. Regulatory Flexibility Certification
IX. Regulatory Analysis
X. Backfitting and Issue Finality
XI. Cumulative Effects of Regulation
XII. Plain Writing
XIII. Environmental Assessment and
Proposed Finding of No Significant
Environmental Impact
XIV. Paperwork Reduction Act
XV. Criminal Penalties
XVI. Voluntary Consensus Standards
XVII. Availability of Guidance
XVIII. Public Meeting
XIX. Availability of Documents
I. Obtaining Information and
Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC–2019–
0062 when contacting the NRC about
the availability of information for this
action. You may obtain publicly
available information related to this
action by any of the following methods:
• Federal Rulemaking Website: Go to
https://www.regulations.gov and search
for Docket ID NRC–2019–0062.
• NRC’s Agencywide Documents
Access and Management System
(ADAMS): You may obtain publicly
available documents online in the
ADAMS Public Documents collection at
https://www.nrc.gov/reading-rm/
adams.html. To begin the search, select
‘‘Begin Web-based ADAMS Search.’’ For
problems with ADAMS, please contact
the NRC’s Public Document Room (PDR)
reference staff at 1–800–397–4209, at
301–415–4737, or by email to
PDR.Resource@nrc.gov. For the
convenience of the reader, instructions
about obtaining materials referenced in
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this document are provided in the
‘‘Availability of Documents’’ section.
• NRC’s PDR: The PDR, where you
may examine and order copies of
publicly available documents, is open
by appointment. To make an
appointment to visit the PDR, please
send an email to PDR.Resource@nrc.gov
or call 1–800–397–4209 or 301–415–
4737, between 8 a.m. and 4 p.m. eastern
time, Monday through Friday, except
Federal holidays.
B. Submitting Comments
The NRC encourages electronic
comment submission through the
Federal rulemaking website (https://
www.regulations.gov). Please include
Docket ID NRC–2019–0062 in your
comment submission. To facilitate NRC
review, please distinguish between
comments on the proposed rule and
comments on the proposed guidance.
The NRC cautions you not to include
identifying or contact information that
you do not want to be publicly
disclosed in your comment submission.
The NRC will post all comment
submissions at https://
www.regulations.gov as well as enter the
comment submissions into ADAMS.
The NRC does not routinely edit
comment submissions to remove
identifying or contact information.
If you are requesting or aggregating
comments from other persons for
submission to the NRC, then you should
inform those persons not to include
identifying or contact information that
they do not want to be publicly
disclosed in their comment submission.
Your request should state that the NRC
does not routinely edit comment
submissions to remove such information
before making the comment
submissions available to the public or
entering the comment into ADAMS.
II. Background
A. NRC Advanced Reactor Readiness
In its ‘‘Policy Statement on the
Regulation of Advanced Nuclear Power
Plants,’’ dated July 8, 1986, the
Commission stated that it considered
the term ‘‘advanced’’ to apply to
reactors that are significantly different
from current (i.e., current in 1986)
generation light-water reactors (LWRs)
then under construction or in operation,
and that ‘‘advanced’’ includes reactors
that provide enhanced margins of safety
or utilize simplified inherent or other
innovative means to accomplish their
safety functions. At the time, certain
high temperature gas-cooled reactors,
liquid metal reactors, and LWRs of
innovative design were considered to be
‘‘advanced.’’ The 1986 policy statement
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provided the Commission’s policy
regarding the review of, and desired
characteristics associated with,
advanced reactors. The NRC updated
this statement in the ‘‘Policy Statement
on the Regulation of Advanced
Reactors,’’ dated October 14, 2008
(Advanced Reactor Policy Statement).
The agency has undertaken many
activities related to advanced reactors,
including issuing an advance notice of
proposed rulemaking titled,
‘‘Approaches to Risk-Informed and
Performance-Based Requirements for
Nuclear Power Reactors,’’ dated May 4,
2006 (71 FR 26267). These efforts were
often done in parallel, and sometimes
interwoven, with the NRC’s efforts to
improve risk-informed and
performance-based approaches within
the agency (e.g., the Commission’s
policy statement, ‘‘Use of Probabilistic
Risk Assessment Methods in Nuclear
Regulatory Activities,’’ dated August 16,
1995 (PRA Policy Statement)).
In 2016, the NRC issued ‘‘NRC Vision
and Strategy: Safely Achieving Effective
and Efficient Non-Light-Water Mission
Readiness’’ (Advanced Reactor Vision
and Strategy Document), in response to
increasing interest in advanced reactor
designs. The NRC considered the
Department of Energy’s (DOE’s)
advanced reactor deployment goals in
developing the Advanced Reactor
Vision and Strategy Document. Since
publication of the document, the NRC
continues to manage its activities to
support the DOE’s deployment goals.
The Advanced Reactor Vision and
Strategy Document identified initiating
and developing a new risk-informed and
performance-based regulatory
framework as a possible long-term goal.
However, the NRC staff’s initial efforts
were focused on resolving policy issues
and developing guidance for licensing
non-LWR technologies under the
existing regulatory frameworks (parts 50
and 52). The NRC staff issues annual
Commission papers on the status and
progress of the NRC staff’s activities
related to advanced reactors (e.g.,
SECY–24–0020, ‘‘Advanced Reactor
Program Status,’’ dated February 27,
2024). These Commission papers
provide status updates for advanced
reactor activities undertaken both prior
to and after initiation of this
rulemaking.
In 2017, the NRC staff prioritized
activities to support the development of
technology-inclusive, risk-informed,
and performance-based licensing
approaches that could be implemented
under the existing regulatory framework
in parts 50 and 52. One key element of
these efforts was the Licensing
Modernization Project (LMP), a cost-
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shared initiative led by nuclear utilities
and supported by DOE. The LMP is a
technology-inclusive, risk-informed,
and performance-based methodology
developed for non-LWR designs. The
LMP provides a systematic and
reproducible process for licensing-basis
event (LBE) selection and evaluation;
classification of structures, systems, and
components (SSCs); and assessment of
defense in depth. The LMP refined the
DOE’s Next Generation Nuclear Plant
Program methodologies to reflect
interactions with the NRC, to address
feedback from industry, and to broaden
the scope of the approach to ensure
applicability to various non-LWR
technologies. The LMP activities led to
the publication and submittal of Nuclear
Energy Institute (NEI) 18–04, Revision 1,
‘‘Risk-Informed Performance-Based
Technology Inclusive Guidance for NonLight Water Reactor Licensing Basis
Development,’’ issued August 2019. The
document indicates that controlling the
frequencies and potential consequences
of a wide spectrum of events is the
primary focus of the LMP approach.
The NRC endorsed the principles and
methodology in NEI 18–04, with
clarifications, in RG 1.233, ‘‘Guidance
for a Technology-Inclusive, RiskInformed, and Performance-Based
Methodology to Inform the Licensing
Basis and Content of Applications for
Licenses, Certifications, and Approvals
for Non-Light-Water Reactors.’’ The
NRC staff sought Commission approval
of the use of LMP and NEI–18–04 in
SECY–19–0117, ‘‘Technology-Inclusive,
Risk-Informed, and Performance-Based
Methodology to Inform the Licensing
Basis and Content of Applications for
Licenses, Certifications, and Approvals
for Non-Light-Water Reactors,’’ dated
December 2, 2019. In that paper, the
staff described the relationship between
the LMP and NEI–18–04 and previous
relevant Commission decisions,
including those described in SECY–93–
092, ‘‘Issues Pertaining to the Advanced
Reactor (PRISM, MHTGR, and PIUS)
and CANDU 3 Designs and their
Relationship to Current Regulatory
Requirements,’’ dated April 8, 1993. The
Commission approved the use of the
LMP methodology and NEI–18–04 as a
reasonable approach for establishing key
parts of the licensing basis and content
of applications for licenses,
certifications, and approvals for nonLWRs in Staff Requirements
Memorandum (SRM) SRM–SECY–19–
0117, dated May 26, 2020. Although the
LMP approach is technology- inclusive,
the industry and NRC staff initially
focused the LMP’s applicability on nonLWRs, both for efficiency and to support
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near-term non-LWR applications under
the existing regulatory framework, such
as the Advanced Reactor Demonstration
Projects supported by DOE.
As stated in the part 53 rulemaking
plan, SECY–20–0032, the NRC staff
developed part 53 by building upon
recent and ongoing activities such as the
LMP approach described in SECY–19–
0117. Such an approach supports
implementing the NEIMA requirement
to use, where appropriate, risk-informed
and performance-based techniques, and
it also capitalizes on previous initiatives
by the industry, DOE, and the NRC,
including the LMP. This approach
highlights the role of PRA in riskinformed and performance-based
approaches to identifying enhanced
safety margins that can be used to justify
operational flexibilities. The proposed
framework is largely based on the
methodology described in SECY–19–
0117 and includes a prominent role for
PRA.
As discussed in section II.B,
‘‘Stakeholder Views on Part 53
Preliminary Proposed Rule Language,’’
of this document, the NRC conducted
extensive public outreach on early
versions of the proposed rule text. Early
versions of the draft proposed rule
included two alternative regulatory
frameworks. One framework (called
‘‘Framework A’’) offered a licensing
approach centered largely on risk
analysis and the other framework
(called ‘‘Framework B’’) largely
replicated the existing licensing
approach in parts 50 and 52 but
modified it to be technology neutral. In
its SRM to SECY–23–0021, ‘‘Proposed
Rule: Risk-Informed, TechnologyInclusive Regulatory Framework for
Advanced Reactors (RIN 3150–AK31),’’
the Commission disapproved the
inclusion of Framework B in this
proposed rule and directed the staff to
provide them within one year an
options paper for possible future use of
the Framework B methodology.
B. Stakeholder Views on Part 53
Preliminary Proposed Rule Language
In SRM–SECY–20–0032, the
Commission directed the NRC staff to
prepare and release preliminary
proposed rule language, followed by
public outreach and dialogue, and then
further revise the language until the
NRC staff had established the rudiments
of its proposed rule for Commission
consideration. To implement the
Commission’s direction, the NRC staff
undertook an unprecedented program of
stakeholder engagement, recognizing the
importance of this rulemaking to the
advanced reactor community and
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interested stakeholders from a broad
range of backgrounds and organizations.
On November 6, 2020, the NRC
published a notification in the Federal
Register (85 FR 71002) describing plans
for the periodic release of preliminary
proposed rule language, meetings with
stakeholders, and the ability of
stakeholders to provide input during the
development of this proposed rule.
Sections of the preliminary proposed
rule language were subsequently
released, and the NRC held numerous
public meetings to discuss the
preliminary proposed rule language and
obtain input from stakeholders. On
December 10, 2021, the NRC published
a second notification in the Federal
Register (86 FR 70423) announcing that
the development of the proposed rule
and related interactions with
stakeholders were being extended until
August 31, 2022.
By the close of the public stakeholder
interactions on August 31, 2022, the
NRC staff had held 24 public meetings
since September 2020. The NRC staff
also met with the Advisory Committee
on Reactor Safeguards (ACRS) in 16
public meetings during this period. By
the close of the public engagement
period on the preliminary proposed rule
language, 126 letters were received on
the preliminary proposed rule language.
Of these 126 letters, 21 were from nongovernmental organizations, 31 were
from the public, one was from Congress,
and the remaining 73 letters were from
NRC licensees, the NEI, and other
industry groups. In addition, the ACRS
wrote four interim letter reports to the
Chair on this rulemaking and issued its
final letter report on November 22,
2022. The letters from stakeholders
provided various points of view and
suggestions for clarifications, additions,
and deletions to the preliminary
proposed rule language. Copies of these
letters may be viewed and downloaded
from the Federal rulemaking website
https://www.regulations.gov, under
docket number NRC–2019–0062. The
inputs received were considered in the
development of this proposed rule.
However, as described during the
various public interactions related to
this rulemaking and in supporting
documents, the NRC will not formally
disposition the questions and
suggestions related to the preliminary
proposed rule language as it will for
public comments received following the
publication of this proposed rule.
III. Discussion
A. Objective and Applicability
The NRC is proposing to add a new,
alternative part to its regulations that
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would set out a risk-informed,
technology-inclusive framework for the
licensing and regulation of commercial
nuclear plants. This new approach
would achieve the following: (1)
continue to provide reasonable
assurance of adequate protection of
public health and safety and the
common defense and security; (2)
promote regulatory stability,
predictability, and clarity; (3) reduce
requests for exemptions from the
current requirements in parts 50 and 52;
(4) establish new requirements to
address non-LWR technologies; (5)
recognize technological advancements
in reactor design; and (6) credit the
possible response of some designs of
commercial nuclear plants to postulated
accidents, including slower transient
response times and relatively small and
slow release of fission products. This
proposed rule would add 10 CFR part
53; subpart M, ‘‘Fitness for Duty
Programs for Facilities Licensed Under
10 CFR Part 53,’’ to Part 26; § 73.100,
‘‘Technology-inclusive requirements for
physical protection of licensed activities
at commercial nuclear plants against
radiological sabotage,’’ § 73.110,
‘‘Technology-inclusive requirements for
protection of digital computer and
communication systems and networks,’’
and § 73.120, ‘‘Access authorization
program for commercial nuclear
plants,’’ as well as make conforming
changes throughout 10 CFR chapter I,
‘‘Nuclear Regulatory Commission.’’
B. Need for Changes to the Existing
Regulatory Framework
The NRC has long recognized that the
licensing and regulation of a variety of
nuclear reactor technologies would
present challenges because the existing
regulatory framework has evolved
primarily to address the LWR designs
that compose the current operating fleet
(widely referred to as Generation II
reactors). The NRC has had many
interactions with designers of various
reactor technologies under
development, sometimes collectively
referred to as advanced reactors (widely
referred to as Generation III/III+ (i.e.,
evolutionary light-water) and
Generation IV (i.e., non-light-water)
reactors). The interactions have
informed the development of policies
and guidance to support the potential
licensing of new and different types of
reactor facilities, some of which may not
utilize LWR designs. The NRC issued its
Advanced Reactor Policy Statement to
provide all interested parties, including
the public, with the Commission’s
views concerning the desired
characteristics of advanced reactor
designs. The NRC further described its
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early efforts to establish a technologyinclusive approach to the regulation of
nuclear reactors in the advance notice of
proposed rulemaking published in 2006.
The NRC acknowledged in its ‘‘Report
to Congress: Advanced Reactor
Licensing,’’ issued August 2012, that
while the safety philosophy inherent in
the current regulations applies to all
reactor technologies, the specific and
prescriptive aspects of those regulations
clearly focus on the current fleet of LWR
facilities.
Congress similarly recognized the
potential benefits of developing a
regulatory infrastructure to support the
development and commercialization of
advanced nuclear reactors.
Consequently, Congress passed NEIMA
in late 2018, and the President signed it
into law in January 2019. NEIMA
directed the NRC to undertake a
rulemaking to establish a technologyinclusive regulatory framework for
optional use by applicants for new
commercial advanced nuclear reactor
licenses. In addition, on July 9, 2024,
the President signed into law the
Accelerating Deployment of Versatile,
Advanced Nuclear for Clean Energy Act
of 2024, also referred to as the
ADVANCE Act. The NRC is evaluating
its plans for implementing the
ADVANCE Act, including how its
regulations, as well as the proposed part
53 or future revisions to it, could be
used to address provisions in the
ADVANCE Act. The ADVANCE Act
contains provisions on a variety of
nuclear-related topics, such as micro
reactors, nuclear reactor license
application reviews, and nuclear fuel. In
Section VI, ‘‘Specific Requests for
Comments,’’ the NRC is requesting
public input on how part 53 could be
revised to better enable its potential use
to implement the ADVANCE Act.
The requirements in part 53 would
support a wide variety of potential
commercial nuclear reactor
technologies. As noted in this
discussion, the current regulatory
framework in parts 50 and 52 evolved
in the context of the current operating
reactor fleet dominated by LWRs and as
a result includes provisions specific to
LWR technologies. While the NRC can
license other reactor technologies under
the current framework by using existing
regulatory flexibilities and the
exemption process, there is significant
interest in developing a regulatory
framework that is flexible enough to
accommodate multiple technologies and
robust enough to ensure a level of safety
equivalent to parts 50 and 52, consistent
with the Commission’s Advanced
Reactor Policy Statement. The
Commission reiterated its safety
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expectations for new reactors in the
SRM for SECY–10–0121, ‘‘Modifying
the Risk-Informed Regulatory Guidance
for New Reactors,’’ dated March 2, 2011:
Because new plant designs incorporate
operating experience from current generation
reactors, severe accident research, and risk
insights from design probabilistic risk
assessments, the Commission expects that
the advanced technologies incorporated in
new reactors will result in enhanced margins
of safety. However, the Commission
continues to expect (consistent with the 2008
Advanced Reactor Policy Statement), as a
minimum, at least the same degree of
protection of the public and the environment
that is required for current-generation lightwater reactors. New reactors with these
enhanced margins and safety features should
have greater operational flexibility than
current reactors.
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However, developing a regulatory
framework that can accommodate a
wide range of technologies while
maintaining an acceptable level of safety
presents significant regulatory
challenges. The existing regulations
have been developed over the course of
decades and reflect changes to address
events discovered through operating
experience. In contrast, part 53 is being
developed to accommodate technologies
that, in some cases, lack significant
operating experience. To address these
challenges, the NRC drew on welldeveloped approaches to licensing to
produce a technology-neutral and robust
regulatory framework. The proposed
regulatory framework would use PRAs
to assess risks, help establish technical
requirements, and manage operations.
The framework builds on the LMP,
which is a technology-inclusive
approach to licensing that leverages
insights from a detailed PRA to provide
applicants with significant design and
operation flexibilities.
C. 10 CFR Part 53: Framework
This proposed rule consists of several
major components, including a new part
53, to be added to 10 CFR chapter I,
revisions for part 26, part 50, and part
73, and conforming changes throughout
10 CFR chapter I.
Part 53 is comprised of subparts A
through M. These provisions are
organized to provide high-level
performance criteria and to specify
requirements to demonstrate
compliance with those performance
criteria throughout major stages of the
life cycle of commercial nuclear plants.
This organization reflects a systemsengineering style approach to the
design, licensing, operation, and
ultimately decommissioning of future
commercial nuclear plants. Organizing
requirements in this manner also
supports performance-based
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approaches. Required programs (e.g.,
radiation protection) and monitoring
(e.g., technical specification (TS)
surveillance) during the operations
phase that are similar to those required
by part 50 would complement the
design and analysis requirements in
subpart C. The performance-based
approach proposed in part 53 also
includes regulatory requirements that
would allow applicants to use a flexible
and graded approach to the performance
of safety functions based on the role of
a particular SSC, human action, or
program in limiting the overall risks to
the public below accepted standards
through balanced measures to prevent
and mitigate possible events.
Proposed subpart M of part 26 would
be new and would be largely consistent
with the objective-based fitness for duty
(FFD) requirements in current subpart
K, ‘‘FFD Programs for Construction,’’ of
part 26 supplemented by select
requirements from subparts A through I,
N, and O of part 26. These requirements
are designed to ensure program
effectiveness, maintain protections
afforded to individuals subject to the
FFD program, and align with FFD
program implementation by parts 50
and 52 licensees. The proposed
requirements are not entirely equivalent
because current subpart K of part 26
only applies during construction of the
commercial nuclear plant, whereas
proposed subpart M of part 26 would
apply during construction, operation,
and decommissioning. Furthermore,
proposed subpart M of part 26 would
allow the use of a variety of biological
specimens for drug testing as well as
innovative technologies for drug and
alcohol screening and testing that are
not described or allowed by the
requirements in subparts A through K,
N, and O of part 26, except under
limited conditions.
Proposed revisions to part 73 would
establish a new technology-inclusive
consequence-based approach for a range
of security areas, including physical
security, cybersecurity, and access
authorization (AA) for commercial
nuclear reactors. The NRC used
operating experience to include
additional regulatory flexibility for a
part 53 licensee’s implementation of
security requirements.
In addition, this proposed rule would
make conforming changes throughout
10 CFR chapter I, by adding ‘‘and part
53’’ where appropriate to account for
the addition of the proposed part 53.
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IV. Part 53: Framework
Subpart A—General Provisions
Subpart A would provide the general
provisions applicable to all applicants
and licensees that would be established
in part 53 for the issuance, amendment,
and termination of licenses, permits,
certifications, and approvals for
commercial nuclear plants licensed
under Section 103 of the Atomic Energy
Act of 1954, as amended (the Act) and
title II of the Energy Reorganization Act
of 1974 (88 Stat. 1242). Subpart A
would include purpose, scope,
definitions, written communications,
employee protections, completeness and
accuracy of information, exemptions,
standards for review, jurisdictional
limits, consideration of attacks and
destructive acts by enemies of the
United States, and information
collection requirements.
The requirements in subpart A would
be largely equivalent to the general
requirements in part 50 that are
applicable to all part 50 applicants and
licensees (specifically, §§ 50.1 through
50.13) but would reference the
corresponding regulations in part 53 in
place of references to part 50.
A. Discussion of Definitions in Proposed
Part 53
This proposed rule would include a
definition section in § 53.020. The
definitions of most terms in § 53.020
would be equivalent to the
corresponding terms defined in: (1)
§§ 50.2, 52.1, and other NRC
regulations; (2) NEI 18–04, as endorsed
by RG 1.233; or (3) American Society of
Mechanical Engineers (ASME)/
American Nuclear Society Risk
Assessment Standard (RA–S)-1.4–2021,
as endorsed for trial use by RG 1.247,
‘‘Acceptability of Probabilistic Risk
Assessment Results for Non-Light-Water
Reactor Risk-Informed Activities.’’ This
is intended to provide clarity and
consistency in terminology where
possible and to utilize past and ongoing
NRC initiatives to support the licensing
of new reactors. Specific deviations
from existing definitions are further
explained in the following paragraphs.
Regarding the definition of
‘‘Commercial nuclear plant’’ and
‘‘Commercial nuclear reactor’’ in
proposed § 53.020, as noted previously,
the NRC initially considered
establishing the scope of part 53 as
being for ‘‘advanced nuclear plants.’’
The preliminary proposed rule language
defined ‘‘advanced nuclear plant’’ as ‘‘a
utilization facility consisting of one or
more advanced nuclear reactors’’ as
defined in NEIMA. NEIMA defines the
term ‘‘advanced nuclear reactor’’ as ‘‘a
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nuclear fission reactor or fusion
machine, including a prototype plant (as
defined in sections 50.2 and 52.1 of title
10, Code of Federal Regulations (as in
effect on the date of enactment of this
Act)), with significant improvements
compared to commercial nuclear
reactors under construction as of the
date of enactment of this Act, including
improvements such as—(A) additional
inherent safety features; (B) significantly
lower levelized cost of electricity; (C)
lower waste yields; (D) greater fuel
utilization; (E) enhanced reliability; (F)
increased proliferation resistance; (G)
increased thermal efficiency; or (H)
ability to integrate into electric and
nonelectric applications.’’
Based on public discussions on the
use of the term, the NRC determined
that the NEIMA definition, although
broad, did not define ‘‘significant
improvements’’ with enough specificity
to implement in NRC regulations.
Additionally, a number of stakeholders
suggested that the descriptor,
‘‘advanced,’’ implied enhanced safety,
while the NEIMA definition includes
‘‘significant improvements’’ in areas
other than safety enhancements. In
response to this feedback, and to be
technology inclusive, the NRC
determined that the broader term
‘‘commercial nuclear plant’’ would be
preferable. The NEIMA definition of
advanced nuclear reactor also includes
fusion technologies. Fusion energy
systems have not been included in the
scope of part 53 but are the subject of
a separate rulemaking activity,
‘‘Regulatory Framework for Fusion
Systems.’’ See NRC docket ID NRC–
2023–0017 on the Federal rulemaking
website https://www.regulations.gov.
The NRC proposes to allow use of part
53 by any ‘‘commercial nuclear plant.’’
The use of the term ‘‘plant’’ versus
‘‘reactor,’’ as used in existing
regulations (i.e., § 50.2), recognizes that
co-located support facilities and
radionuclide sources need to be
considered in the licensing of a facility.
The phrase ‘‘commercial purposes,’’ as
used in the definition of ‘‘commercial
nuclear plant,’’ includes purposes such
as providing process heat for a variety
of industrial applications (e.g.,
desalination, oil refining, hydrogen
production). The NRC has not compiled
a complete list of such commercial
purposes. The definition of
‘‘Commercial nuclear plant’’ refers to a
‘‘Commercial nuclear reactor,’’ which is
defined based on the definition of
‘‘Nuclear reactor’’ in § 50.2. However,
the phrase ‘‘in a self-supporting chain
reaction’’ was removed from the
definition to enable applying part 53 to
accelerator driven systems that use
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special nuclear material (SNM) but that
do not involve self-sustaining chain
reactions. Relatedly, ‘‘Utilization
facility’’ is also defined in § 53.020
based on the definition of that term in
§ 50.2 but is also revised to refer to a
‘‘Commercial nuclear plant’’ as defined
in § 53.020.
The NRC proposes to include a
definition of ‘‘Consensus code or
standard’’ in part 53 that is based on the
use of these terms in the National
Technology Transfer and Advancement
Act of 1995 (NTTAA) (Pub. L. 104–113)
and the Office of Management and
Budget (OMB) Circular No. A–119,
‘‘Federal Participation in the
Development and Use of Voluntary
Consensus Standards and in Conformity
Assessment Activities.’’ As required by
NTTAA, the NRC undertakes the
following activities: (i) consults with
voluntary consensus standards bodies;
(ii) participates with voluntary
consensus bodies in the development of
consensus standards; and (iii) uses
consensus standards as a means to carry
out the NRC’s policy objectives. In part
53, the NRC is not proposing to
incorporate by reference specific codes
and standards as is done under the
existing regulations in § 50.55a, ‘‘Codes
and standards,’’ because some codes
and standards are LWR-specific. Part 53
would require that design features must
be designed using generally accepted
consensus codes and standards but
would not incorporate the specific code
or standard into the NRC’s regulations.
During public meetings, significant
discussions with stakeholders indicated
that future reactor designers were
interested in the use of international
consensus standards that have not yet
been endorsed by the NRC. The
definition proposed in part 53 would
allow for the use of international codes
and standards not previously used in
NRC licensing but recognizes that the
use of any consensus code or standard
would ultimately need to be found
acceptable by the NRC, either through
generic efforts to endorse a code or
standard or on an application-specific
basis during an individual licensing
review.
The proposed definition of
‘‘Construction’’ is slightly different than
the definition in § 50.10—it would cover
the same concept but be applied to a
slightly different scope of activities
based on how SSCs are classified under
part 53. In part 53, the definition of
‘‘Construction’’ is based on the
definition in § 50.10 but modified to
apply to safety-related (SR) and nonsafety-related but safety-significant
(NSRSS) SSCs identified by the design
and analysis requirements in subparts B
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and C to ensure the safety criteria are
met.
Section 53.020 would also add
definitions for terms related to event
selection (LBEs, design-basis accidents
(DBAs), anticipated event sequences,
unlikely event sequences, and very
unlikely event sequences); equipment
classifications (SR, NSRSS, and nonsafety-significant SSCs); performance
metrics (e.g., safety criteria and
functional design criteria); and special
treatment.
The regulation would define ‘‘Safety
criteria’’ in terms of the plant-level
performance-based metrics that would
be provided in §§ 53.210 and 53.220.
The term ‘‘Functional design criteria’’
would be defined as metrics for the
performance of specific SSCs that are
determined from the role of the SSC in
meeting the safety criteria. These are
new terms that have not previously been
defined or used in NRC regulation.
The term ‘‘Safety-related SSCs’’
would refer to those SSCs needed to
meet the safety criteria in § 53.210. The
term ‘‘Non-safety-related but safetysignificant SSCs’’ would mean those
SSCs that are not SR because they are
not relied upon to perform any function
necessary to demonstrate compliance
with § 53.210 but warrant special
treatment because they are relied on to
achieve adequate defense in depth or
perform risk-significant functions. The
term ‘‘Special treatment’’ would be
defined as requirements, such as quality
assurance and programmatic controls,
identified for each design feature to
ensure that the safety criteria are
satisfied and the safety functions are
fulfilled. These requirements would also
ensure that SR and NSRSS SSCs will
provide defense in depth, or perform
risk-significant functions, under service
conditions and with SSC reliabilities
that are consistent with the analysis
required in proposed subpart C.
Structures, systems, and components
designated as SR would also contribute
to defense in depth and risk-significant
functions and may warrant special
treatments beyond those defined for the
SR functions needed for compliance
with § 53.210. The term ‘‘Non-safetysignificant SSCs’’ would mean those
SSCs that are not SR or NSRSS.
The terms ‘‘Design-basis accidents,’’
‘‘Anticipated event sequences,’’
‘‘Unlikely event sequences,’’ and ‘‘Very
unlikely event sequences’’ would be
defined to be different types of
‘‘Licensing-basis events’’ and would also
be largely equivalent to the LMP’s
definitions of DBAs, anticipated
operational occurrences (AOOs), designbasis events (DBEs), and beyond-designbasis events, respectively. The term
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‘‘Design-basis accidents’’ would be
defined as postulated event sequences
that are used to set functional design
criteria and performance objectives for
the design of SR SSCs through
deterministic analyses. Design-basis
accidents would be derived from the
unlikely event sequences from the PRA
and then analyzed in a conservative
approach by prescriptively assuming
that only SR SSCs are available to
mitigate postulated accident scenarios.
Within the LMP methodology, event
sequences with mean frequencies of 1 ×
10¥2/plant-year and greater would be
classified as anticipated event
sequences. Within the LMP
methodology, infrequent event
sequences with mean frequencies of 1 ×
10¥4/plant-year to 1 × 10¥2/plant-year
would be classified as unlikely event
sequences. ‘‘Very unlikely event
sequences’’ would be less likely to occur
than unlikely event sequences. Within
the LMP methodology, rare event
sequences with frequencies of 5 × 10¥7/
plant-year to 1 × 10¥4/plant-year would
be classified as very unlikely event
sequences. While the proposed
terminology for these event sequences
would create some differences between
part 53 and the LMP, part 53 would use
new terms for these event sequences
specifically to avoid conflicts with
terms already used within part 50 and
part 52 to represent different concepts.
Further, because some stakeholder
comments demonstrated confusion
related to the history of beyond-designbasis accidents terminology, these
definitions seek to clarify the event
categories in part 53. The sections of
this preamble related to subparts B and
C provide additional discussion of
LBEs.
B. Other General Provisions
Section 53.040 would govern written
communications and how applications
and other required information must be
submitted to the NRC. These
requirements would be equivalent to
those in § 50.4.
Section 53.050 would establish
requirements for enforcement action to
which a licensee, an applicant, or a
licensee’s or applicant’s contractor or
subcontractor, or an employee of any of
them may be subject for engaging in
deliberate misconduct. These
requirements would be equivalent to
those in § 50.5.
Section 53.060 would prohibit
discrimination against an employee of a
holder or applicant for an NRC license,
permit, design certification (DC), or
design approval, or a contractor or
subcontractor of a holder or applicant
for an NRC license, permit, DC, or
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design approval for engaging in certain
protected activities. Section 53.060 also
would prescribe a procedure for seeking
a remedy for employees who believe
they have been discriminated against for
engaging in such protected activities.
These requirements would be
equivalent to those in §§ 50.7 and 52.5.
Section 53.070 would govern the
completeness and accuracy of
information provided to the NRC. These
requirements would be equivalent to
those in §§ 50.9 and 52.6.
Section 53.080 would govern
exemptions from the requirements of
the regulations in part 53. These
requirements would be equivalent to
those in §§ 50.12 and 52.7.
Paragraphs (a) through (d) of § 50.90
would establish requirements for
standards that the NRC would consider
in determining whether a construction
permit (CP), operating license (OL),
early site permit (ESP), combined
license, or manufacturing license (ML)
under part 53 would be issued to an
applicant. These requirements would be
equivalent to those in §§ 50.40, 50.42,
50.43 and 50.22, respectively.
Requirements equivalent to those in
§§ 50.41 and 50.21 would not be
included in part 53 because they apply
to Class 104 licenses, and part 53 would
not apply to those licenses.
Section 53.100 would require that no
license issued under part 53 would
cover activities which are not under or
within the jurisdiction of the United
States. These requirements would be
equivalent to those in § 50.53.
Section 53.110 would state that
licensees and applicants would not be
required to provide design features or
other measures for the specific purpose
of protection against the effects of
attacks and destructive acts by enemies
of the United States directed against the
facility or deployment of weapons
incident to U.S. defense activities.
These requirements would be
equivalent to those in § 50.13.
Section 53.115 would establish
requirements for rights related to SNM.
These requirements would be
equivalent to those in § 50.54(b) and (c).
Section 53.117 would establish
requirements for license suspension and
rights of recapture of the material or
control of the facility in a state of war
or national emergency declared by
Congress. These requirements would be
equivalent to those in § 50.54(d).
Section 53.120 would establish
requirements for information collection
requirements and OMB approval. These
requirements would be equivalent to
those in § 50.8.
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Subpart B—Technology-Inclusive Safety
Requirements
Proposed subpart B, ‘‘TechnologyInclusive Safety Requirements,’’ would
provide technology-inclusive safety
criteria that would serve as performance
standards for the subsequent
performance-based requirements used
throughout part 53. Subsequent subparts
would define how specific activities
during various stages of the life cycle of
a commercial nuclear plant contribute
to satisfying these high-level
performance standards. The
performance standards in subpart B
would also establish a means to
determine appropriate regulatory
controls for SSCs, human actions, and
programs in the following subparts. For
example, the classification of SR SSCs
would be built upon the proposed safety
criteria in § 53.210, ‘‘Safety criteria for
design-basis accidents.’’ The more
detailed requirements for those SSCs
would then be further defined in the
design and analysis requirements in
subpart C, ‘‘Design and Analysis
Requirements.’’ The activities for
manufacturing, constructing, and
maintaining the SR SSCs would be
governed by subpart E, ‘‘Construction
and Manufacturing Requirements,’’ and
subpart F, ‘‘Requirements for
Operation.’’
Requirements for NSRSS SSCs
warranting special treatment would
likewise be determined under § 53.220,
‘‘Safety criteria for licensing-basis
events other than design-basis
accidents,’’ in subpart B and § 53.460,
‘‘Safety categorization and special
treatment,’’ in subpart C. Regulatory
requirements related to the NSRSS SSCs
would be distinguished from the
regulatory requirements for SR SSCs
throughout part 53. Part 53 would afford
more flexibility to applicants and
licensees regarding how NSRSS SSCs
would be used in the design and
maintained during plant operations, as
compared to SR SSCs.
The collective set of performancebased requirements in part 53 would be
sufficient, if met, for the NRC to make
the findings required to grant an
application for a utilization facility
under Section 182 of the Act that the
utilization of SNM will be in accord
with the common defense and security
and will provide adequate protection to
the health and safety of the public. This
construct would be similar to existing
NRC regulations, which the Commission
has said on many occasions do not
specifically define ‘‘adequate
protection.’’ However, compliance with
NRC regulations may be presumed to
assure adequate protection at a
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minimum. The requirements throughout
part 53 that support demonstrating
compliance with § 53.220 would be
similar to current regulations that both
contribute to assuring adequate
protection of public health and safety
and are desirable to promote the
common defense and security or to
protect health or to minimize danger to
life or property under Section 161 of the
Act.
Consistent with historical practice,
Sections 182 and 161 of the Act are
cited as authorizing legislation within
this proposed rule. However, specific
language from the Act would not be
incorporated into the safety objectives
or safety criteria in part 53. This is
because, again consistent with historical
practice, the NRC would not be defining
‘‘adequate protection’’ through the
individual safety requirements in part
53. Rather, part 53 would enable the
NRC to make its required findings under
the Act by providing sufficient
performance standards, safety criteria,
and related requirements on how
applicants must demonstrate
compliance with subpart B and other
subparts.
Section 53.210 would provide safety
criteria for DBAs that would be required
to be identified under § 53.240 and
analyzed under § 53.450(f) in subpart C
of part 53. Subsequent sections in part
53 would require that the SSCs relied
upon to demonstrate compliance with
the criteria in § 53.210 be classified as
SR. The use of SR SSCs and the 25 rem
reference values for potential
radiological consequences would align
with traditional deterministic
approaches for LWRs from §§ 50.34,
52.79, and 100.11 for evaluating the
effectiveness of plant design features
with respect to postulated reactor
accidents. A footnote similar to that
included in § 50.34(a)(1)(ii)(D)(1) and
§ 52.79(a)(1)(vi)(A) would be included
in § 53.210 to explain that the use of the
25 rem value would not be intended to
imply that this number constitutes an
acceptable limit for an emergency dose
to the public under accident conditions.
Rather, this dose value has been set
forth in this proposed section as a
reference value that would be used in
the evaluation of plant design features
with respect to DBAs to verify that the
proposed designs would provide
assurance of low risk of public exposure
to radiation in the event of an accident.
The inclusion of the safety criteria for
DBAs in subpart B would provide a
logical structure supporting the
identification and treatment of SR SSCs
and establishing the corresponding
functional design criteria for those
SSCs.
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Section 53.220 would provide safety
criteria for LBEs other than DBAs that
would be required to be identified
under § 53.240 and analyzed under
§ 53.450(e) in subpart C. Whereas
§ 53.210 and the related requirements
for SR SSCs would provide that a
defined success path exists for DBAs,
the safety criteria for LBEs other than
DBAs would establish the connections
between SSC design, human actions,
and programmatic controls and a
broader set of potential internal and
external hazards. These safety criteria
would also address defense-in-depth
matters such as a balanced
consideration of prevention and
mitigation.
The safety criterion in § 53.220(b)
would include a requirement to use a
comprehensive risk metric or set of
metrics and associated risk performance
objectives against which calculated
values of the risk metrics are compared.
The comprehensive risk metrics or set of
metrics and associated risk performance
objectives would support a
performance-based approach to
developing an appropriate combination
of design features and programmatic
controls to prevent or mitigate LBEs
other than DBAs. The applicant must
propose the comprehensive risk metric
or set of metrics and associated risk
performance objectives, and the
comprehensive risk metric or set of
metrics and associated risk performance
objectives must provide an appropriate
level of safety. Comprehensive risk
metrics should consist of a proposed
plant risk metric or set of proposed risk
metrics that approximate the total,
overall risk from the facility and that
address the range of possible plant
configurations and associated internal
and external hazards to the extent
practicable. The associated risk
performance objectives are
preestablished, indicative values of the
comprehensive risk metrics that are
used as part of risk-informed decisionmaking. The methodology for
developing and using proposed
comprehensive risk metrics and
associated risk performance objectives is
defined by the proposed requirements
for analyses in § 53.450. Therefore, the
application must include a description
of that methodology and, among other
things, should explain the initial
conditions, boundary conditions, and
key assumptions used to develop and
calculate the risk metrics. Screening
tools and bounding or simplified
methods may be used for any mode or
hazard, provided that the applicant
provides an acceptable technical basis.
As with all risk-informed
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methodologies, treatment of
uncertainties must be addressed.
The risk performance objectives
established under this methodology are
likely to involve assessing and averaging
the risks over a period of time (e.g.,
plant year) and would not constitute a
real-time requirement that must be
continuously demonstrated by the
licensee. The use of a comprehensive
risk metric or set of risk metrics and risk
performance objectives that reflect an
average risk to establish performance
goals for SR and NSRSS SSCs is
consistent with current practices that
use other risk assessment techniques to
address short-term plant configurations
during plant maintenance activities.
It is worth noting that the evaluation
of plant risks, as represented by a
comparison of analysis results to
acceptable risk performance objectives
for comprehensive risk metrics, would
be one of several performance standards
used in subpart B. The proposed use of
multiple performance standards,
including deterministic criteria and
defense-in-depth measures, reflects an
integrated decision-making process
similar to that described in RG 1.174,
‘‘An Approach for Using Probabilistic
Risk Assessment in Risk-Informed
Decisions on Plant-Specific Changes to
the Licensing Basis,’’ Revision 3. The
NRC’s approval of using a
comprehensive risk metric or set of
metrics with associated risk
performance objectives is not, by itself,
an indicator of adequate protection.
Rather, the comparison of
comprehensive risk metrics to
associated risk performance objectives
that are acceptable to the NRC is part of
a suite of regulatory requirements that,
when considered holistically, form the
basis for the NRC’s decision-making.
This is analogous to the approach used
for plants licensed under part 50 and
part 52, where no single regulatory
requirement governs whether a plant is
‘‘safe enough.’’
The RG 1.233, ‘‘Guidance for a
Technology-Inclusive, Risk-Informed,
and Performance-Based Methodology to
Inform the Licensing Basis and Content
of Applications for Licenses,
Certifications, and Approvals for NonLight-Water Reactors,’’ describes an
example of an acceptable approach for
identifying and analyzing LBEs under
part 50 and part 52, including the use
of the quantitative health objectives
(QHOs) stated in the NRC’s policy
statement, ‘‘Safety Goals for Nuclear
Power Plant Operation,’’ dated August
4, 1986 (51 FR 28044), as corrected and
republished August 21, 1986 (51 FR
30028) (Safety Goals Policy Statement),
as acceptable performance objectives for
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comprehensive risk metrics. The use of
comprehensive risk metrics, such as the
individual early fatality risk (IEFR) and
the individual latent cancer fatality risk
(ILCFR), and associated risk
performance objectives, such as the
QHOs, from the Safety Goals Policy
Statement, could form the basis for one
approach to meet § 53.220(b). The
requirement for comprehensive risk
metrics, in combination with the other
proposed requirements in subparts B
and C, would bring the approach
endorsed in RG 1.233 for parts 50 and
52 into part 53. Additionally, the use of
comprehensive risk metrics and
associated risk performance objectives
would provide a logical performance
objective to support the risk
management approaches in the various
subparts comprising proposed part 53.
The Commission stated in the
introduction of the Safety Goals Policy
Statement that improvements to thencurrent regulatory practices could lead
to a more coherent and consistent
regulation of nuclear power plants, a
more predictable regulatory process, a
better public understanding of the
regulatory criteria that the NRC applies,
and public confidence in the safety of
operating plants. Accordingly, the
Commission announced the safety goals
with a focus on the risks to the public
from nuclear power plant operation.
Following the issuance of the Safety
Goals Policy Statement, the NRC has
used the comprehensive risk metrics
and performance objectives provided in
the safety goals within the criteria for
many decisions involving safety
judgments during the licensing and
regulation of operating reactors and
proposed nuclear reactor designs.
Consistent with NUREG–0880, the
proposed comprehensive risk metrics
and associated risk performance
objectives required under § 53.220(b)
could be expressed in terms of a
biologically average individual in terms
of age and other risk factors. Although
some comprehensive risk objectives
such as the IEFR and ILCFR are defined
in terms of fatality risks, the
Commission continues to make clear
that no death attributable to nuclear
power plant operation will ever be
‘‘acceptable’’ in the sense that the
Commission would regard it as a routine
or permissible event. Comprehensive
risk metrics and associated risk
performance objectives as used in this
proposed rule would establish
acceptable risks, not acceptable deaths.
Applicants under the proposed part
53 may choose to develop and seek NRC
approval of comprehensive risk metrics
or sets of risk metrics and associated
risk performance objectives beyond
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those discussed above, including the
use of surrogate measures for use in
specific analyses to satisfy the proposed
requirements in § 53.220(b). Such
surrogate measures for comprehensive
risk metrics and associated risk
performance objectives could be used in
a manner similar to the use of core
damage frequency and conditional
containment failure probability for
LWRs within the safety goal evaluation
process in NUREG/BR–0058,
‘‘Regulatory Analysis Guidelines of the
U.S. Nuclear Regulatory Commission,’’
and other assessments of LWRs using
the NRC’s safety goals. The NRC would,
as appropriate, review novel approaches
for comprehensive metrics and
associated risk performance goals
proposed by applicants, industry
organizations, or standard development
organizations and would engage
stakeholders during the development of
the related regulatory guidance or
specific licensing actions.
Section 53.230 would require safety
functions needed to ensure that the
safety criteria under §§ 53.210 and
53.220 can be met if an assumed LBE
were to occur at a commercial nuclear
plant. Section 53.230 would specify that
limiting the release of radioactive
materials from the facility is the primary
safety function, and therefore, limiting
potential offsite consequences (i.e., dose
to a hypothetical individual) would be
used as the primary performance metric
throughout part 53. The additional or
subsidiary safety functions needed to
limit the release of radionuclides may
include, without limitation, controlling
processes related to reactivity, heat
generation, heat removal, and chemical
interactions. This proposed rule
provides flexibility to applicants and
licensees in identifying, implementing,
and maintaining the safety functions
supporting retention of radionuclides
for commercial nuclear plants of varying
sizes and technologies.
Proposed § 53.240 would require
applicants to identify and address LBEs.
LBEs are unplanned events, resulting
from both internal and external hazards,
that are used in the design and analyses
required under part 53 for licensing
commercial nuclear plants. This ensures
estimates of offsite consequences from
analyses performed under proposed
§ 53.450 are below the safety criteria
identified under proposed §§ 53.210 and
53.220 and that SSCs, personnel, and
programs address the safety functions
from proposed § 53.230. Including a
high-level performance requirement
related to the identification and analysis
of LBEs in subpart B would reflect the
historical and continuing importance of
evaluating unplanned events as part of
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the licensing of commercial nuclear
plants. Proposed § 53.240 would require
identification and analysis of LBEs
under § 53.450, which would require a
PRA. Examples of acceptable methods
of using PRAs to identify and assess
LBEs would be the methodology in RG
1.233, as discussed in Draft Regulatory
Guide (DG)–1413, ‘‘TechnologyInclusive Identification of Licensing
Events for Commercial Nuclear Plants.’’
Section 53.250 would establish
defense-in-depth requirements based on
the longstanding philosophy of
providing defense in depth to address
uncertainties about the design,
operation, and performance of
commercial nuclear plants. For
example, parts 50 and 52 address
defense in depth through layered
prescriptive technical requirements
(e.g., fuel performance, cladding
integrity, reactor coolant system
integrity, containment performance) for
LWRs. In contrast, the flexibility
afforded to applicants in how they
propose to demonstrate compliance
with the high-level safety criteria within
part 53 would necessitate this specific
requirement to ensure defense in depth
is provided. The requirements in this
section would state that no single
engineered design feature, human
action, or programmatic control, no
matter how robust, should be
exclusively relied upon to address LBEs
other than DBAs. The phrase
‘‘engineered design feature’’ would not
preclude the possible crediting of
inherent characteristics within the
design and analysis for commercial
nuclear reactors. While defense in depth
would only be assessed for LBEs other
than DBAs, the need to ensure
dedicated success paths for DBAs would
contribute to the overall defense in
depth for each commercial nuclear plant
under part 53.
Section 53.260 would govern normal
operations and would establish a level
of safety based on current requirements
in 10 CFR part 20, ‘‘Standards for
Protection Against Radiation,’’ which
limits doses to members of the public
and dose rates in unrestricted areas.
Section 53.270 would provide for the
protection of plant workers and would
establish a level of safety based on
current requirements in 10 CFR part 20
which limits occupational dose.
Subpart C—Design and Analysis
Requirements
This subpart would provide
requirements for the design of
commercial nuclear plants and the
supporting analyses, including the
analyses of LBEs, to demonstrate that
the performance standards in proposed
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subpart B can be satisfied. The sections
within subpart C would reflect the
overall hierarchy throughout part 53,
which would cover: (1) plant-level
safety criteria (§§ 53.210, 53.220, and
53.470); (2) safety functions (§ 53.230)
needed to demonstrate compliance with
the safety criteria; (3) design features
(§ 53.400), human actions, and
programmatic controls needed to fulfill
the safety functions; and (4) functional
design criteria (§§ 53.410 and 53.420)
that must be defined for each design
feature relied on to demonstrate the
safety criteria (§§ 53.210, 53.220, and
53.470) are met. Subpart C would also
contribute to the logic and structure of
part 53 by distinguishing between SR
SSCs and NSRSS SSCs and licenseecontrolled programs that address LBEs
other than DBAs. Specifically, SR SSCs,
human actions, and programmatic
controls needed to protect against DBAs
are used to satisfy the safety criteria in
§ 53.210. Non-safety-related but safetysignificant SSCs, human actions, and
licensee-controlled programs that
address LBEs other than DBAs generally
contribute to the appropriate measures
considering potential risks to public
health and safety.
Section 53.400 would establish a
requirement that design features be
provided for each commercial nuclear
plant to satisfy the safety criteria and
fulfill safety functions from proposed
subpart B during LBEs. Other sections
in subpart C would, in turn, further
address the necessary capabilities and
reliabilities for SSCs by establishing
functional design criteria, fulfilling
design requirements, performing
analyses of LBEs, performing other
supporting analyses, and categorizing
SSCs based on their roles in preventing
or mitigating LBEs.
Section 53.410 would require that
functional design criteria be defined for
design features relied upon to
demonstrate that the consequences from
DBAs would be below the criteria in
§ 53.210 through analyses performed
under § 53.450(f), which includes
insights from both PRAs and
deterministic analyses. Other sections
within part 53 would establish
appropriate controls on these design
features (e.g., safety classification,
protection from external hazards,
quality assurance, and TS) to ensure the
functional design criteria are satisfied.
The performance requirements for the
SSCs needed to address DBAs and the
corresponding human actions and
programmatic controls would contribute
to ensuring that a commercial nuclear
plant licensed under part 53 would
meet the safety criteria in § 53.210.
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Section 53.415 would require that SR
SSCs be protected against or designed to
withstand the effects of natural
phenomena (e.g., earthquakes,
tornadoes, hurricanes, floods, tsunami,
and seiches) and constructed hazards
(e.g., from dams, transportation routes,
and military or industrial facilities).
Specifically, § 53.415 would require that
SR SSCs remain capable of performing
the safety functions stated in § 53.230
for which they are credited up to the
design-basis external hazard levels as
determined under § 53.510. As used in
§ 53.415 and subpart D of part 53, a
hazard level would refer to such things
as the magnitude and recurrence rate of
an earthquake and the resultant ground
motions, the height of a flood, the force
of hurricane winds, or the
concentrations of chemicals resulting
from a release from a nearby facility.
These requirements would support
either traditional deterministic
approaches for determining and
protecting against external hazards or
probabilistic approaches that are being
developed for seismic and some other
external hazards.
Section 53.420 would require that
functional design criteria be defined for
design features that play a significant
role in demonstrating that the safety
criteria for LBEs other than DBAs are
satisfied. The analyses required for this
demonstration would be described in
proposed § 53.450(e), which would
require that those events be identified
and assessed using a PRA methodology
in combination with other generally
accepted approaches for systematically
evaluating engineered systems. The
SSCs determined to be safety significant
(i.e., either SR or NSRSS) would have
associated special treatment
requirements as specified in § 53.460.
Special treatment would be defined in
subpart A of part 53 and generally refers
to measures (e.g., quality assurance,
testing, monitoring) taken beyond the
procurement and installation of
commercial grade products to provide
confidence that the SSC will comply
with the applicable functional design
criteria. The inclusion of a systematic
approach to identifying the functional
design criteria for SSCs and tailoring the
special treatments to specific LBEs and
safety functions is an important
contributor to satisfy the proposed
safety criteria in subpart B. Therefore,
designers and licensees for commercial
nuclear plants would be provided
flexibility on how LBEs other than
DBAs are either prevented or mitigated
and how the calculated comprehensive
plant risks satisfy the safety criterion
established under § 53.220(b).
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Section 53.425 would establish
requirements for design features and
related functional design criteria
limiting doses to members of the public
during normal operations to satisfy the
criteria in part 20. Section 53.430 would
provide similar requirements for design
features and related functional design
criteria for protection of plant workers
to meet the safety criteria in part 20.
Similar to existing regulations, the NRC
considers that licensees would generally
comply with the requirements of part 20
to keep doses as low as reasonably
achievable by meeting a design objective
of keeping doses to the public from
routine plant effluents less than 10
millirem per year. This goal is similar to
that provided by appendix I to part 50
and would assist designers, applicants,
and licensees in performing the
evaluations of possible reductions in
public dose from routine effluents when
considering costs and other factors. As
emphasized in existing regulations in
part 50, the design objective of keeping
doses to the public from routine plant
effluents less than 10 millirem per year
should not be construed as a radiation
protection standard. The NRC
anticipates that future guidance will
continue to reflect this performance
goal.
The proposed requirements in
§§ 53.425 and 53.430 for design features
and functional design criteria to support
radiation protection activities have
parallels in existing regulations such as
§ 50.34(a) and (b)(3), which require in
part that the means be provided for
meeting the requirements of part 20 and
General Design Criterion 60, 61, 63, and
64 in appendix A to part 50, which
provide radiation protection related
design criteria.
Section 53.440 would address various
design requirements that warrant
specific mention to ensure that the
design features required by § 53.400
comply with the functional design
criteria required by §§ 53.410 and
53.420. These requirements would be
met through design practices,
consideration of testing and operating
experience, and various assessments of
LBEs and other potential challenges to
commercial nuclear plants. Discussions
of some of the key design requirements
included in this section follow.
• § 53.440(a): An essential element to
ensuring a proposed design can comply
with the performance criteria in
proposed part 53 would be that the
abilities of design features to fulfill their
safety functions are demonstrated by a
combination of analyses, test programs,
prototype testing, and operating
experience. This requirement closely
aligns with the language in § 50.43(e)
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and is proposed in part 53 as the same
foundational requirement. In addition,
the proposed § 53.440(a) would require
the design processes for SSCs under this
section to include administrative
procedures for evaluating operating,
design, and construction experience for
considering applicable important
industry experiences in the design of
those SSCs. This proposed requirement
corresponds to the existing requirement
under § 50.34(f)(3)(i) that was developed
in response to the 1979 accident at
Three Mile Island Nuclear Generating
Station.
• § 53.440(b): The design and
licensing of commercial nuclear plants
should use generally accepted
consensus codes and standards. Such
codes and standards ensure sufficient
testing and qualification of materials
and equipment and provide defined
processes, specifications, and
acceptance criteria for use by designers
and suppliers. The NRC would indicate
acceptance of consensus codes and
standards used in the design and
licensing of a specific commercial
nuclear plant either through the NRC’s
generic endorsement of a code or
standard (i.e., through regulatory
guidance), including any limitations or
conditions, that can be referenced
within an application, or through the
review of a referenced code or standard
as part of the review of a specific
application.
• § 53.440(c): The design
requirements in subpart C would
require the materials used for SR and
NSRSS SSCs to be qualified for their
service conditions over the design life of
the SSC.
• § 53.440(d): The requirements in
§ 53.440 would include the need to
consider possible degradation
mechanisms for materials and
equipment to inform both the design
process and the development of
integrity assessment programs to be
executed during plant operations in
accordance with subpart F of part 53.
The inclusion of requirements related to
designing and monitoring for possible
degradation mechanisms reflects
important lessons learned from the
history of LWRs as well as operating
experience with structures and systems
in countless other engineering
endeavors.
• § 53.440(e) and (f): The design
requirements in subpart C would state
specific design requirements similar to
existing requirements in parts 50, 52,
and 73 for protections against fires and
explosions and consideration of safety
and security together in the design
process.
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• § 53.440(g) and (h): Specific design
requirements are proposed to ensure
that commercial nuclear reactors under
part 53 have the capability to achieve
and maintain subcriticality and longterm cooling. The requirements would
be included to address the potential that
some reactor designs may be able to
achieve a stable end state for the
purpose of event analyses but might
need further actions to completely shut
down and service the facility.
• § 53.440(i): The design, analysis,
and development of programmatic
controls under part 53 would consider
the number of reactor units and other
significant inventories of radioactive
materials contributing to the risks to
public health and safety. This would
reflect the definition of ‘‘Commercial
nuclear plant’’ in subpart A and
reinforce that the evaluation of LBEs is
performed on a plant-wide basis. This
aspect of part 53 would be different
from parts 50 and 52, which generally
define safety requirements on the
assumption of events involving only
individual reactor units.
• § 53.440(j): A design requirement is
proposed to provide a technologyinclusive requirement that would be
equivalent to the requirements in
§ 50.150 to address the possible impact
of a large commercial aircraft.
• § 53.440(k): The inclusion of a
specific proposed requirement to
address the risks to public health from
potential chemical hazards of licensed
material is appropriate given the
diversity of reactor technologies and
designs that might be licensed under
part 53. The requirement in part 53
would be similar to the existing
requirements in 10 CFR part 70,
‘‘Domestic Licensing of Special Nuclear
Material,’’ that address both potential
radiological and chemical hazards for
licensed materials at fuel cycle facilities.
• § 53.440(l): Provisions are proposed
to require that measures be taken during
the design of commercial nuclear plants
to minimize contamination of the
facility and the environment, facilitate
eventual decommissioning, and
minimize the generation of radioactive
waste in accordance with § 20.1406.
• § 53.440(m): A design requirement
is proposed to provide a technologyinclusive equivalent to the requirements
in § 50.68 by including options for
commercial nuclear plants to either
have a monitoring system capable of
detecting a criticality as described in
§ 70.24 or to have restrictions on SNM
handling and storage that would prevent
inadvertent criticality events.
• § 53.440(n): The design would need
to reflect state-of-the-art human factors
principles for safe and reliable
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performance in all settings that human
activities are expected for performing or
supporting the continued availability of
plant safety or emergency response
functions.
Section 53.450 would establish
analysis requirements and would center
upon the use of a PRA in combination
with other generally accepted
approaches for systematically evaluating
engineered systems. The reliance on
PRAs as a key component in the
proposed analysis requirements for part
53 would reflect the decades of
improvements in PRA methodologies
and the increasing use of PRA
techniques in the design, licensing, and
oversight of both operating and future
nuclear reactors. Part of the
Commission’s PRA Policy Statement is
that the use of PRA technology should
be increased in all regulatory matters to
the extent supported by the state of the
art in PRA methods and data and in a
manner that complements the NRC’s
deterministic approach and supports the
NRC’s traditional defense-in-depth
philosophy. The need to supplement
PRA insights with other engineering
approaches and judgments reflects the
NRC’s longstanding policy described in
the SRM to SECY–98–144, ‘‘Staff
Requirements—SECY–98–144—White
Paper on Risk-Informed and
Performance-Based Regulations,’’ dated
February 24, 1999, for regulatory
decision-making to be risk-informed but
not solely based on numerical results of
a risk assessment (i.e., not a risk-based
approach). Part 53 would maintain a
role for NRC’s traditional deterministic
approaches (particularly for DBAs) and
defense-in-depth philosophy by
including specific requirements
utilizing these regulatory tools in
subparts B and C.
PRA would be used in combination
with other techniques in part 53 to
identify and categorize LBEs, classify
SSCs, and evaluate defense in depth.
This increased role for the PRA
necessitates that it would be developed,
performed, and maintained in
accordance with NRC-approved
standards and practices (see § 53.450(c)
and (d)). The computer codes used to
model the plant response and the
behavior of the barriers to the release of
radionuclides would need to be
qualified for the range of conditions
being simulated across a wide range of
unplanned events. These analyses
would need to use realistic approaches
and address uncertainties associated
with states of knowledge, modeling, and
performance of SSCs.
While industry consensus PRA
standards and peer review processes
endorsed in RGs 1.200 and 1.247 remain
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acceptable for developing a PRA, they
are not regulatory requirements and an
application under part 53 need not
follow every aspect of the applicable
consensus PRA standard. Existing
processes for defining the scope and
capability of a PRA supporting an
application offer flexibility in
determining the degree to which the
PRA needs to be developed and may be
informed by other factors such as design
complexity and the needed degree of
realism and level of detail, consistent
with the use of the PRA and substance
of the application. Such processes are
currently available for appropriately
defining the scope of the PRA and
determining applicability of supporting
requirements in consensus PRA
standards needed to satisfy the
proposed regulatory requirements for
the specific uses of analyses under
§ 53.450(b). Likewise, NRC
determinations of the acceptability of
such PRAs would include consideration
of the appropriateness of the applicantdefined scope as part of determining the
applicability of and conformance to
consensus PRA standard supporting
requirements consistent with the
current state of practice. In addition,
these determinations would include
consideration of other aspects of the
development of the PRA, such as PRA
peer reviews. An NRC determination of
the acceptability of a PRA includes but
is not limited to assessing the initial and
boundary conditions and key
assumptions used in the analysis,
treatment of uncertainties, and the use
of screening tools and bounding or
simplified methods for any mode or
hazard, provided the use of those tools
and methods is justified by an
acceptable technical basis. In that
regard, the consensus PRA standards
would not be applied by the NRC as a
strict checklist of requirements for part
53 PRA acceptability determinations.
The proposed § 53.450(c) would
require periodic maintenance and
upgrading of the PRA to maintain an
alignment between the supporting
analyses and the design and
performance of plant equipment,
programs and procedures, and other
factors associated with meeting the
safety criteria of the proposed § 53.220
and the evaluation criteria of proposed
§ 53.450(e)(2). The periodic
maintenance of the PRA would also be
a means to consider new or revised
information related to external hazards,
industry operating experience,
performance issues with or degradation
of SSCs, and other contributors to the
frequency and potential consequences
of various event sequences. The
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periodic assessments performed by
licensees to support the maintenance of
the PRA and other requirements in the
proposed part 53 would be
complemented by NRC inspections and
programs to assess new or revised
information related to topics such as
natural hazards, operating experience,
and potential generic safety issues.
The categories of LBEs used in part 53
would include anticipated event
sequences, unlikely event sequences,
and very unlikely event sequences. The
unlikely event sequences would include
those events with estimated frequencies
well below the frequency of events
expected to occur during the lifetime of
a commercial nuclear plant. An
important aspect of the analysis
requirements is that, under proposed
§ 53.450(e), the analyses of LBEs other
than DBAs would not only be used to
show the performance criteria of
§ 53.220 are satisfied but to also show
that evaluation criteria defined for each
LBE or category of LBEs would also be
satisfied. Such evaluation criteria for
specific LBEs or categories of LBEs
would be defined in terms of limits on
the release of radionuclides or
maintaining the integrity of one or more
barriers used to limit the release of
radionuclides and reflect a graded
approach of allowing lesser potential
consequences from more frequent
events. An example of such evaluation
criteria for a range of LBEs that could
likely be expanded for part 53 is
provided in RG 1.233. Another
proposed requirement for the proposed
§ 53.450(e) analyses is that the
methodology would need to include a
means to identify event sequences
deemed risk-significant such that those
event sequences can be given special
attention within other sections of part
53.
Part 53 would maintain an important
role for a deterministic analysis of DBAs
in the performance criteria of § 53.210
and the related analytical requirements
in § 53.450(f). The analysis of DBAs
would be required to address event
sequences drawn from those with
estimated frequencies below the
expected lifetime of a generation of
reactors (e.g., event sequences with
frequencies as low as one in ten
thousand years). As proposed in this
section, DBAs would need to be
analyzed using deterministic methods
and ensure a safe, stable end state with
reliance upon only SR SSCs and human
actions, if needed, to be performed by
operators licensed under the provisions
of §§ 53.760 through 53.795.
While the DBAs analyzed under part
53 would be similar to the traditional
DBAs analyzed under parts 50 and 52,
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there are important distinctions between
the overall role of DBA analyses in part
50 and proposed part 53. In part 53, the
role of the DBA analysis would be more
narrowly focused on selecting SR SSCs
and determining functional design
criteria for those SSCs to ensure the
commercial nuclear plant meets the
safety criteria in § 53.210. The overall
control of risks posed by commercial
nuclear plants under part 53 would be
provided by the analyses of and
measures taken for both DBAs and other
LBEs, including very unlikely event
sequences. This would contrast with the
traditional deterministic approach in
part 50 wherein the analyses of DBEs
such as DBAs were used to provide
bounding assessments, incorporate
standard design rules such as
assumptions related to single failures,
and to define conservative performance
requirements for SR SSCs. Limitations
related to the traditional deterministic
approach were addressed in part 50
through case-by-case assessments and
specific actions for beyond-design-basis
events such as anticipated transients
without scram and station blackout.
Section 53.450 would also include
provisions to ensure that analyses are
performed to support the design
requirements of § 53.440(e) on fire
protection, § 53.440(j) on aircraft impact
assessments, and § 53.425 on using
design features and plant programs to
control doses to members of the public
from routine effluents and direct
radiation from contained sources. The
proposed analysis requirements related
to fire protection would support either
a traditional, deterministic approach or
a more risk-informed approach where
the risks from fires are addressed within
the identification and analyses of LBEs.
Section 53.460 would establish
criteria for the safety classification of
SSCs and determination of appropriate
special treatments. As noted in subpart
A, the term ‘‘Special treatments’’ would
be defined to mean those items, such as
measures taken to satisfy functional
design criteria, quality assurance, and
programmatic controls, which provide
assurance that certain SSCs will provide
defense in depth or perform risksignificant functions. These
requirements would also provide
confidence that the SSCs will perform
under the service conditions and with
the reliability credited in the analysis
performed in accordance with § 53.450
to satisfy the safety criteria in §§ 53.210
and 53.220. The terminology used in
part 53 would include the following
categories for SSC classification: (1) SR;
(2) NSRSS; and (3) non-safety
significant. Requirements for SR SSCs
would be defined in other sections of
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part 53 and would include using TSs for
controls during operation and the
application of quality assurance
requirements from appendix B of part
50.
Requirements for NSRSS SSCs would
include the need to identify necessary
special treatments such as performance
measures on reliability. Licensees
would generally be afforded flexibility
in maintaining and changing special
treatments for SSCs categorized as
NSRSS. Non-safety-significant SSCs
would be addressed under normal
licensee programs for commercial grade
equipment and typical industry
practices for general plant design and
maintenance. Safety-related SSCs would
also contribute to defense in depth and
risk-significant functions and may
warrant special treatments beyond those
defined for their SR functions to reflect
their role in meeting the safety criteria
in § 53.220 and the evaluation criteria in
§ 53.450(e).
Section 53.470 would allow an
applicant or licensee to seek operational
flexibilities by adopting more restrictive
criteria than those provided in § 53.220
and that might otherwise be used in the
analysis of LBEs under § 53.450(e). Such
an approach might be taken to ensure
sufficient safety margins to gain
operational flexibilities in areas such as
justifying siting in relation to
population centers or staffing levels. As
an example, an applicant or licensee
could propose to justify siting proposals
by adopting alternate criteria for very
unlikely event sequences. Such
alternate criteria could require
calculated consequences for an
individual at the exclusion area
boundary to be less than one rem total
effective dose equivalent (TEDE). This
section would establish requirements to
ensure that, if more restrictive
evaluation criteria than those required
by a methodology were used to justify
operational flexibilities, then the
analysis, design features, and
programmatic controls would be
established and maintained accordingly.
Section 53.480 would establish
seismic design considerations. This
proposed section would relate to the
safety criteria in subpart B, the
analytical requirements related to
external hazards in § 53.450, and
subpart D, ‘‘Siting Requirements.’’ For
licenses issued under part 53, this
section in subpart C would support a
variety of approaches to seismic design.
For example, a design for a commercial
nuclear plant could show that SSCs are
able to withstand the effects of
earthquakes by adopting an approach
similar to that in appendix S to part 50.
Alternatively, an applicant could follow
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the more recent risk-informed
alternatives afforded by standards
development organizations (e.g.,
American Society of Civil Engineers
(ASCE)/Structural Engineering Institute
(SEI) 43–19, ‘‘Seismic Design Criteria for
Structures, Systems, and Components in
Nuclear Facilities.’’) Because the agency
has not endorsed ASCE/SEI–43–19, an
applicant can propose to use ASCE/SEI
43–19 on an application specific basis to
meet § 53.480 and the NRC would
evaluate the adequacy of the standard as
applied in that application. The design
could also be done with the full
integration of seismic PRAs into the
design and licensing of a particular
commercial nuclear plant. This section
has been developed to accommodate a
variety of potential risk-informed,
performance-based seismic design
approaches. The analyses required by
§ 53.450 would need to address seismic
hazards as well as other external
hazards. The expected responses of
SSCs to a range of seismic events would
be included in the analyses when
ensuring that the safety criteria defined
under § 53.220 would be met. The
potential SSC responses to seismic
hazards could be addressed in the
analyses using a fragility model
(conditional probability of its failure at
a given hazard input level), a high
confidence of low probability of failure
value, or other method endorsed or
otherwise found acceptable by the NRC.
Subpart D—Siting Requirements
Proposed subpart D in part 53 would
state requirements for the siting of
commercial nuclear plants and would
serve the role provided by 10 CFR part
100, ‘‘Reactor Site Criteria,’’ for nuclear
reactors licensed under parts 50 and 52.
As reflected in proposed § 53.500, the
reason for establishing siting
requirements would remain the same as
it has been historically, which is to
ensure that licensees and applicants
assess what impact the site environs
may have on a commercial nuclear plant
(e.g., external hazards) and, conversely,
what potential adverse health and safety
impacts a commercial nuclear plant may
have on nearby populations in view of
the site characteristics.
Proposed § 53.510 would require that
design-basis external hazard levels be
identified and characterized based on
site-specific assessments of natural and
constructed hazards with the potential
to adversely affect plant functions. The
site-specific assessments would be used
in the proposed § 53.415, which would
require that SR SSCs be designed to
withstand the effects of natural
phenomena and constructed hazards of
levels or severities up to design-basis
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external hazard levels. The design-basis
levels for external hazards relevant to a
site would need to account for
uncertainties and variabilities in data,
models, and methods used to
characterize those hazards. Existing
approaches could be used to
demonstrate compliance with this
requirement. The historical importance
of assessing seismic events as risks to
commercial nuclear plants and the
associated development of riskinformed approaches to address seismic
events would be reflected in proposed
§ 53.480, ‘‘Earthquake engineering,’’ and
specific requirements in subpart C. The
NRC is developing a graded approach
for seismic design by grouping SSCs
into different seismic design categories
(SDCs) based on their risk significance.
While the agency has not endorsed
ASCE/SEI–43–19, an applicant can
propose to use ASCE/SEI 43–19 on an
application-specific basis to meet
§ 53.480 and the NRC will evaluate the
adequacy of the standard as applied in
that application. The NRC staff will
continue to review ASCE/SEI–43–19 as
part of its efforts to further develop
guidance in this area. The approach
described in RG 1.208, ‘‘A PerformanceBased Approach to Define the SiteSpecific Earthquake Ground Motion,’’
would be an acceptable way to develop
site-specific ground motion response
spectra for SSCs under appendix S to
part 50, which corresponds to SSCs that
are categorized as the highest SDC
(SDC–5) in ASCE/SEI 43–19.
The evaluation of seismic hazards
under subpart D would need to be
sufficient to inform a site-specific
design (e.g., a CP or custom COL) or
confirm the use of a standard design for
a commercial nuclear plant under
§ 53.480 and other sections of subpart C.
A risk-informed approach could use
several design-basis ground motions
(DBGMs) to assess SSCs in various SDCs
(i.e., one DBGM per SDC). Section
53.510(d) would state that geologic and
seismic siting factors must also include
related hazards such as seismically
induced flooding and volcanic activity
that may affect the design and operation
of a proposed commercial nuclear plant
for the proposed site.
Section 53.520 would require
applicants to identify and assess site
characteristics related to topics which
might include meteorology, geology,
hydrology, or other areas in the design
and analyses required under subpart C.
Proposed section 53.530 would set
requirements for population-related
considerations and maintain
requirements and definitions similar to
those currently in part 100 for an
exclusion area, low population zone,
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and population center distance. The
NRC recognizes that some applicants
may propose to essentially collapse the
exclusion area and low population zone
to the site boundary. This approach
would rest on a demonstration that the
calculated consequences of DBAs
remain below the proposed dose
guidelines used in § 53.210, which are
the same as those in the existing
regulations in parts 50, 52, and 100. The
proposed definitions in § 53.020 would
allow such configurations, assuming
they were justified by the design and
analyses from subpart C. This approach
should provide flexibility to justify
alternative exclusion areas and low
population zones without foreclosing
the option for an applicant to define
more conventional exclusion areas and
low population zones outside of a
defined site boundary. The NRC’s longstanding preference for siting reactors in
areas of low population density would
be maintained in part 53 by using the
current language from part 100 in
proposed § 53.530(c). The NRC revised
guidance related to population densities
surrounding a commercial nuclear plant
in Revision 4 to RG 4.7, ‘‘General Site
Suitability Criteria for Nuclear Power
Stations’’ to reflect Commission
direction in SRM–SECY–20–0045,
‘‘Population Related Siting
Considerations for Advanced Reactors.’’
Site-related requirements in part 20
(restricted area) and part 73 (protected
and owner-controlled areas) would
remain applicable to commercial
nuclear plants licensed under part 53.
Proposed section 53.540 would
require that site characteristics be
appropriately considered in other
activities such as the design and
analysis performed under proposed
subpart D and the emergency planning
and security programs under proposed
subpart F.
Subpart E—Construction and
Manufacturing Requirements
The proposed part 53 language would
establish construction and
manufacturing requirements in subpart
E. The proposed language for
construction-related activities would
largely reflect current requirements in
part 50 without any fundamental
changes. Limited changes would be
made in several places, as described in
the following paragraphs, to be
technology-neutral and for consistency
with the organization and language of
part 53. The proposed language for
requirements for manufacturing
activities would largely mirror those for
construction-related activities. However,
the proposed manufacturing
requirements have been updated from
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the current requirements in subpart F of
part 52 to better accommodate the
possible factory fabrication of
manufactured reactors. The
manufacturing of specific components
outside the scope of an ML would not
be addressed by these proposed
subparts.
Section 53.600 would establish the
overall construction and manufacturing
requirements for CPs, OLs, COLs, MLs,
and limited work authorizations
(LWAs). This section would connect the
construction and manufacturing
requirements to the safety criteria,
quality assurance requirements, and
other requirements located in other
subparts. These requirements would
require that construction and
manufacturing activities be managed
and conducted such that when
combined with associated design
features and programmatic controls, the
constructed plant would satisfy the
relevant requirements in subpart B.
Section 53.605 would establish
requirements for the reporting of defects
and instances of noncompliance during
construction. This section would
provide equivalent requirements to
those in § 50.55(e).
Section 53.610(a) would establish the
requirement to have in place a welldefined command and control structure
to manage construction activities. The
requirements would generally reflect
current requirements, with an emphasis
on the quality assurance programs for
complying with the requirements in
appendix B to part 50. The proposed
§ 53.610(a)(6) would require
programmatic controls for implementing
special treatment for NSRSS SSCs to
align with requirements in other
subparts in part 53. The section would
also refer to other NRC regulations to
address matters such as requirements to
have a FFD program, a radiation
protection program if radioactive
materials are brought onto the site, and
security programs to protect sensitive
information and protect against cyber
threats.
Section 53.610(b) would provide
requirements governing construction
activities, including the equivalent of
the requirement in § 50.10(e) that
prohibits starting construction until the
NRC has authorized the activities by
issuing a CP, COL, ESP, or LWA.
Section 53.610(b)(1)(iii) would require
procedures to be in place prior to
beginning construction to ensure that
construction-related activities do not
undermine important features such as
slope stability and that constructionrelated activities such as backfilling of
excavated portions of the site
appropriately address potential pre-
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construction activities such as the
emplacement of retaining walls or
drainage systems. Other requirements in
these paragraphs would be equivalent to
requirements in parts 50 and 52 with
appropriate references to other parts for
items such as possession of byproduct
material or SNM, protecting operating
units from construction activities for
commercial nuclear plants with
multiple reactor units, and having a
redress plan in case LWA activities are
terminated.
Section 53.610(c) would address
inspection and acceptance activities by
including requirements in part 53
equivalent to specific quality assurance
criteria in appendix B to part 50 and
inspections, tests, analyses, and
acceptance criteria (ITAAC) in part 52
for COLs.
Section 53.620(a) would include
proposed requirements covering the
activities performed under an ML issued
under part 53. Provisions related to MLs
were first adopted by the NRC in 1973
through the addition of appendix M to
part 50. The regulation supported the
manufacture of a nuclear power reactor
to be incorporated into a commercial
nuclear plant under a CP and operated
under an OL at a different location from
the place of manufacture.1 The
regulations and processes for MLs were
changed substantially in the part 52
rulemaking in 2007 (72 FR 49352). The
most important shift in the ML concept
in that rulemaking was that a final
reactor design, which would be
equivalent to that required for a
standard DC under part 52 or an OL
under part 50, must be submitted and
approved before issuance of an ML. The
rationale for that change was that
approval of a final design ensures early
consideration and resolution of
technical matters before there is any
substantial commitment of resources
associated with the actual manufacture
of the reactor, which greatly enhances
regulatory stability and predictability.
The proposed part 53 sections in
subpart E for manufacturing and in
subpart H for licensing matters would
maintain requirements equivalent to
those in part 52 for MLs. The NRC
approval of a standard design and
related manufacturing processes,
coupled with a stable workforce and
established procedures, has the
potential for maintaining and even
improving the quality and consistency
of manufacturing, as compared to the
traditional method of constructing
1 On December 17, 1982, the NRC issued
‘‘Manufacturing License ML–1 to Offshore Power
Systems for the manufacture of a maximum of eight
floating nuclear plants,’’ dated September 30, 1982,
but the project was subsequently canceled.
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reactors onsite by a variety of
contractors and subcontractors.
Subpart E would include
requirements that would apply to
portions of a manufactured reactor in
recognition that some activities covered
by an ML may occur at different
fabrication facilities. As with the
preceding sections on construction,
§ 53.620 would establish the
requirements to have in place programs,
procedures, and a well-defined
command and control structure to
manage manufacturing-related
activities.
Section 53.620(b) in subpart E would
propose requirements for executing the
manufacturing activities following
receipt of an ML under part 53.
Information about the design and
manufacturing processes should be
provided by the applicant. The
importance of the ML is reflected in
several of the proposed requirements in
§ 53.620(b) that would refer to
complying with the ML, including
conducting manufacturing processes
within facilities for which the license
holder can control activities. The
essential role of post-manufacturing
inspections would also be incorporated
into this proposed section by requiring
the holder of the ML to perform
inspections and have acceptance
processes for manufactured reactors or
portions of a manufactured reactor.
Section 53.620(c) would provide
proposed requirements for the control of
radioactive materials if the holder of an
ML plans to possess and use source,
byproduct, or SNM as part of the
manufacturing process. By and large,
the proposed subpart E would refer to
NRC regulations in 10 CFR part 30,
‘‘Rules of General Applicability to
Domestic Licensing of Byproduct
Material,’’ 10 CFR part 40, ‘‘Domestic
Licensing of Source Material,’’ and part
70 for the requirements on controlling
radioactive materials. Several specific
requirements to address the potential
hazards of radioactive materials are
proposed in areas such as having a fire
protection program, an emergency plan,
training programs, and procedures to
minimize contamination.
The most significant change proposed
for MLs in part 53 as compared to MLs
under part 52 relates to § 53.620(d) in
subpart E and the associated licensing
provisions in subpart H. These
provisions would allow and establish
requirements for the loading of fuel into
a manufactured reactor at the
manufacturing site for subsequent
transport to a commercial nuclear
facility that will operate pursuant to a
COL. The first requirement in the
proposed § 53.620(d) would establish
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limitations on when a license under part
70 would authorize the loading of fuel
into a reactor manufactured under an
ML. The proposed regulation would
require the manufactured reactor to
include at least two independent
physical mechanisms that will each
prevent criticality should conditions
most favorable to critical operation be
introduced (e.g., optimum neutron
moderation and reflection). This
requirement would contribute to the
NRC’s longstanding practice of requiring
defense in depth for preventing
accidents in any facility dealing with
SNM, including requirements in § 70.64
for certain part 70 licensees to adhere to
the ‘‘double contingency principle.’’
The requirements to have in place
mechanisms to prevent criticality could
likewise support meeting other
provisions in subpart H to part 70, such
as those related to having a safety
program and integrated safety
assessment. The mechanisms to
preclude criticality in the proposed
requirements would reasonably ensure
that a manufactured reactor would not
become critical assuming optimum
neutron moderation, and optimum
neutron reflection conditions. With the
proposed requirements for mechanisms
to prevent criticality and all criticality
safety controls required by 10 CFR part
70 in place, the presence of fuel in the
manufactured reactor would not create
a nuclear hazard different than the
hazard from the presence of the same
fuel in a storage location or container
licensed under 10 CFR part 70.
Collectively, the proposed measures
would reasonably ensure that the
manufactured reactor would not be
capable of operations, thereby obviating
the need for a COL under §§ 53.1416
and 53.1440 to authorize fuel loading.
Additionally, this approach would focus
the ML application and its review on
the design, manufacture, and
deployment of the manufactured
reactor.
The activities involving SNM within
the manufacturing facility, including the
loading of fuel, would be regulated
primarily under the part 70 license. The
reference to the requirements in subpart
H of part 70 in section 53.620(d) assures
that the activities involving the receipt,
storage, and loading of a variety of
possible fuel forms and enrichments at
the manufacturing facility will be
analyzed in a systematic manner and
appropriate protection will be provided
against equipment malfunctions, human
errors, external hazards, and other
adverse conditions. The regulations in
part 51 provide a flexible approach for
environmental review to address the
range of regulated activities under part
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70. The flexibility in part 51 will enable
the NRC to determine the appropriate
type of environmental review based on
the circumstances associated with the
loading of fuel into a specific
manufactured reactor.
The proposed § 53.620(d) cites the
requirements in parts 70, 71, and 73 to
ensure important features and programs
are in place prior to the receipt of SNM.
The features and programs required to
be in place prior to receipt of SNM
include (1) radiation monitoring
instrumentation and alarms; (2)
measures to detect potential criticality
accidents; (3) appropriate procedures,
equipment, and personnel qualified for
the fuel loading; (4) programs for
physical security and cybersecurity; and
(5) material control and accounting
(MC&A) programs. Section
53.620(d)(2)(i) proposes requirements to
address security programs for any ML
authorizing possession of a
manufactured reactor into which fuel
has been loaded at the manufacturing
facility. Currently, for category II SNM,
security measures may be required in
addition to requirements included in
§ 73.67, ‘‘Licensee fixed site and intransit requirements for the physical
protection of special nuclear material of
moderate and low strategic
significance,’’ on a case-by-case basis.
Including appropriate security measures
in the proposed part 53 regulations will
provide additional openness and
transparency for applicants applying for
an ML who seek to load fuel into
manufactured reactors at a
manufacturing site.
Currently, § 73.67 only requires a
security plan for licensees who possess,
use, transport, or deliver to a carrier for
transport SNM of moderate strategic
significance, or 10 kg or more of SNM
of low strategic significance. However,
the proposed physical security program
for fueled manufactured reactors would
require a security plan for any ML
authorizing possession of a
manufactured reactor into which fuel
has been loaded at the manufacturing
facility, regardless of fuel type,
enrichment, and quantity. This is
consistent with other controls for MLs,
including reactivity and criticality
controls.
The proposed requirements would
also require a holder of an ML and part
70 license to address cybersecurity to
ensure a cyberattack would not
adversely impact the functions
performed by digital assets used by the
licensee for physical security, radiation
monitoring, or criticality prevention.
The proposed regulations in part 53
covering the activities related to the
storage, movement, and loading of fresh
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fuel into a manufactured reactor in the
manufacturing facility would likewise
refer to the applicable regulations in
part 70. The proposed § 53.620(d) would
also require the loading or unloading of
unirradiated fuel into or from a
manufactured reactor and any changes
to the configuration of reactivity-related
systems to be performed by a certified
fuel handler meeting the requirements
in subpart F. The NRC is aware of
proposals to introduce reprocessing of
existing or future spent nuclear fuel into
the fuel cycle for some potential
commercial nuclear plants. This
proposed rule does not address the
loading of spent nuclear fuel or fuel
resulting from reprocessing of spent
nuclear fuel into a manufactured
reactor.
Section 53.620(e) would limit the
transport and delivery of a
manufactured reactor or portions of a
manufactured reactor only to a site for
which the Commission has issued a
COL authorizing the construction of a
commercial nuclear plant using a
manufactured reactor under the specific
ML. This proposed requirement is
similar to the limitations in § 52.153,
with the difference being that part 53
would allow the installation of a
manufactured reactor at the site of a
COL but would not include provisions
for installation at a site under a CP. The
possible combination of a manufactured
reactor and the licensing option of CP
and OL seems unlikely and would
require the introduction of ITAAC into
the licensing provisions for a CP and
OL. An additional proposed paragraph
in § 53.620(e) would provide
requirements for protecting fueled
manufactured reactors during transport
to the site of the commercial nuclear
plant by referencing the transportation
and security requirements in 10 CFR
part 71, ‘‘Packaging and Transportation
of Radioactive Material,’’ and part 73.
Section 53.620(f) would include
proposed requirements for the
acceptance and installation of a
manufactured reactor at the site of a
commercial nuclear plant. The proposed
requirements would reference the
construction requirements in § 53.610 to
govern the integration of the
manufactured reactor into the
construction of a commercial nuclear
plant. Other proposed requirements in
the section would address required
receipt inspections and verification that
interface requirements between the
manufactured reactor and the balance of
the commercial nuclear plant have been
met.
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Subpart F—Requirements for Operation
Proposed subpart F would provide the
requirements for the operations phase of
a commercial nuclear plant to ensure
that the safety criteria in subpart B are
satisfied throughout the plant’s lifetime
and during all modes of normal
operation and unplanned events.
Section 53.700 would provide the
overall objectives and general
organization of subpart F, which would
be to establish requirements during
operations for: (1) plant SSCs; (2) plant
personnel; and (3) plant programs.
Proposed § 53.710 would provide the
requirements for maintaining
capabilities, availability, and reliability
of SSCs to demonstrate compliance with
the safety criteria and design
requirements for unplanned events that
are described in proposed subparts B
and C. The basic structure of this
proposed section would be that controls
for SR SSCs are provided by TS and
controls for NSRSS SSCs are required to
be addressed with licensee-controlled
documents and procedures.
The general content and control of TS
under the proposed part 53 would be
similar to the requirements in part 50.
The proposed requirements for TS
would include limits on the inventories
of radioactive materials, plant operating
limits, and specific requirements for
each SR SSC, including limiting
conditions for operation (LCO) and
required surveillances. The proposed
requirements for TS would also include
a section on important design elements,
which is similar to design features in
§ 50.36, and a section for administrative
controls. A provision addressing the
development and submittal of TS to
address decommissioning activities
would also be included in the proposed
subpart G.
The proposed requirements for TS
under part 53 would not carry over
safety limits or associated limiting
safety system settings from § 50.36,
which contains TS requirements for
operating reactors under parts 50 and
52. As discussed in SECY–18–0096,
systematic assessments and more
mechanistic approaches to evaluating
source terms support an alternative
approach to establishing barrier-based
safety limits. An example provided in
that paper is a comparison of: (1) the
traditional specified acceptable fuel
design limits (SAFDL) that support
protecting a specific barrier from
potential failure mechanisms (e.g.,
departure from nucleate boiling to
protect fuel cladding); and (2) the
specified acceptable system
radionuclide release design limit
(SARRDL) concept, which limits the
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possible increase in circulating
radionuclide inventory during normal
operations or an AOO as part of an
integrated or ‘‘functional containment’’
approach. Additional discussion of the
use of SARRDL in the design and
licensing of advanced reactors is
provided in RG 1.232. The SARRDL
could be addressed as an operating limit
within this proposed construct of
requirements for TS. In cases, such as
LWRs, where a SAFDL approach might
be used as part of a mechanistic
approach to meeting the design and
analysis requirements in subpart C, the
associated functional design criteria
proposed in § 53.410 and TS under the
proposed § 53.710(a) would define
similar requirements as those provided
by the safety limit and limiting safety
system setting requirements in § 50.36.
The proposed requirements for TS
under part 53 would not include
specific criteria for identifying when
LCOs must be established (i.e., would
not include an equivalent to
§ 50.36(c)(2)(ii)). Instead, consistent
with subparts B and C, the TS
requirements in subpart F of part 53
would define TS LCOs as providing
limits on SR SSCs. The SR SSCs protect
against DBAs to demonstrate
compliance with the safety criteria in
the proposed § 53.210. In the proposed
construct for part 53, risk-significant
SSCs would be addressed through a
combination of TS for the SR SSCs and
establishment and monitoring of
performance standards for NSRSS SSCs.
In addition to addressing TS for SR
SSCs, proposed § 53.710 would require
appropriate controls be developed and
implemented for NSRSS SSCs.
Examples include appropriate
surveillances and controls established
through reliability assurance programs.
Configuration management and other
special treatments would provide that
the capabilities, availabilities, and
reliabilities of NSRSS SSCs are
maintained consistent with the
underlying risk assessments while
providing flexibility to licensees
through maintaining the management
functions within licensee-controlled
programs. Controls on NSRSS SSCs are
appropriate as part of the overall
performance-based approach within
proposed part 53. Special treatments
beyond those defined for their SR
functions may also be warranted for SR
SSCs to reflect their role in meeting the
safety criteria in § 53.220 and the
evaluation criteria in § 53.450(e). The
performance objectives for NSRSS SSCs
would reflect that the comprehensive
risk metrics and related risk
performance objectives established
under § 53.220 may involve assessing
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and averaging the risks over a defined
period (e.g., plant year) and would not
constitute a real-time requirement that
must be continuously demonstrated by
the licensee. The controls under
§ 53.710(b) justify proposed changes in
part 53 from the traditional or
deterministic approaches in parts 50
and 52 in areas such as replacing the
single-failure criterion with a
probabilistic reliability criterion (see
SRM–SECY–03–0047, ‘‘Policy Issues
Related to Licensing Non-Light-Water
Reactor Designs,’’ dated June 26, 2003).
This approach could also support the
incorporation of risk insights and
analytical margins to gain operational
flexibilities in areas such as siting and
staffing requirements described in
subsequent sections of proposed subpart
F.
Proposed § 53.715 would provide the
requirements for developing and
implementing a program to do the
following: (1) control maintenance
activities; (2) take appropriate corrective
action when performance issues are
identified; (3) conduct routine
evaluations of effectiveness; and (4)
assess and manage risks resulting from
maintenance activities. These proposed
requirements are similar to those
included in § 50.65 (maintenance rule),
including the need to assess and manage
the increase in risk that may result from
the proposed maintenance activities.
While, for the maintenance rule,
specific criteria must be developed to
capture both SR and non-SR but
otherwise important SSCs, the proposed
§ 53.715 would cover SR SSCs and
NSRSS consistent with other subparts in
part 53.
Proposed § 53.720 would provide the
requirements for responding to a
seismic event during the operating
phase of the life cycle of a commercial
nuclear plant and would be equivalent
to the requirements in paragraph
IV(a)(3) of appendix S, ‘‘Earthquake
Engineering Criteria for Nuclear Power
Plants,’’ to part 50.
The proposed part 53 would include
provisions to address staffing, training,
personnel qualifications, and human
factors engineering (HFE) in a manner
that is risk informed, technology
inclusive, performance based, and
flexible in nature. During the
development of part 53, the staff
prepared a draft white paper on ‘‘Risk
Informed and Performance Based
Human-System Considerations for
Advanced Reactors,’’ to support
interactions with stakeholders and the
ACRS. Key considerations include the
recognition that staffing, operator
qualifications, and HFE are
interconnected areas that must be
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approached in an integrated manner
and, furthermore, that safety functions,
including the means by which they are
fulfilled, provide an effective method
for informing technology-inclusive
requirements.
The requirements associated with this
approach would be in §§ 53.725 through
53.830. Section 53.725 discusses
applicability and defines specific terms.
Some definitions draw from those in
§ 55.4. Several new definitions would be
introduced for use within the context of
subpart F. These new definitions would
be the following: ‘‘Automation,’’
‘‘Auxiliary operator,’’ ‘‘Generally
licensed reactor operator,’’ ‘‘Interactiondependent-mitigation facility,’’ ‘‘Load
following,’’ ‘‘Self-reliant-mitigation
facility.’’
Sections 53.725 through 53.830 would
be divided into four portions that would
cover general operational requirements,
operator and senior operator licensing
requirements, generally licensed reactor
operator (GLRO) requirements, and
general training requirements for plant
staff. The NRC intends to provide
guidance addressing the review of
operator staffing plans; the review of
operator, senior operator, and GLRO
examination programs; and the
implementation of scalable HFE
reviews. Licensees would be required to
use GLROs upon demonstrating
compliance with the criteria in § 53.800.
Certain routine communications are
necessary to facilitate the operator
licensing process. The NRC is proposing
to adapt the requirements of §§ 55.5 and
50.74 to § 53.726 to accomplish this.
Specific information must be
collected in order to facilitate the initial
issuance of operator licenses, as well as
to allow for license renewals and
required updates thereafter. Such
information collection activities must
also be approved by the OMB. The NRC
is proposing to adapt the requirements
of § 55.8, to include any needed updates
in OMB approval information, to
§ 53.120 to accomplish this.
The information used within the
regulatory processes of the NRC must be
free from omissions and inaccuracies to
facilitate effective regulation. Consistent
with this, the NRC is proposing to adapt
the requirements of § 55.9 to § 53.728 to
require the completeness and accuracy
of material information provided by
individual applicants and license
holders.
Section 53.730 would provide
performance-based and technologyinclusive requirements for assessing the
role of personnel in facility safety,
applying human-system considerations
within facility design, and incorporating
operational approaches that are
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consistent with design-specific safety
considerations. Most of these
requirements would be adapted from
portions of §§ 50.34(f) and 50.54 and 10
CFR part 55, ‘‘Operators’ Licenses,’’
with considerable modification in order
to reflect the introduction of new
technologies and possible changes in
the roles of personnel in preventing and
mitigating events. The NRC is proposing
that these technical requirements
would, together, serve as a component
of the required content of applications
for OLs and COLs under part 53.
Additionally, the NRC proposes that the
specific technical requirements
associated with HFE, human-system
interface design, concept of operations,
functional requirements analysis, and
function allocation would serve as a
component of the required content of
applications for standard DCs, standard
design approvals, MLs, and CPs, as well.
Human factors engineering is
essential to facilitate the role of
personnel in facility safety in a manner
that is both effective and reliable. The
NRC proposes to adapt § 53.730(a) from
the HFE design requirements of
§ 50.34(f)(2)(iii). A key difference would
be that the requirement would now be
focused on settings where personnel
fulfill their safety or emergency
response roles wherever they may
occur. The NRC additionally proposes
to include within the scope of this
requirement activities for assuring the
continued availability of plant
equipment that is needed for safety, and
envisions that this may encompass
relevant maintenance, inspections, and
testing as well. The NRC intends that
this requirement would be associated
with staff guidance for conducting
scalable reviews of HFE that is planned
to accompany part 53.
Human-system interfaces provide
vital information to operators across a
spectrum of operating conditions that
can range from normal operations
through severe accident conditions. The
specific types of information that must
be available to support operations staff
during such conditions include, in part,
those associated with safety function
parameters, safety system status,
possible core damage states, barrier
integrity, and radioactive leakage. Due
to the importance of such information,
the NRC proposes under § 53.730(b) to
require such human-system interface
design features for all facilities,
irrespective of other flexibilities
proposed under part 53. Therefore, the
NRC proposes to adapt specific postThree Mile Island requirements of
§ 50.34(f) in a technology-inclusive
manner as detailed in the following:
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• Paragraph (b)(1) would be adapted
from § 50.34(f)(2)(iv).
• Paragraph (b)(2) would be adapted
from § 50.34(f)(2)(v).
• Paragraph (b)(3) would be adapted
from § 50.34(f)(2)(xi), 50.34(f)(2)(xii),
and 50.34(f)(2)(xxi).
• Paragraph (b)(4) would be adapted
from § 50.34(f)(2)(xvii), 50.34(f)(2)(xviii),
50.34(f)(2)(xix), and 50.34(f)(2)(xxiv).
• Paragraph (b)(5) would be adapted
from § 50.34(f)(2)(xxvi).
• Paragraph (b)(6) would be adapted
from § 50.34(f)(2)(xxvii).
In addition to the requirements of
§ 53.730(b)(1) through (6), a further set
of human-system interface design
requirements applicable only to those
facilities that will be staffed by GLROs
would be provided under § 53.730(b)(7).
This prescriptive set of design
requirements for those facilities which
demonstrate compliance with the
criteria of § 53.800 would recognize that
the application of HFE under § 53.730(a)
is anticipated to be significantly
reduced at such facilities in the absence
of an expected operator role for the
fulfillment of safety functions. However,
it should be noted that the capability for
an immediately initiated, manual
reactor shutdown would be
conservatively mandated irrespective of
any other design considerations.
The NRC proposes § 53.730(c) to
require the submittal of a concept of
operations that is of sufficient scope and
detail to appropriately inform the staff.
The development of a concept of
operations can facilitate a clear
understanding on the part of the NRC
for potential novel operating concepts.
Additionally, such information is likely
to reduce the degree of resources and
interactions needed for the NRC to
obtain the understanding necessary to
enable flexible requirements in areas
such as staffing, operator qualifications,
and HFE.
The NRC proposes § 53.730(d) to
require the submittal of both a
Functional Requirements Analysis and a
Function Allocation. The identification
of design-specific safety functions and
how they are fulfilled serves as a
primary means for achieving
technology-inclusive requirements
within areas such as staffing, operator
qualifications, and HFE. The Functional
Requirements Analysis and Function
Allocation processes (which are both
HFE methods derived from systems
engineering principles), provide an
effective means to identify both how
safety functions will be satisfied and
how to characterize any associated
operator role in doing so. A Functional
Requirements Analysis shows what
features, systems, and human actions
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are relied upon to demonstrate safety
(i.e., fulfill safety functions). A Function
Allocation then describes how safety
functions are assigned to both personnel
and automatic systems. However, an
important adaptation of the Function
Allocation for use under the proposed
rule would be the further need to not
only describe allocations of safety
functions to human action and
automation, but also to identify
allocations made to active safety
features, passive safety features, or
inherent safety characteristics as well.
Operating experience provides an
important source of information by
which to inform various aspects of
facility design and operations.
Accordingly, the NRC proposes in
§ 53.730(e) to adapt the requirements of
§ 50.34(f)(3)(i) for requiring an operating
experience program.
New technologies may involve
concepts of operations that are more
conducive to customizable licensed
operator staffing requirements than the
prescriptive requirements of § 50.54(m).
Analyses and assessments that are based
on HFE principles provide a
performance-based means of
determining licensed operator and
senior operator staffing needed to
support safe operations. In contrast, for
those facilities required to be staffed by
GLROs, the NRC anticipates that the
operator staffing plans will reflect a
simpler approach of showing that a
continuity of responsibility will be
maintained for facility operations
throughout the operating phase, with at
least one GLRO providing continuous
oversight and remaining immediately
available when any units are fueled.
Additionally, a revised approach to the
traditional position of the shift technical
advisor that focuses on the availability
of engineering expertise as a means of
addressing uncertainties and abnormal
circumstances is more suitable within
the context of part 53 and is intended
to be applicable to all facilities,
irrespective of other design and staffing
considerations.
Consistent with this approach, the
NRC proposes under § 53.730(f) to
require the submittal of a staffing plan
that details operations staffing, how
engineering expertise will be provided,
and what staffing will be available to
provide other needed support functions.
The NRC intends that this requirement
would be associated with staff guidance
for reviewing operations staffing plans
that is planned to accompany part 53
and that, following NRC approval of the
OL or COL, the staffing plan would
become a condition of the facility
license. The NRC intends that, at a
minimum, the approved licensed
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operator and senior operator (or, if
applicable, GLRO) staffing, positions,
and personnel locations will be
incorporated into corresponding
requirements within the facility TS and
that a license amendment would thus be
required for any subsequent changes.
Operator training and qualification
programs provide an essential
component of supporting human
performance in implementing tasks with
safety implications. Such programs
must include components that cover the
stages of initial training, examination,
and continuing training. Additionally,
recognizing the potential for varying
concepts of operations to affect
traditional, prescriptive approaches to
operator proficiency, the NRC proposes
under part 53 to allow facilities to
develop operator proficiency programs
based on facility-specific
considerations.
Therefore, the NRC proposes in
§ 53.730(g)(1) to require approval as part
of its approval of the OL or COL, of the
programs that will be used for the initial
training, initial examination,
requalification training and
examination, and proficiency of both
licensed operators and senior operators.
In a corresponding manner, the NRC
proposes in § 53.730(g)(2) to require
approval of the programs that will be
used for the GLRO equivalents of each
of these programs for facilities with
such staffing. The NRC intends that
examination program requirements
would be associated with staff guidance
for the review of tailored examination
processes that are planned to
accompany part 53. Following the
completion of an initial training
program, continuing training programs
provide an important means of
sustaining the knowledge and abilities
of individuals. The NRC is proposing to
adapt the requirements of § 50.54(i–1) in
§ 53.730(g)(3) to require that operator
continuing training programs be in
effect to support operator performance.
Under part 53, the NRC proposes to
require these programs to be in effect
concurrent with when the initial
operator examinations first commence,
in effect putting the programs in place
only when they are needed. This
represents a modification of the
comparable requirement of § 50.54(i–1),
which links the commencement of these
programs to a timeline driven by the
licensing of the facility.
The authorization to manipulate
controls of the facility that directly
affect reactivity or power level is
restricted to individuals who are either
licensed operators, licensed senior
operators, or GLROs. However, for
practical purposes, situations in which
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an individual is participating in an
approved training program or
reestablishing proficiency may also call
for them to operate the controls of the
facility under the cognizance of a
licensed individual. The NRC is
proposing to adapt the requirements of
§ 55.13 in § 53.735 to accomplish this,
with a notable difference being the
incorporation of GLROs.
Section 53.740 would provide
requirements for OL and COL holders
under part 53. Portions of § 53.740
would be adapted from the conditions
of § 50.54. In general, the conditions for
operations staffing under part 53 would
reflect considerations for potential
technological differences and varying
concepts of operation that are expected
among part 53 facility licensees.
Additionally, certain requirements
would be specific to the operating phase
while others would remain in effect
following the permanent cessation of
facility operations during the
decommissioning phase.
All commercial nuclear plants
licensed under part 53 would require
some form of licensed operator staffing,
whether it be by specifically or
generally licensed operators. Consistent
with this, the NRC is proposing under
§ 53.740(a) to require facility licensees
to demonstrate compliance with the
programmatic requirements for either
specifically licensed operators and
senior operators or for GLROs, as
applicable to the facility.
The NRC recognizes that technologyinclusive facility staffing will need to
account for a potentially wide range of
concepts of operations; for this reason,
flexible and performance-based
approaches for establishing required
facility staffing are appropriate.
However, once the appropriate facility
staffing has been determined and
approved by the NRC, such staffing
must be maintained to ensure that the
appropriately qualified individuals will
be available when needed to support the
safe operation of the facility. Therefore,
the NRC is proposing under § 53.740(b)
to require that the staffing described
within the approved facility staffing
plan be maintained as a condition of the
facility license as opposed to
prescriptive staffing requirements like
those of § 50.54(k) and (m).
Because operation of facility controls
directly affects reactivity or power level,
only those individuals who possess
appropriate levels of qualification and
authorization are permitted to operate
those controls. The NRC is proposing to
adapt the requirements of § 50.54(i) in
§ 53.740(c) to require that only
specifically licensed operators and
senior operators or, alternatively,
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GLROs, may operate facility controls,
with allowance for specified exceptions
for the purposes of operator training or
proficiency.
Senior operators, by virtue of their
license level, are qualified and
authorized both to perform certain
important responsibilities and to direct
the licensed activities of licensed
operators. Therefore, facilities that are
required to be staffed by specifically
licensed operators must also include
senior operators within their staffing. In
contrast, facilities staffed with GLROs
only have a single license level available
and, therefore, there is no equivalent
provision for such facilities. The NRC is
proposing to adapt the requirements of
§ 50.54(l) in § 53.740(d) to require the
licensing and designation of senior
operators at facilities staffed by
specifically licensed operators.
In contrast with control
manipulations that directly affect
reactor power and reactivity (e.g.,
control rod movement, control drum
rotation, recirculation pump speed
adjustment, reactor coolant system
boration or dilution, etc.) and are
therefore restricted to performance only
by licensed operators, other types of
plant operations that may result in
reactor power and reactivity changes via
means that are indirect in nature (e.g.,
electrical generation changes, turbine
bypass valve operation, steam usage by
process heat applications, etc.) may be
implemented by non-licensed
personnel. However, due to the
potential influence of such operations
on reactor power and reactivity, the
continuous oversight of reactor
parameters by a licensed operator is
necessary during these operations. The
NRC is therefore proposing to adapt the
requirements of § 50.54(j) in § 53.740(e)
to require appropriate oversight of
operations, other than those associated
with the controls themselves, that may
affect reactivity or power level.
Load following where plant output
automatically changes in response to
externally originated instructions or
signals is not permitted under the
existing regulations of § 50.54. However,
new technological considerations and
concepts of operation may justify such
an operational approach under
appropriate circumstances. The NRC
recognizes that, beyond electrical power
generation, load following may also
affect other applications of plant output,
such as hydrogen production,
desalination, or district heating. For
load following to be permissible,
measures must be in place to provide
assurance that plant output
considerations are not permitted to lead
to challenges to safe reactor operations.
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These measures may consist of
automated control systems, automatic
protective features, or the continuous
oversight and immediate intervention
capability of an appropriately qualified
and authorized individual. Section
53.740(f) would allow for load
following, provided that appropriate
measures are in place. In considering
the acceptability of the measures
associated with load following, the NRC
expects that any automatic protection
relied upon would be separate from that
credited for reactor protection purposes
and would employ setpoints that are set
so as to prevent actuation of the reactor
protection system while accomplishing
its functions to the extent practical.
Core alterations such as refueling are
associated with specific considerations
that warrant limiting the oversight of
such operations to appropriately
qualified and authorized individuals.
Unlike other types of fuel handling
operations, core alterations occur within
the confines of a reactor vessel that is
specifically designed to support and
sustain nuclear criticality, thereby
justifying the imposition of higher
qualification levels within such
contexts. The NRC is proposing to adapt
the requirements of § 50.54(m)(2)(iv) in
§ 53.740(g) to require the supervision of
core alterations by either a specifically
licensed senior operator, a specifically
licensed senior operator whose license
is limited to fuel handling, or by a
GLRO, as applicable to the facility.
Because certain commercial reactor
designs may be capable of refueling
while at power and, in any event,
overall facility oversight would already
be required by either a specifically
licensed senior operator or by a GLRO,
the NRC proposes to omit this
requirement as redundant during
periods where core alterations occur
while the plant is operating.
It is impossible to predict every
possible scenario that a commercial
nuclear plant might potentially
encounter. Therefore, it is prudent to
grant the authority for appropriately
qualified individuals to depart from
facility license conditions when
emergency circumstances dictate that
doing so is in the interest of public
health and safety. The NRC is proposing
to adapt the requirements of § 50.54(x)
and (y) in § 53.740(h) to permit specific
individuals to authorize departures from
facility license conditions or TSs when
emergency conditions warrant doing so
for the protection of the public health
and safety. Recognizing that certain
facilities licensed under part 53 may be
staffed by GLROs in lieu of specifically
licensed senior operators, the NRC
proposes to extend this authority to
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GLROs. While it is not anticipated that
GLROs will have a role in the
fulfillment of safety functions at selfreliant-mitigation facilities and,
furthermore, that operators at such
facilities would not be in a position by
which to significantly influence
radiological safety outcomes, the very
nature of the § 50.54(x) and (y) and the
proposed § 53.740(h) provisions concern
situations that are unanticipated and,
therefore, unforeseeable. Thus, it is
appropriate to grant GLROs a
comparable authority to that of senior
licensed operators and certified fuel
handlers as it relates to invoking this
provision under emergency conditions
as a means of accounting for such
possibilities.
Due to the unique authorities and
responsibilities of both specifically and
generally licensed reactor operators, it is
essential that any individual fulfilling
such a role demonstrate compliance
with the regulatory requirements for
operator licensing. Section 107 of the
Act authorizes the Commission to
prescribe conditions for the licensing of
operators and to issue licenses
consistent with those conditions. The
NRC is proposing to adapt the
requirements of § 55.3 in § 53.745 to
require that any person performing the
function of an operator, senior operator,
or GLRO must be authorized by a
license issued by the Commission.
The NRC proposes to license
individuals as operators under both
specific and general licensing
frameworks. Specific licenses would be
for licensed operators (i.e., reactor
operators) and senior operators (i.e.,
senior reactor operators) and would be
issued to a named person upon approval
by the Commission of an application for
that named person. In contrast, GLROs
would perform duties under the
provisions of a general license that
would be effective without the filing of
an application with the Commission or
the issuance of licensing documents to
a particular person. The NRC proposes
requirements for the use of a specific
licensing process for licensed operators
and senior operators under §§ 53.760
through 53.795, with § 53.760
addressing applicability.
Medical fitness is an important
component of the overall process of
specifically licensing operators because
it provides assurance that operators will
be able to carry out important duties
without being precluded from doing so
by health-related issues. Medical fitness
also provides assurance that such issues
will not adversely affect the
performance of assigned job duties or
cause operational errors that endanger
public health and safety. In addition to
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a requirement for medical fitness, a
medical examination by a physician to
confirm compliance with this
requirement is necessary. The NRC is
proposing to adapt the requirements of
§§ 55.21, 55.23, and 55.27 under
§ 53.765 to require medical fitness,
examinations by physicians, and
medical certification for specifically
licensed operators and senior operators.
In recognition of the fact that GLROs are
not expected to have a role in the
fulfillment of safety functions at the
facilities at which they are licensed, the
NRC proposes to not extend a
comparable medical requirement to
GLROs.
The NRC is also proposing to adapt
the requirements of §§ 55.25 and
50.74(c) in § 53.770 to require that
timely notifications be made to the NRC
if a specifically licensed operator or
senior operator develops a permanent
physical or mental condition that
adversely affects the performance of
assigned operator job duties or could
cause operational errors endangering
public health and safety.
Notwithstanding this requirement
related to permanent medical
conditions, the NRC continues to
recognize that it is appropriate for
facility licenses to impose
administrative restrictions and
conditions upon specifically licensed
operators and senior operators in
response to temporary medical
conditions.
The process of specifically licensing
individuals as licensed operators or
senior operators requires the submittal
of applications to the NRC for review.
These applications must detail certain
elements associated with licensing,
including the demonstration of
compliance with examination,
experience, and medical requirements.
The NRC is proposing to adapt the
requirements of §§ 55.31 through 55.35
in § 53.775 to include requirements for
the applications associated with the
specific licensing of licensed operators
and senior operators at commercial
nuclear plants licensed under part 53. In
contrast with the part 55 requirements,
the NRC proposes to provide additional
flexibility by locating certain details
associated with the preparation and
submittal of these applications within
guidance in lieu of placement within
this proposed rule itself.
The NRC proposes overall
programmatic requirements for
specifically licensed operator and senior
operator training, examination, and
proficiency in § 53.780. In general, the
proposed requirements are adapted from
those in part 55, with several additional
flexibilities being incorporated to better
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account for potential variations in
reactor technologies and concepts of
operations. The requirements proposed
in § 53.780 cover, in part, the initial
training, initial examination,
requalification training, requalification
examination, and proficiency of
specifically licensed operators and
senior operators.
The initial training process provides
individuals with the knowledge and
abilities needed to subsequently fulfill
assigned duties as licensed operators or
senior operators in a safe and reliable
manner. The use of a systems approach
to training (SAT) ensures that the
training program is based upon job
requirements in a manner that can be
adapted to account for differences in
plant technology, concepts of
operations, and operator roles in the
fulfillment of design-specific safety
functions. The NRC is proposing under
§ 53.780(a) to require facility licensees
to implement a SAT-based training
program for the initial training of
licensed operator and senior operator
applicants. The program must be
adequate to ensure that applicants will
be capable of performing the duties
necessary both to protect public health
and safety and to maintain plant safety
functions. The NRC further proposes
that such programs be subject to NRC
approval and subsequent change control
processes of an appropriate nature.
Examinations provide a means of
assessing that individuals have achieved
a degree of knowledge and ability that
is sufficient to carry out assigned duties
as licensed operators or senior operators
in a manner that is safe and reliable.
The NRC is proposing to adapt the
requirements of §§ 55.40, 55.41, 55.43,
and 55.45 in § 53.780(b) to require that
facilities establish and implement an
initial examination program. However, a
key difference from the comparable
requirements of part 55 would be that
facilities have the flexibility to propose,
subject to NRC approval, the
examination methods and criteria to be
used in assessing satisfactory applicant
performance. Such examination
programs (including those used within
the scope of requalification training)
would need to provide for acceptable
levels of both test validity and test
reliability in order to be considered
acceptable. The NRC intends that staff
guidance would be available to facilitate
the review of licensing examination
programs that are proposed by facility
licensees and that, following NRC
approval, initial examination programs
would be subject to an appropriate
change control process. Furthermore,
the NRC proposes that holders of
licenses to operate commercial nuclear
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plants under part 53 be provided the
alternative of administering their own
approved licensing examinations. The
NRC would continue to exercise
appropriate oversight of the program,
make operator licensing decisions based
upon the examination results, and
reserve the right to administer the
examinations in lieu of permitting the
facility to do so. However, irrespective
of the provided flexibilities in
examination format and structure, at a
minimum, topics from the following
general categories of knowledge and
abilities should be sampled in such
examinations:
• Reactor Theory, Thermodynamics,
and Chemical Interactions
• Plant Systems and Components
• Reactivity Management and
Manipulations
• Radiation Control and Safety
• Emergency, Abnormal, and Normal
Operations
• Administrative Requirements and
Conditions of the Facility License
Requalification training programs
provide for the continuing training and
examination of specifically licensed
operators and senior operators to ensure
that they maintain the knowledge and
abilities needed to support the safe and
reliable performance of job duties
following the completion of an initial
training and examination program. The
NRC is proposing to adapt the
requirements of § 55.59 in § 53.780(c) to
require that facilities implement both a
SAT-based requalification training
program and a biennial requalification
examination program. However, a
notable difference from the biennial
requalification examinations required
under part 55 would be that distinct
annual operating test and biennial
written examination components would
not be mandated, with the facility
licensee instead proposing the
examination methods and criteria to be
used in assessing satisfactory
performance. The NRC intends that
guidance would be available to facilitate
the review of the requalification
examination programs that are proposed
by facility licensees and that, following
NRC approval, requalification
examination programs would be subject
to an appropriate change control
process.
For examinations to provide for valid
assessments of the knowledge and
abilities of individuals, the
examinations must remain free from
compromises that could affect their
underlying integrity. The NRC is
proposing to adapt the requirements of
§ 55.49 in § 53.780(d) to require that
examinations and related activities
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remain free from any compromise that
might affect the integrity of the
examination process.
Simulators provide a valuable means
of training and evaluating plant
operators, and the NRC is specifically
authorized under the Nuclear Waste
Policy Act of 1982, as amended
(NWPA), section 306 (42 U.S.C. 10226)
to establish regulations for the use of
simulators within such context. The
NRC is proposing to adapt the
requirements of § 55.46 in § 53.780(e) to
address the use of simulation facilities
for training, examinations, and
applicant experience requirements, as
well as to address the maintenance of
simulator fidelity. However, the
proposed requirements of part 53 would
not mandate that full scope, plantreferenced simulators be used and
would allow the use of alternative
simulation facilities consisting of, for
example, partial scope simulators or the
plant itself, provided that all associated
requirements can be demonstrated to be
met using alternative approaches and
methods. Additionally, in allowing for
the possibility that an applicant or
licensee might demonstrate compliance
with training, examination, or
experience requirements using the plant
itself, the NRC is not allowing the
initiation of transients on the actual
plant. Consistent with this, aside from
controlled reactivity manipulations that
are conducted for the purposes of
demonstrating compliance with
experience requirements, actual plant
components may not be operated for
these purposes. Rather, the NRC
perspective is that the use of the plant
for training and examination purposes
should be restricted to techniques such
as walkthroughs, job performance
measures, simulated tasks, use of
augmented reality technology, and
similar approaches that provide training
and examination value while avoiding
the operation of actual plant
components.
There may be situations in which
applicants for operator or senior
operator licenses have previous training
and experience that justifies waiving
some, or all, of the initial examination
requirements. The NRC is proposing to
adapt the requirements of § 55.47 in
§ 53.780(f) to allow for consideration of
requests for waivers of examinations
requirements. In contrast with the part
55 requirements, the NRC proposes to
locate certain details associated with
such waiver requests within guidance
documentation in lieu of placement
within the rule itself.
For licensed operators and senior
operators to perform their assigned
duties safely and reliably, it is essential
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that they perform those duties
frequently enough so as to maintain a
sufficient degree of proficiency. The
NRC is proposing to adapt the
requirements of § 55.53(e) and (f) in
§ 53.780(g) to require that specifically
licensed operators and senior operators
maintain proficiency and, if proficiency
is not maintained, regain proficiency
prior to resuming licensed duties.
However, in recognition of the fact that
varying concepts of operations are
possible for advanced reactor facilities,
the NRC is proposing, in contrast with
the requirements of part 55, to allow
facility licensees to establish their own
programs for operator proficiency,
subject to NRC approval.
As the holders of specific licenses,
licensed operators and senior operators
must be subject to license conditions on
an individual basis to ensure that the
basis upon which the licenses were
issued remains valid. The NRC is
proposing to adapt the requirements of
§ 55.53 in § 53.785 to require
appropriate conditions of licenses for
specifically licensed operators and
senior operators. However, in contrast
with the requirements of § 55.53(e) and
(f), the NRC is proposing to allow
certain aspects of operator proficiency
to be addressed by an NRC-approved
facility proficiency program.
Licenses for specifically licensed
operators and senior operators are
issued by the NRC and must remain
subject to modification or revocation.
The NRC is proposing to adapt the
requirements of §§ 55.51 and 55.61 in
§ 53.790 to address the issuance,
modification, and revocation of licenses
issued to specifically licensed operators
and senior operators.
The licenses issued to specifically
licensed operators and senior operators
are valid for a period of six years, after
which they expire, unless otherwise
renewed. The NRC is proposing to adapt
the requirements of §§ 55.55 and 55.57
in § 53.795 to address the expiration and
renewal of licenses issued to
specifically licensed operators and
senior operators.
In developing this proposed rule, the
NRC has discussed with stakeholders
the considerations that might justify the
omission of the specifically licensed
operators and senior operators.
However, even for an inherently safe
reactor with autonomous operation
features, certain important
administrative functions (e.g.,
compliance with TS, operability
determinations, NRC notifications,
emergency declarations, risk
assessment, maintenance oversight, and
radiological release limit compliance)
would still need to be accomplished by
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appropriately qualified and authorized
individuals. Additionally, the NRC
recognized that manual manipulations
of facility reactivity controls must only
be performed by individuals who have
been appropriately licensed by the
Commission. The NRC therefore
proposes under § 53.800 to establish a
new class of facility (defined as a selfreliant-mitigation facility), according to
the criteria contained in § 53.800 for
part 53. These facilities would employ
GLROs rather than specifically licensed
operators and senior operators. The
GLRO regulations offer enhanced
flexibilities and targeted relaxations in a
manner that is commensurate with the
modified role of such operators to
ensure the safe operation of the
associated facilities. In contrast, those
facilities not meeting the criteria of
§ 53.800 would instead be considered
interaction-dependent-mitigation
facilities and would require staffing by
specifically licensed operators and
senior operators. The terminology used
to designate these facility types reflects
differences in how operators are
anticipated to need to interact with their
plant systems in mitigating events and
achieving safe outcomes; such systems
may either need operators to interact
with them in some manner (i.e., be
interaction-dependent) or may instead
be able to rely fully upon their own
capabilities independent of operator
interaction (i.e., be self-reliant).
Generally licensed reactor operators
would differ from specifically licensed
operators because the latter would be
directly and independently evaluated by
the NRC as part of their licensing
process. This direct and independent
evaluation remains appropriate when
operators may reasonably be expected to
exert a significant influence on public
health and safety outcomes. Therefore, a
key determinant as to whether generally
licensed reactor operators can be
utilized in facility staffing is the
assessment of the operator’s role in
maintaining and fulfilling safety
functions at the facility, such as through
the performance of credited actions for
the mitigation of plant events.
The criteria proposed in § 53.800
would designate self-reliant-mitigation
facilities. These criteria are derived from
the following set of considerations:
• no human action needed to satisfy
radiological consequence criteria;
• no human action needed to address
LBEs;
• safety functions not allocated to
human action;
• reliance upon robust and highly
reliable safety features; and
• adequate defense in depth achieved
without reliance on human action.
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It should be noted that those facilities
not meeting the criteria proposed in
§ 53.800 would instead be classified as
interaction-dependent-mitigation
facilities and would require staffing by
specifically licensed operators and
senior operators instead.
Generally licensed reactor operators
would perform duties under the
provisions of a general license that
would be effective without the filing of
an application with the Commission or
the issuance of licensing documents to
a particular person. The NRC proposes
requirements for the general licensing
process for GLROs under §§ 53.805
through 53.820. The requirements for
GLROs would parallel those for senior
operators in regard to their comparable
administrative responsibilities.
Nonetheless, the requirements for
GLROs would be relaxed and
incorporate greater flexibilities
compared to the requirements for
specifically licensed operators in a
manner that is consistent with the
GLRO’s role in safety at self-reliantmitigation facilities.
In order to use GLROs in lieu of
specifically licensed operators and
senior operators, a OL/COL applicant
would need to demonstrate that its
proposed facility is a self-reliantmitigation facility, i.e., that it will
comply with the following requirements
on an ongoing basis: maintaining GLRO
qualifications for the performance of
important functions and tasks;
incorporating relevant programmatic
controls into TS; administering the
related programs for training,
examination, and proficiency; and
ensuring that the relevant provisions of
parts 26 and 73 are met. Additionally,
to provide for an accurate accounting of
what individuals are licensed under the
general license, facility licensees would
be required to report the identities of all
generally licensed reactor operators to
the NRC on an annual basis.
Furthermore, a facility licensee must
ensure that the facility design and
performance continue to meet the
technological criteria to be classified as
a self-reliant-mitigation facility (i.e., the
criteria of § 53.800) on a continual basis
during the operating phase, as the
relaxations afforded to such facilities in
the areas of operator licensing, staffing,
and HFE would be predicated on this
assumption. The NRC therefore
proposes under § 53.805 to establish
requirements for facility licensees that
address issues such as these. Finally,
the failure of a self-reliant-mitigation
facility to subsequently meet the criteria
of § 53.800 after the issuance of an OL
or COL would constitute a reportable
event (i.e., an unanalyzed condition that
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significantly degrades plant safety)
under the provisions of § 53.1630.
The NRC proposes the general license
for GLROs under § 53.810. GLROs
would be licensed as a class of
individuals under the provision of
§ 53.810(a) and would be subject to the
conditions specified in § 53.810(b)
through (g). Portions of these conditions
are adapted from § 55.53 and from those
conditions currently included in the
licenses issued to specifically licensed
operators and senior operators. The NRC
would retain the ability to suspend or
prohibit individuals from operating
under the general license should such
action be warranted.
The NRC proposes overall
programmatic requirements for GLRO
training, examination, and proficiency
under § 53.815. In general, these
proposed requirements are adapted from
those of part 55 and parallel those also
proposed for specifically licensed senior
operators in § 53.780. These
requirements include increased
flexibilities and several targeted
relaxations that reflect the limited role
of GLROs in facility safety. The
requirements proposed under § 53.815
cover, in part, the initial training, initial
examination, continuing training,
requalification examination, and
proficiency of GLROs. Section 53.805
would require the facility licensee to
develop, implement, and maintain these
programs. Section 53.810, in turn,
would prescribe that the requirements
of § 53.805 would need to be met as a
requirement of the general license. The
implication of this structure is that the
facility licensee would need to
implement these programs for training,
examination, and proficiency, and
GLROs would need to participate in
these programs to demonstrate
compliance with the requirements of the
general license.
The initial training process provides
GLROs with the knowledge and abilities
needed to fulfill assigned duties as
GLROs. The use of a SAT serves to
ensure that the training program is
based upon job requirements in a
manner that can be adapted to account
for differences in plant technology and
concepts of operations. The NRC is
proposing under § 53.815(b) to require
facility licensees to implement a SATbased training program for the initial
training of GLROs that is adequate to
ensure that they have the necessary
knowledge, skills, and abilities to
perform their duties. The NRC further
proposes that such programs would be
subject to NRC approval, oversight, and
appropriate change control processes.
The training program must ensure that
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GLROs maintain the necessary
knowledge, skills, and abilities.
Examinations provide a means of
assessing that individuals have achieved
a degree of knowledge and ability that
will be sufficient to enable them to carry
out assigned duties as GLROs in a
manner that is both safe and reliable.
The NRC proposes to adapt the
requirements of §§ 55.40, 55.41, 55.43,
and 55.45 in § 53.815(b) to require that
facility licensees establish and
implement an initial examination
program. A key difference from the
comparable requirements of part 55
would be that facility licensees would
be afforded the flexibility to propose,
subject to NRC approval, the
examination methods and criteria to be
used in assessing satisfactory individual
performance. Such examination
programs (including those used within
the scope of continuing training) would
need to provide for acceptable levels of
both test validity and test reliability in
order to be considered acceptable. The
NRC intends that staff guidance would
be available to facilitate the review of
initial examination programs that are
proposed by facility licensees and that
approved initial examination programs
would be subject to an appropriate
change control process. In contrast with
both the requirements of part 55 and the
proposed requirements of § 53.780, the
NRC does not intend to administer or
evaluate these initial examinations.
However, the examination processes
themselves will continue to be subject
to ongoing NRC oversight. Irrespective
of the provided flexibilities in
examination format and structure,
topics from the following general
categories of knowledge and abilities
should be sampled in such
examinations:
• Reactor Theory, Thermodynamics,
and Chemical Interactions
• Plant Systems and Components
• Reactivity Management and
Manipulations
• Radiation Control and Safety
• Emergency, Abnormal, and Normal
Operations
• Administrative Requirements and
Conditions of the Facility License
Continuing training programs provide
the ongoing training and examination of
GLROs to ensure that they maintain the
knowledge and abilities needed to
support the safe and reliable
performance of job duties following the
completion of an initial training and
examination program. The NRC is
proposing to adapt the requirements of
§ 55.59 in § 53.815(b) to require that
facility licensees implement both a
SAT-based continuing training program
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and a requalification examination
program. However, a notable difference
from the examinations required under
part 55 would be that distinct annual
operating test and biennial written
examination components would not be
mandated. The facility licensee would
instead propose examination methods
and criteria to be used in assessing
satisfactory performance. Furthermore,
unlike the comparable requirements of
part 55 and those proposed for
specifically licensed operators and
senior operators, a biennial periodicity
for requalification examinations would
not be prescribed. However, adequate
justification for the proposed periodicity
of requalification examinations would
be required. The NRC intends that staff
guidance would be available to facilitate
the review of the requalification
examination programs that are proposed
by facility licensees. Approved
requalification examination programs
would be subject to an appropriate
change control process.
For examinations to provide for valid
assessments of the knowledge and
abilities of individuals, the
examinations must remain free from
compromises that could affect their
underlying integrity. The NRC is
proposing to adapt the requirements of
§ 55.49 in § 53.815(d) to require that
examinations and related activities
remain free from any compromise that
might affect the integrity of the
examination process.
Simulators provide a valuable means
of training and evaluating plant
operators and the NRC is specifically
authorized under the NWPA, section
306 (42 U.S.C. 10226) to establish
regulations for the use of simulators
within such context. The NRC is
proposing to adapt the requirements of
§ 55.46 in § 53.815(e) to address the use
of simulation facilities for training and
examinations, and experience
requirements, as well as to address the
maintenance of simulator fidelity. The
use of full scope, plant-referenced
simulators would not be mandated. The
potential use of alternative simulation
facilities consisting of, for example,
partial scope simulators or the plant
itself, would be allowed provided that
all associated requirements could be
demonstrated to be met using
alternative approaches and methods.
Additionally, in allowing for the
possibility that an applicant or licensee
might demonstrate compliance with
training and examination requirements
using the plant itself, the NRC is not
allowing the initiation of transients on
the actual plant. Consistent with this,
aside from controlled reactivity
manipulations that are conducted for
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the purposes of demonstrating
compliance with experience
requirements, actual plant components
may not be operated for these purposes.
Rather, the use of the plant for training
and examination purposes should be
restricted to techniques such as
walkthroughs, job performance
measures, simulated tasks, use of
augmented reality technology, and
similar approaches that provide training
and examination value while avoiding
the operation of actual plant
components.
There may be situations in which
GLROs have previous training and
experience that justifies waiving some,
or all, of the initial examination.
Therefore, the NRC is proposing under
§ 53.815(f) to allow facility licensees to
waive some, or all, portions of initial
examinations provided that such
waivers are consistent with a program
that has been approved by the NRC.
For GLROs to safely and reliably
perform their assigned duties, it is
essential that they perform those duties
frequently enough so as to maintain a
sufficient degree of proficiency.
However, the NRC recognizes that
facilities that utilize GLROs may have
concepts of operation that warrant
unique proficiency considerations.
Therefore, the NRC is proposing in
§ 53.815(g) to require that facility
licensees develop, implement, and
maintain programs to maintain and
reestablish, if needed, the proficiency of
GLROs. This could occur, for example,
if an individual’s extended absence
from watch standing has rendered
proficiency requirements unmet.
The general license should remain in
effect for an individual only while that
individual remains employed in a
position that may call for the individual
to manipulate the reactivity controls of
the facility. The NRC proposes under
§ 53.820 to require that the general
license would cease to be applicable on
an individual basis when an
individual’s employment status
becomes such that this is no longer the
case. However, the NRC recognizes that
for some types of self-reliant-mitigation
facilities, very long periods may elapse
between circumstances that necessitate
manual manipulation of reactivity
controls. Therefore, the general license
remains in effect for an individual as
long as the individual’s current position
could potentially require that individual
to manipulate reactivity controls at
some point within the course of the
individual’s assigned job duties.
The NWPA, section 306 (42 U.S.C.
10226) authorizes and directs the NRC
to, in part, issue regulations and
guidance that address the training and
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qualifications of civilian nuclear power
plant operators, supervisors,
technicians, and other appropriate
operating personnel. The NRC
implements this in part 50 through the
requirements of § 50.120, ‘‘Training and
qualification of nuclear power plant
personnel.’’ The NRC is proposing
under § 53.830 to adapt, with
modifications, the requirements of
§ 50.120 for use in part 53 to provide
more flexible personnel training and
qualification requirements than those in
§ 50.120 and better reflect diverse
concepts of operations.
The NRC recognizes that the
categories of nuclear power plant
personnel in § 50.120 may not be
needed for the diverse concepts of
operations, staffing models, and nontraditional personnel roles and
responsibilities anticipated under
proposed part 53; conversely, and for
the same reasons, additional categories
of plant personnel may need to be
covered by part 53. The NRC also
recognizes that the timeframe prescribed
in § 50.120 for the establishment of
training programs may not be aligned
with the schedules associated with the
startup of certain types of commercial
nuclear plant facilities. However, the
NRC also recognizes that the SAT-based
training required under § 50.120
remains an appropriate means by which
training programs should continue to be
developed and implemented. Therefore,
the approach taken by the NRC in
addressing the training of certain plant
staff under the proposed part 53 reflects
greater flexibilities in personnel
categories and programmatic
timeframes, while still retaining the
requirement that such training programs
be based on SAT.
The NRC is proposing under § 53.830
to require SAT-based training programs
with the timeframe for when such
programs are required being based upon
when the associated personnel are
needed to support facility-specific
needs. The training programs would
cover the training and qualification of
plant personnel in the general categories
of supervisors, technicians, and other
appropriate operating personnel. The
licensee would not be required to seek
NRC approval of a training program
prior to usage. However, the licensee is
required to accommodate NRC
inspection of the training program. The
NRC intends to develop guidance to
facilitate the inspection of these training
programs but does not intend for such
guidance to preclude the potential for
the training programs to be maintained
by a separate, NRC-approved
accreditation process.
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The proposed § 53.845 would require
programs to be developed,
implemented, and maintained to help
ensure that design features and human
actions have the capabilities and
reliabilities necessary to demonstrate
compliance with the safety criteria in
subpart B throughout the operating life
of each commercial nuclear plant. The
proposed programmatic requirements in
subpart F would also address areas such
as radiation protection needed to
control routine effluents during normal
operations. The proposed §§ 53.850
through 53.910 would require programs
to support specific activities needed to
ensure the prevention or mitigation of
unplanned events or to support normal
operations for any reactor design.
However, each holder of an OL or COL
would be required to assess whether
additional programs are needed for the
specific reactor design and location of
the commercial nuclear plant. Licensees
would be able to combine, separate, and
otherwise organize programs and related
documents as appropriate for the
technologies and organizations
associated with the commercial nuclear
plant.
Proposed § 53.850 would require a
radiation protection program associated
with the requirements in subparts B and
C for public doses resulting from normal
operations and the protection of plant
workers. The proposed requirements
related to doses from normal operations,
including routine effluents, would be
similar to those specified in § 50.36a,
‘‘Technical specifications on effluents
from nuclear power reactors,’’ and
related requirements in standard TS for
offsite dose calculation manuals. While
the proposed section would include
requirements that are technically and
programmatically similar to part 50,
proposed § 53.850 would not include a
requirement for effluent-related TS as is
required in § 50.36a. A proposed
requirement similar to that found in the
administrative controls section of TS for
operating reactors licensed under parts
50 and 52 would be included for
programmatic controls of solid wastes to
complement the design requirements in
proposed § 53.425.
Proposed § 53.855 would require an
emergency response plan that
demonstrates compliance with the
requirements in appendix E to part 50
and § 50.47(b) or § 50.160. The
regulations in § 50.47 stating that the
NRC will not issue certain licenses
unless it finds that there is reasonable
assurance that adequate protective
measures can and will be taken to
protect public health and safety in the
event of a radiological emergency apply
equally to applications under part 53
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86941
complying with the applicable
standards set forth in either § 50.160 or
the requirements in appendix E to part
50 and § 50.47(b).
In its 2008 Advanced Reactor Policy
Statement, the Commission stated their
expectation that ‘‘the safety features of
advanced reactor designs will be
complemented by the operational
program for Emergency Planning (EP).
This EP operational program, in turn,
must be demonstrated by inspections,
tests, analyses, and acceptance criteria
to ensure effective implementation of
established measures.’’ Consistent with
this policy statement, emergency plans
and emergency planning zones are not
safety features in the design. In SECY–
97–020, ‘‘Results of Evaluation of
Emergency Planning for Evolutionary
and Advanced Reactors,’’ dated January
27, 1997, the staff indicated that the
rationale upon which EP for current
reactor designs is based, that is,
potential consequences from a spectrum
of accidents, is appropriate for use as
the basis for EP for evolutionary and
passive advanced LWR designs and is
consistent with the Commission’s
defense-in-depth safety philosophy.
Also, in its Safety Goals Policy
Statement the Commission stated that:
‘‘A defense-in-depth approach has been
mandated in order to prevent accidents
from happening and to mitigate their
consequences. Siting in less populated
areas is emphasized. Furthermore,
emergency response capabilities are
mandated to provide additional defensein-depth protection to the surrounding
population.’’ Consistent with this policy
statement, proposed § 53.855
contributes an additional independent
layer of defense in depth for commercial
nuclear plants. Therefore, the
emergency plans and emergency
planning zones under proposed § 53.855
are not used to demonstrate compliance
with subpart B and subpart C of this
part. Rather, compliance with the
requirements in proposed § 53.855
would provide reasonable assurance
that adequate protective measures can
and will be taken to protect public
health and safety in the event of a
radiological emergency.
Proposed § 53.860 would identify the
applicable regulations for part 53
applicants related to the programs for
physical security, cybersecurity, FFD,
AA, and information security. These
programs are discussed in more detail in
section V, ‘‘Changes to Other Parts of 10
CFR,’’ of this document.
Proposed § 53.860(a) would establish
the physical protection program and
present a graded approach to physical
protection requirements. If a licensee
can meet the proposed criterion in
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§ 53.860(a)(2)(i), then the requirement to
protect against the design-basis threat
(DBT) of radiological sabotage would
not be applicable. The criterion in
§ 53.860(a)(2)(i) would require a
licensee to show that potential
consequences resulting from a DBT
initiated event would result in offsite
doses below the values in § 53.210 even
if licensee mitigation and recovery
actions, including any operator action,
are unavailable or ineffective. Where the
criterion is met, the resulting physical
protection requirements would be those
for protection of SNM and Category 1
and Category 2 radioactive material, if
applicable. This proposal would apply a
new regulatory approach for certain
commercial nuclear plants in which the
DBT of radiological sabotage would not
be applicable.
For those licensees able to meet the
criterion in § 53.860(a)(2), the NRC
would not conduct Force-On-Force
(FOF) exercise inspections. Section
170D.a of the Act permits the
Commission to determine which
licensed facilities are part of a class of
licensed facilities where NRCconducted FOF exercises are
appropriate to assess the ability of a
private security force of a licensed
facility to defend against any applicable
DBT. For the class of licensees that meet
the criterion of § 53.860(a)(2), it would
not be appropriate to conduct FOF
exercises to evaluate performance at
commercial nuclear plants where the
DBT of radiological sabotage is not
applicable and the facility poses a lower
risk to public health and safety from
potential radiation exposure. These
facilities would still have tailored
security requirements and oversight
consistent with their relatively low risk.
For those licensees not able to meet
the criterion in § 53.860(a)(2), proposed
§ 53.860(a) would permit the licensee to
choose one of two paths to provide
physical protection: (1) the current set
of requirements in § 73.55, which would
include any changes resulting from the
ongoing proposed rulemaking on
Alternative Physical Security
Requirements for Advanced Reactors 2
that provides pre-determined physical
security alternatives; or (2) the
performance-based requirements in
proposed § 73.100. In either case, the
licensee would be subject to NRCconducted FOF inspections.
Proposed § 53.860(b) would require
licensees to establish, implement, and
maintain an FFD program under part 26.
Section 53.860(c) would require
2 SECY–22–0072, ‘‘Proposed Rule: Alternative
Physical Security Requirements for Advanced
Reactors,’’ dated August 2, 2022.
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licensees to establish, implement, and
maintain an AA program in accordance
with either § 73.56 or proposed § 73.120,
as appropriate. Section 53.860(d) would
require licensees to establish,
implement, and maintain a
cybersecurity program in accordance
with either § 73.54 or proposed § 73.110.
Section 53.860(e) would require
licensees to establish, implement, and
maintain an information protection
system that complies with the
requirements of §§ 73.21, 73.22, and
73.23, as applicable.
Proposed § 53.865 would establish
requirements for quality assurance and
refer to appendix B of part 50 for the
part 53 requirements for SR design
features. Proposed requirements related
to evaluating and reporting changes to
the quality assurance program would be
included in proposed subpart I and
would be equivalent to those found in
§ 50.54.
The proposed § 53.870 would require
licensees to actively assess possible
degradation of SSCs from the effects of
aging, fatigue, and environmental
conditions. The proposed inclusion of
requirements related to designing and
monitoring for possible degradation
mechanisms reflects important lessons
learned from the history of LWRs and
the likely introduction of new design
features and materials in future
commercial nuclear plants. The
allowable combinations of design
features, operating experience, testing,
and monitoring during operations
would support performance-based
approaches to the initial licensing of
new technologies. The proposed
performance-based approach to integrity
assessment programs would also allow
for the subsequent consideration of
operating experience and appropriate
corrective actions or allowable
relaxations for ensuring that design
features comply with the proposed
functional design criteria of §§ 53.410
and 53.420. The proposed program
would be based upon a comprehensive
and integrated evaluation of the aging
and other degradation mechanisms
applicable to the design; identification
of the affected SSCs; the allowances
provided in the design of the SSCs for
degradation; and schedules and
procedures for determining if and at
what rate degradation is occurring, as
well as its cause. Risk insights could be
used to prioritize the monitoring,
evaluation, and management of
degradation based upon the importance
of the SSC to safety and the time frame
for when the effects of degradation
could be of concern.
Proposed § 53.875 would establish
requirements for a fire protection
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program supporting operations similar
to § 50.48. The proposed fire protection
program during operations would work
in concert with specific fire protection
requirements proposed in subpart C for
design and analyses and in proposed
subpart E for construction and
manufacturing.
Proposed § 53.880 would establish
requirements for an inservice inspection
(ISI) and inservice testing (IST) program,
which are historically important
activities conducted in accordance with
ASME codes and regulations in
§ 50.55a. While the proposed part 53
would not incorporate specific
consensus codes and standards into the
regulations, § 53.880 allows for the use
of generally accepted codes and
standards. The proposed requirement
for an ISI and IST program would
reinforce the need to develop
monitoring programs to be conducted
during a plant’s operations phase to
complement the design process and
address inherent uncertainties. The NRC
encourages the continued use of
consensus codes and standards
supporting design, testing, and
inspections to support integrated and
performance-based approaches in
demonstrating compliance with the
proposed requirements in part 53.
Proposed § 53.910 would establish
requirements for developing,
implementing, and maintaining
procedures (e.g., operations and
emergency operating procedures) and
guidelines (e.g., accident management
guidelines). The programmatic
requirements for many of the
procedures listed in this proposed
section would be similar to the
requirements found in the
administrative controls section of TS for
plants licensed under parts 50 and 52.
The proposed inclusion, where
appropriate, of accident management
guidelines in these requirements is
intended to ensure that an integrated set
of procedures and guidelines would be
established by licensees to ensure
command and control across the
spectrum of possible event sequences.
The proposed required procedures
would also include those needed to
complement the design requirements in
proposed § 53.440(m) related to
criticality alarms and the equivalent of
the procedures required in § 50.54(hh)
to address notifications of potential
aircraft threats.
Subpart G—Decommissioning
Requirements
The proposed subpart G would
provide the regulatory requirements for
the decommissioning phase of the life
cycle of a commercial nuclear plant.
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The requirements being proposed in
subpart G for the decommissioning of a
commercial nuclear plant are adapted
from the current regulations in § 50.75,
‘‘Reporting and recordkeeping for
decommissioning planning,’’ § 50.82,
‘‘Termination of license,’’ and § 50.83,
‘‘Release of part of a power reactor
facility or site for unrestricted use.’’
Although the requirements from those
sections of part 50 have been copied
into proposed subpart G with relatively
few changes, the requirements are
reorganized to fit within the part 53
structure. The few changes made were
primarily to make the proposed
requirements more technology inclusive
by adding alternatives within sections,
whereas some requirements in part 50
were developed specifically for LWRs.
As an example, § 50.75 provides
minimum amounts of decommissioning
funds required to demonstrate
reasonable assurance of funds for
decommissioning LWRs. Such generic
amounts have not been developed for all
reactor technologies that may be
licensed under part 53. Therefore, the
Commission proposes in § 53.1020,
‘‘Cost estimates for decommissioning,’’
that site-specific cost estimates for
decommissioning must be developed
considering costs in such areas as
engineering, labor, and waste disposal.
The derivation of the generic cost
estimates for LWRs in § 50.75 is
provided in NUREG/CR–5884, ‘‘Revised
Analyses of Decommissioning for the
Reference Pressurized Water Reactor
Power Station,’’ and NUREG/CR–6187,
‘‘Revised Analyses of Decommissioning
for the Reference Boiling Water Reactor
Power Station.’’ Similar to part 50, a
provision for an annual adjustment of
decommissioning cost estimates would
be included in proposed § 53.1030.
The NRC is currently pursuing
another rulemaking, ‘‘Regulatory
Improvements for Production and
Utilization Facilities Transitioning to
Decommissioning,’’ which was
published as a proposed rule for public
comment on March 3, 2022 (87 FR
12254). As these rulemakings progress,
the NRC will consider revisions to part
53 to align the two rulemaking efforts.
For example, the proposed § 53.1075
could be expanded to include or
reference requirements for
decommissioning in areas such as EP
and security in addition to the proposed
decommissioning fire protection plans
that would provide an equivalent to
§ 50.48(f).
Subpart H—Licenses, Certifications, and
Approvals
Proposed subpart H would provide
requirements related to applications
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under part 53 for NRC licenses,
certifications, or approvals for
commercial nuclear plants.
Proposed subpart H would specify
requirements applicable to all part 53
applications as well as requirements
specific to part 53 applications for
LWAs, ESPs, standard design approvals,
standard DCs, MLs, CPs, OLs, and COLs.
Proposed subpart H would be
equivalent to and include all existing
licensing, certification, and approval
processes currently covered under parts
50 and 52, with the exception of the
process for early review of site
suitability issues. Interactions with
external stakeholders during the
development of the proposed rule did
not identify significant interest in or
need for including the process for early
review of site suitability issues in part
53.
Much of the proposed subpart H
regulatory text is identical to the
corresponding language in parts 50 and
52, with minor changes to account for
cross references in part 53, to make
language technology neutral, or to
reflect the unique analytical approach in
part 53. In these instances, this
preamble discussion will describe the
language as ‘‘equivalent’’ to the existing
corresponding requirement in part 50 or
part 52 and will describe any
deviations, where applicable.
Because part 53 carries over the
majority of the licensing options from
parts 50 and 52, there are several
sections in proposed subpart H that are
similar to existing regulations in parts
50 and 52. Proposed § 53.1100 would
address filing of applications for
licenses, certifications, or approvals
under oath or affirmation and is
equivalent to § 50.30. The proposed
§ 53.1100 does not include the current
requirement in § 50.30(a)(2) that the
applicant maintain the capability to
generate additional copies, because it is
unnecessary in the age of electronic
submissions. In addition, the existing
requirement on applications for OLs in
§ 50.30(d) is included in proposed
§ 53.1124(g)(2), ‘‘Relationship between
sections,’’ covering OLs, rather than in
proposed § 53.1100.
Proposed § 53.1101 would lay out
activities requiring an NRC license and
is equivalent to § 50.10(b). Proposed
§ 53.1103 would address combining
applications and is equivalent to
§§ 50.31, 50.52, and 52.8. Proposed
§ 53.1103(b) would continue the
Commission’s practice of combining
multiple authorizations for a facility
under parts 30, 40, 50, 52, and 70 into
one license based on the Commission’s
authority under Section 161h. of the Act
to combine NRC licenses. Proposed
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§ 53.1106 would address elimination of
repetition and is equivalent to § 50.32.
Proposed § 53.1109 would provide
general information requirements for the
content of applications submitted to the
NRC under part 53 and is equivalent to
§ 50.33, with the exception of § 50.33(f)
on financial qualifications, which is
covered in proposed subpart J, and
§ 50.33(h) on earliest and latest dates for
completion of construction, which is
covered in § 53.1306 of this subpart.
Each application would need to include
information to address the items in
proposed § 53.1109 as cited in the
appropriate section of this subpart for
the application type.
One change from current
requirements can be found in proposed
§ 53.1109(i), which is not limited to
electricity generation as it is currently in
part 50. Some prospective NRC
applicants are considering development
of nuclear plants for other commercial
ventures, such as process heat
generation or hydrogen production. In
addition, § 53.1109(j), which requires
applications containing classified
information to separate that information
from the unclassified information in the
application, refers to ‘‘Restricted Data or
classified National Security
Information’’ instead of the term used in
the corresponding provision in
§ 50.33(j), ‘‘Restricted Data or other
defense information.’’ This change was
made to use the defined term in part 95
rather than ‘‘defense information’’ as
used in § 50.33(j). The usage in § 50.33(j)
dates back to the Atomic Energy
Commission amendment of that section
on January 19, 1956 (21 FR 355, 357)
and was not changed with the issuance
of part 95 (45 FR 14476; March 5, 1980)
after the establishment of the NRC and
the 1975 reissuance of the former
Atomic Energy Commission regulations.
The revised terminology also aligns
with its usage in § 53.1115.
Proposed § 53.1112 would address
environmental conditions and is
equivalent to § 50.36b. Proposed
§ 53.1115 would address requirements
for agreements limiting access to
classified information and is equivalent
to § 50.37. Proposed § 53.1118 would
address ineligibility of certain
applicants and is equivalent to § 50.38.
Proposed § 53.1120 would address
exceptions and exemptions from
licensing requirements for Department
of Defense and DOE facilities and is
equivalent to § 50.11. Proposed
§ 53.1121 would address public
inspection of applications and is
equivalent to § 50.39.
Proposed § 53.1124 would address the
relationship between the various
licenses, certifications, and approvals
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provided in this subpart, and the
requirements are equivalent to a number
of similar provisions in parts 50 and 52
including §§ 50.10, 52.13, 52.43, 52.73,
52.133, and 52.153. New provisions are
provided in § 53.1124(c) and (d), that
would allow an application for either a
standard design approval or a standard
DC under part 53 to reference applicable
licensing-basis information that
supported issuance of an OL or COL
under part 53. These provisions are
being proposed to offer additional
flexibility beyond what is currently
allowed under parts 50 or 52 for an
applicant who may wish to license a
first-of-a-kind reactor for operation prior
to seeking generic approval or
certification of the standard design.
Proposed § 53.1124(e) would address
the limitations that a manufactured
reactor may only be transported to a site
with a COL and is equivalent to
§ 52.153. Proposed § 53.1130 would
address LWAs and is equivalent to
§ 50.10.
Proposed §§ 53.1140 through 53.1188
would govern the content of ESP
applications. Proposed § 53.1140 is
equivalent to § 52.12. Proposed
§ 53.1143 would address filing of
applications and is equivalent to
§ 52.15. Proposed § 53.1144 would
address general information
requirements for the content of
applications and is equivalent to
§ 52.16.
Proposed § 53.1146 would specify
requirements for the technical contents
of applications and is equivalent to
§ 52.17. Proposed § 53.1146(b)(2)
provides applicants for ESPs a
regulatory option to propose major
features of the emergency plans or
complete and integrated emergency
plans in accordance with either the
requirements in § 50.160 of this chapter,
or the requirements in appendix E to
part 50 of this chapter and § 50.47(b) of
this chapter, as applicable.
Proposed § 53.1149 would address
standards for review of ESP applications
and administrative review of
applications, including hearings, and is
equivalent to §§ 52.18 and 52.21.
Proposed § 53.1155 would address
referral to the ACRS and is equivalent
to § 52.23. Proposed § 53.1158 would
address issuance of ESPs and is
equivalent to § 52.24. Proposed
§ 53.1161 would address the extent of
activities permitted and is equivalent to
§ 52.25. Proposed § 53.1164 would
address the duration of an ESP and is
equivalent to § 52.26. Proposed
§ 53.1167 would address provisions for
requesting a LWA after issuance of an
ESP and is equivalent to § 52.27.
Proposed § 53.1170 would address
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transfers of ESPs and is equivalent to
§ 52.28. Proposed § 53.1173 would
address applications for ESP renewals
and is equivalent to § 52.29. Proposed
§ 53.1176 would address criteria for
renewal of an ESP and is equivalent to
§ 52.31. Proposed § 53.1179 would
address the duration of an ESP renewal
and is equivalent to § 52.33. Proposed
§ 53.1182 would address the use of a
site for purposes other than those
described in the permit and is
equivalent to § 52.35. Proposed
§ 53.1188 would address finality of ESP
determinations and is equivalent to
§ 52.39.
Proposed §§ 53.1200 through 53.1221
would govern the contents of standard
design approval applications. Proposed
§ 53.1200 is equivalent to § 52.131.
Proposed § 53.1203 would address filing
of applications and is equivalent to
§ 52.135. Proposed § 53.1206 would
address general information
requirements for the content of
applications and is equivalent to
§ 52.136.
Proposed § 53.1209 would address
requirements for the technical content
of applications and is largely equivalent
to § 52.137. In proposed § 53.1209(a),
the NRC proposes text that expands the
discussion of ‘‘major portion’’ standard
design approvals. Additional discussion
regarding standard design approvals for
a major portion of a standard design can
be found in the NRC’s ‘‘A Regulatory
Review Roadmap for Non-Light Water
Reactors,’’ which considers the Nuclear
Innovation Alliance report ‘‘Clarifying
‘Major Portions’ of a Reactor Design in
Support of a Standard Design
Approval.’’ Proposed § 53.1209(b)
outlines the required content of the
Final Safety Analysis Report (FSAR).
Proposed requirements in
§ 53.1209(b)(2) for portions of the
application addressing design
information state that the application
must include design information
equivalent to that required for a
standard DC. This reference to the
pertinent DC requirements (specifically,
those in proposed § 53.1239(a)(2)
through (27)) is an efficiency that would
prevent the need to repeat many of the
same requirements for the content of a
standard design approval application.
Proposed § 53.1210 would address
requirements for the content of a
standard design approval application
other than the FSAR. Proposed
§ 53.1210(a) would require the inclusion
of a description of availability controls
that are not included in the FSAR.
Proposed § 53.1212 would address
standards for review of applications and
is equivalent to § 52.139. Proposed
§ 53.1215 would address referral to the
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ACRS and is equivalent to § 52.141.
Proposed § 53.1218 would address staff
approval of designs and duration of
design approvals and is equivalent to
§§ 52.143 and 52.147. Proposed
§ 53.1221 would address finality of
standard design approvals and
information requests and is equivalent
to § 52.145 with the exception that it
extends such finality to a standard
approval referenced in a DC application.
Standard design approvals issued to
date under part 52 have been issued
during the NRC’s review of the standard
DC application and have relied on the
same application content. However, a
future scenario could arise where the
DC application is not submitted until
after a design approval has been
granted. The NRC would apply the same
finality provisions in this situation as in
the situation where a standard design
approval is referenced in a COL
application.
There is no equivalent to proposed
§ 53.1221(d) in part 52 for standard
design approvals. This provision would
state that the Commission will require,
before granting a CP, COL, OL, or ML
which references a standard design
approval, that engineering documents
be completed and available for audit. A
similar provision is included in part 52
in relation to a standard DC; and the
NRC would require that design and
analysis information needed for the
Commission to make its safety
determination be complete and
available for any application the NRC is
reviewing. Making this explicit provides
increased clarity to future standard
design approval applicants under part
53.
Proposed §§ 53.1230 through 53.1263
would address standard DCs. Proposed
§ 53.1230 would address general
provisions for standard DCs and is
equivalent to § 52.41. Proposed
§ 53.1233 would address filing of
applications and is equivalent to
§ 52.45. Proposed § 53.1236 would
address general information
requirements for the content of
applications and is equivalent to
§ 52.46. Proposed § 53.1239 would
address requirements for the technical
content of applications and is
equivalent to § 52.47(a). The
requirements in proposed § 53.1239
have been modified from the analogous
requirements in § 52.47(a) to align with
the technical requirements in proposed
part 53.
Proposed § 53.1241 would address
requirements for the content of a
standard DC application other than the
FSAR and is equivalent to § 52.47(b)
and (d).
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Proposed § 53.1242 would address
review of applications and is equivalent
to §§ 52.48 and 52.51. Proposed
§ 53.1242(c) would include a provision
that would allow a DC applicant to
reference applicable licensing-basis
information for an OL or COL issued
under part 53. As explained previously,
this provision is being proposed to
explicitly allow flexibility for an
applicant who may wish to license a
first-of-a-kind reactor for operation prior
to seeking certification of the generic
reactor design. For NRC findings on a
reactor design in an OL or COL
proceeding, this proposal would
provide finality in a subsequent DC
application that references information
on the OL or COL proceeding’s docket.
This finality accorded to the OL or COL
findings would bind the NRC staff and
the ACRS but would not bind members
of the public or the Commission. (To the
extent an Atomic Safety and Licensing
Board (ASLB) might have a role in a DC
rulemaking, the OL or COL findings
would not bind the ASLB either.)
Specifically, members of the public
would have the opportunity to comment
on a proposed DC rule under wellestablished NRC practice. The rationale
for binding the NRC staff and ACRS is
similar to the rationale for a COL
applicant referencing a standard design
approval under part 52.
Proposed § 53.1245 would address
referral to the ACRS and is equivalent
to § 52.53. Proposed § 53.1248 would
address issuance of standard DCs and is
equivalent to § 52.54. Proposed
§ 53.1251 would address duration of
certifications and is equivalent to
§ 52.55(c). Proposed § 53.1254 would
address application for renewal and is
equivalent to § 52.57. Proposed
§ 53.1257 would address criteria for
renewal and is equivalent to § 52.59.
Proposed § 53.1260 would address
duration of renewals and is equivalent
to § 52.61. Proposed § 53.1263 would
address finality of standard DCs and is
equivalent to § 52.63.
Proposed §§ 53.1270 through 53.1291
would address MLs covering
manufacturing activities at one or more
licensee facilities. Proposed § 53.1270
would address the scope of these
sections and is equivalent to § 52.151.
Proposed § 53.1273 would address
filing of applications for an ML and is
equivalent to § 52.155(a).
Proposed § 53.1276 would address
general information requirements for the
content of ML applications and is
equivalent to § 52.156, with one
exception. Proposed § 53.1276 would
require each application for an ML to
also include the information required by
§ 53.1109(e). This information includes
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the type of license applied for, the use
to which the facility will be put, the
period of time for which the license is
sought, and a list of other licenses,
except operator’s licenses, issued or
applied for in connection with the
proposed facility to address the
potential variations in how MLs might
be formulated under the proposed part
53.
Proposed § 53.1279 would address
requirements for the technical content
of applications for MLs to be included
in the FSAR and is equivalent to
§ 52.157. In addition, the requirements
in proposed § 53.1279(a) and (b) have
been modified from the analogous
requirements in § 52.157 to align with
the technical requirements in proposed
part 53. Proposed § 53.1279(a)(2)
outlines the required content of the
application addressing design
information and states that the
application must include design
information equivalent to that required
for a standard DC. This reference to the
pertinent DC requirements is an
efficiency that would prevent the need
to repeat the same requirements for the
content of an ML application.
Proposed § 53.1279(c) would provide
application requirements related to the
deployment of the completed
manufactured reactor. Proposed
§ 53.1279(c)(1) would require inclusion
of information related to the procedures
governing the preparation of the
manufactured reactor for shipping to the
site where it is to be operated, the
conduct of shipping, and the
verification of the condition of the
shipped items upon receipt at the site.
Proposed § 53.1279(c)(2) would require
that the application include information
on the interaction of the design,
manufacture, and installation of a
manufactured reactor within the
applicant’s organization and the manner
by which the applicant will ensure close
integration between the designer,
contractors, and any licensee of a
facility in which the manufactured
reactor is to be installed. Finally,
proposed § 53.1279(c)(3) would require
that the application include a
description of the measures used for the
control of interfaces between the holder
of the ML and the holder of the COL for
the commercial nuclear plant at which
the manufactured reactor is to be
installed. This information is necessary
for the NRC to determine whether the
applicant would have appropriate
controls in place to ensure coordination
between parties involved in the design,
manufacture, and eventual operation of
any reactor manufactured under an ML.
Proposed § 53.1279(d) would include
additional requirements for application
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content for applicants seeking an ML for
manufactured reactors that will be
fueled at the factory under a 10 CFR part
70 license, consistent with the
requirements in § 53.620(d). These
provisions would require the
application to include information
related to loading fuel and the required
independent physical mechanisms to
prevent criticality and to otherwise
provide assurance that the fueled
manufactured reactor can be
successfully transported, installed, and
operated at a site for which the
Commission has issued a COL that
authorizes construction and operation of
a commercial nuclear plant using the
manufactured reactor.
Proposed § 53.1282 would provide
requirements for other application
content for MLs and is equivalent to
§ 52.158. Proposed § 53.1282(a)(1)
would provide requirements to include
in the ML application the ITAAC within
the scope of the ML that the COL holder
referencing the ML must satisfy.
Proposed § 53.1282(a)(2) would require
that the ITAAC from a referenced
standard design apply to the portions of
the ML design within the scope of the
referenced standard design. Proposed
§ 53.1282(a)(3) would state that the COL
application may include a notification
that required referenced standard DC
ITAAC have been satisfied at the
manufacturing facility.
Proposed § 53.1282(b) would require
an ML application to include an
environmental report and, consistent
with existing requirements, proposed
§ 53.1282(b)(2) would note that if the
ML application references a standard
DC, the environmental report need not
contain a discussion of severe accident
mitigation design alternatives for the
manufactured reactor as used in a
commercial nuclear plant.
Proposed § 53.1285 would provide
standards for review of applications and
administrative review of applications
for MLs, including hearings, and is
equivalent to §§ 52.159 and 52.163.
Proposed § 53.1286 would address
referral of applications to the ACRS and
is equivalent to § 52.165. Proposed
§ 53.1287 would address issuance of an
ML and is equivalent to § 52.167.
Proposed § 53.1288 would address
finality of MLs and is equivalent to
§ 52.171. Proposed § 53.1291 would
address the duration of MLs and is
equivalent to § 52.173. Proposed
§ 53.1293 would address the transfer of
MLs and is equivalent to § 52.175.
Proposed § 53.1295 would address the
renewal of MLs and is equivalent to
§§ 52.177, 52.179 and 52.181, with a
minor exception. Proposed
§ 53.1295(a)(3) would state that an ML
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for which a timely application for
renewal has been filed remains in effect
until the Commission has made a final
determination on the renewal
application, provided, however, that the
holder of an ML may not begin
manufacture of a manufactured reactor
less than six months before the
expiration of the license. The proposed
6-month time frame for this provision is
changed from the 3-year period in the
equivalent provision in part 52 because
future reactor applicants may present
smaller, simpler designs, to include
micro-reactor designs, in ML
applications than those that were
envisioned when the existing
requirements were written. A 6-month
time frame for this provision would
provide greater flexibility for ML
holders related to manufactured reactors
being produced when the ML expires.
Proposed §§ 53.1300 through 53.1348
would address licensing requirements
for CPs. Proposed § 53.1300 would set
out general requirements for CPs and is
equivalent to § 50.23. Proposed
§ 53.1306 would address the general
information requirements for the
content of applications for CPs and is
equivalent to § 50.33(f) and (h).
Proposed § 53.1309 would address
requirements for the technical content
of applications for CPs and includes the
requirement to submit a Preliminary
Safety Analysis Report (PSAR) that
describes the facility and presents a
preliminary safety analysis of the
facility as a whole. This is in contrast to
an OL application which is required to
include an FSAR that describes the
facility and presents a final safety
analysis of the facility as a whole.
Proposed § 53.1309 is equivalent to
§ 52.17(a)(1)(iv) through (a)(1)(x) and
52.17(b), with two exceptions. First,
proposed § 53.1309 would replace the
analysis of the dose criteria required by
§ 52.17(a)(1)(ix) with analysis to
demonstrate compliance with the safety
criteria defined in §§ 53.210 and 53.220.
Second, proposed § 53.1309(a)(2) would
add a requirement for a CP application
to include several categories of detailed
design information, although
§ 53.1309(a)(2)(ii) would allow certain
relaxations of this requirement in view
of aspects of a design that may not yet
be fully developed. Section 53.1309
would reference the requirements for
the content of an ESP application to
address application requirements
related to siting and would reference the
requirements for the content of a DC
application to address application
requirements related to design of the
commercial nuclear plant. Proposed
§ 53.1309(a)(2)(ii) would address the
treatment of preliminary design
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information and notes that information
provided in the application may include
some aspects of the design that are not
fully developed. This provision would
require that the completed design,
including any changes during
construction, be described in the FSAR
in an application for an OL. This would
include the requirement for a
description of the PRA required by
§ 53.450(a) and its results. Probabilistic
risk assessments developed for
commercial nuclear plants prior to
construction would be based on the
design and other information available
at the time of the CP application. PRAs
performed in early design stages or prior
to construction may be inherently less
detailed and may include projected
information that will be subsequently
verified or revised when the plant is
built. Proposed § 53.1309(a)(4) would
address preliminary description of the
plans for coping with emergencies.
Proposed § 53.1312 would address
other application content for CPs.
Proposed § 53.1312(a)(1) is equivalent to
§ 52.80(b) but is adapted for a CP
application. Proposed § 53.1312(a)(2) is
equivalent to § 52.80(c) but is adapted
for a CP application. Proposed
§ 53.1312(b)(1) is equivalent to
§ 52.79(b), (c), and (d) but is adapted for
a CP application. Section 53.1312(b)(2)
is equivalent to portions of
§§ 52.63(b)(1), 52.79(b)(1) through (b)(3),
(c), and (d)(1) and (d)(3), 52.80, and
52.93(b), but is adapted for a CP
application. Guidance for equivalent
requirements in parts 50 and 52 is also
addressed in RG 1.206, ‘‘Applications
for Nuclear Power Plants,’’ Revision 1,
section C.1.7.
Proposed § 53.1315 would address
standards for review of applications and
administrative review of applications,
including hearings, and is equivalent to
§§ 52.81 and 52.85, but is adapted for a
CP application.
Proposed § 53.1318 would address
finality of NRC approvals, licenses, and
certifications referenced in a CP
application and is equivalent to
§ 52.83(a) but is adapted for a CP
application.
Proposed § 53.1324 would address
referral to the ACRS and is equivalent
to § 50.58(a) and to § 52.87 but is
adapted for a CP application.
Proposed § 53.1327 would address
authorization to conduct LWA activities
and is equivalent to § 52.91 but is
adapted for a CP application. Proposed
§ 53.1327(a) is equivalent to § 52.91(a)
but is adapted for a CP application.
Proposed § 53.1327(b) is equivalent to
§ 52.91(b) but is adapted for a CP
application. Proposed § 53.1330 would
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address exemptions, departures, and
variances for CP applicants.
Proposed § 53.1333 would address
issuance of CPs. Proposed § 53.1333(a)
is equivalent to § 50.35(a). Proposed
§ 53.1333(b) is equivalent to § 50.35(b)
and to § 52.97(c) but is adapted for a CP
application. Proposed § 53.1336 would
address the effect of CPs and is
equivalent to § 50.35(b). Proposed
§ 53.1342 would address the duration of
CPs. Proposed § 53.1342(a) is equivalent
to § 50.55(a). Proposed § 53.1342(b) is
equivalent to § 50.55(b). Proposed
§ 53.1345 would address the transfer,
assignment, and disposal of CPs and is
equivalent to § 50.80. Proposed
§ 53.1348 would address the
termination of CPs and is equivalent to
§§ 52.3(b)(8) and 52.110(a)(1) but is
adapted for a CP application.
Proposed §§ 53.1360 through 53.1405
address requirements for OLs.
Proposed § 53.1366 would address
requirements for the general content of
applications for OLs. It would refer to
general content requirements in
proposed § 53.1109 and would require
supplemental information. Proposed
§ 53.1366(a) is equivalent to § 50.33(f).
Proposed § 53.1366(b) is equivalent to
§ 50.33(k).
Proposed § 53.1369 would provide
requirements for the technical content
of applications for OLs to be included
in the FSAR and is equivalent to
§ 50.34(b) but has been modified to align
with the technical requirements in part
53. It would require that the FSAR
include and, as needed, update
information provided in the PSAR that
was submitted and reviewed to support
the associated CP application.
Similar to the proposed requirements
for the content of CP applications,
proposed § 53.1369(a) would reference
the requirements for the content of an
ESP application to address application
requirements related to the site. Section
53.1369(b) would reference the
requirements for the content of a DC
application to address some of the
application requirements related to
design of the commercial nuclear plant.
Proposed § 53.1369(c) is equivalent to
§ 50.34(b)(7). Proposed § 53.1369(d)
would require a description of the
Integrity Assessment Program that
would be required by proposed
§ 53.870. Proposed § 53.1369(e) is
equivalent to § 50.34(e). Proposed
§ 53.1369(g) would provide
requirements for OL application content
to support proposed § 53.730 related to
the role of personnel in the operation of
the commercial nuclear plant and is
adapted from requirements in part 55
and § 50.34(f). Likewise, proposed
§ 53.1369(h) would provide
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requirements for OL application content
related to training programs to support
proposed §§ 53.730(g) and 53.830 and
includes requirements equivalent to
§ 50.34(b)(8), § 52.79(a)(33), and part 55.
Proposed § 53.1369(i) would provide
requirements for OL application content
related to emergency plans to support
proposed § 53.855 and is equivalent to
§ 50.34(b)(6)(v).
Proposed § 53.1369(j) would provide
requirements for OL application content
related to the applicant’s organizational
structure and is equivalent to
§ 50.34(b)(6)(i). Proposed § 53.1369(k)
would provide requirements for OL
application content related to the
applicant’s proposed maintenance
program to support proposed § 53.715
and is equivalent to § 50.34(b)(6)(iv).
Proposed § 53.1369(l) would provide
requirements for OL application content
related to the applicant’s quality
assurance program to support proposed
§ 53.865 and is equivalent to
§ 50.34(b)(6)(ii). Proposed § 53.1369(m)
would provide requirements for OL
application content related to the
applicant’s proposed radiation
protection program to support proposed
§ 53.850 and is equivalent to
§ 50.34(b)(3).
Proposed § 53.1369(n) through (p)
would provide requirements for OL
application content related to the
applicant’s proposed physical security
program to support proposed § 53.860(a)
and are equivalent to § 50.34(c) and (d).
Proposed § 53.1369(q) would provide
requirements for OL application content
related to the applicant’s proposed
cybersecurity plan to support proposed
§ 53.860(d) and is equivalent to
§§ 52.79(a)(36)(iv) and 73.54. Proposed
§ 53.1369(r) would provide
requirements for OL application content
related to the implementation of
proposed security, safeguards, and
cybersecurity plans to support proposed
§ 53.860 and is equivalent to
§ 52.79(a)(35)(ii) and 52.79(a)(36)(iv)
and (v).
Proposed § 53.1369(s) would provide
requirements for OL application content
related to the applicant’s proposed fire
protection program to support proposed
§ 53.875 and is equivalent to
§ 52.79(a)(40). Proposed § 53.1369(t)
would provide requirements for OL
application content related to the
applicant’s proposed ISI and IST
program to support proposed § 53.880
and is equivalent to part of
§ 52.79(a)(11). Proposed § 53.1369(w)
would provide requirements for OL
application content related to the
applicant’s general employee training
program to support proposed § 53.830
and is equivalent to § 52.79(a)(33).
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Proposed § 53.1369(x) would provide
requirements for OL application content
related to the applicant’s FFD program
to support part 26 and is equivalent to
§ 52.79(a)(44). Proposed § 53.1369(y)
would provide requirements for OL
applicant’s programs to demonstrate
that any safety questions identified at
the CP stage have been resolved and is
equivalent to § 50.34(b)(5). Proposed
§ 53.1369(z) would provide
requirements for OL applicants to
describe how the performance of each
safety design feature has been
demonstrated capable of fulfilling
functional design criteria considering
interdependent effects through either
analysis, appropriate test programs,
prototype testing, operating experience,
or a combination thereof to support
proposed § 53.440(a). It is largely
equivalent to §§ 50.34(b)(5) and
50.43(e). Proposed § 53.1369(aa) would
provide requirements for OL application
content related to the applicant’s
proposed TS to support proposed
§ 53.710(a) and is equivalent to
§ 50.34(b)(6)(vi).
Proposed § 53.1372 would address
requirements for the content of OL
applications other than the FSAR.
Proposed § 53.1372(a) would require
submission of an environmental report
and is equivalent to § 50.30(f) and
§ 51.53(b). Proposed § 53.1372(b) does
not have a direct parallel in parts 50 and
52 and would require the inclusion of
a description of availability controls that
are not included in the FSAR to support
proposed § 53.710(b).
Proposed § 53.1375 would address
standards for review of OL applications
and the administrative review of
applications, including hearings, and is
equivalent to §§ 52.81 and 52.85, except
that the NRC has omitted 10 CFR part
54, ‘‘Requirements for Renewal of
Operating Licenses for Nuclear Power
Plants,’’ from the list of standards in the
proposed § 53.1375(a). Proposed part 53
does not include detailed requirements
related to renewal of licenses, although
a general provision and possible
placeholder for future requirements has
been included as proposed § 53.1595.
The NRC will decide after the part 53
final rule is published whether this
future section will be retained in part 53
to address license renewal or whether
the agency will take another approach to
address license renewal for part 53
licensees, such as amending part 54 to
address part 53 licensees.
Proposed § 53.1381 would address
referral to the ACRS and is equivalent
to §§ 50.58 and 52.87. Proposed
§ 53.1384 would address exemptions,
departures, and variances for OL
applicants. Section 53.1384(a) is
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86947
equivalent to § 52.93 but is adapted for
OLs. Proposed § 53.1384(b) is equivalent
to §§ 52.39(d) (with respect to ESPs) and
52.93 but is adapted for OLs.
Proposed § 53.1387 would address
issuance of OLs. The proposed
introductory paragraph is equivalent to
§ 50.56. Proposed § 53.1387(a)(1)(i) is
equivalent to §§ 50.50 and 50.57(a)(1).
Proposed § 53.1387(a)(1)(ii) is
equivalent to § 50.50. Proposed
§ 53.1387(a)(1)(iii) is equivalent to
§ 50.57(a)(2). Section 53.1387(a)(1)(iv) is
equivalent to § 50.57(a)(3). Proposed
§ 53.1387(a)(1)(v) is equivalent to
§ 50.57(a)(4). Proposed
§ 53.1387(a)(1)(vi) is equivalent to
§ 50.57(a)(6). Proposed
§ 53.1387(a)(1)(vii) is equivalent to
§ 50.57(a)(5). Proposed
§ 53.1387(a)(1)(viii) is equivalent to
§ 52.97(a)(1)(vi) but is adapted for OLs.
Proposed § 53.1387(c) is equivalent to
§ 50.57(b). Proposed § 53.1387(d) is
equivalent to §§ 50.36(b) and 50.50.
Proposed § 53.1390 would address
backfitting of OLs and is equivalent to
§ 52.98(a) but adapted for an OL
application. Proposed § 53.1396 would
address duration of an OL and is
equivalent to § 50.51(a) and § 52.104.
Proposed § 53.1399 would address
transfer, assignment, and other
disposition of an OL and is equivalent
to § 50.80. Proposed § 53.1402 would
address applications for renewal of an
OL and refers to proposed § 53.1595.
Proposed § 53.1405 would address
continuation of an OL and is equivalent
to § 52.109 but is adapted to address an
OL.
Proposed §§ 53.1410 through 53.1461
would address requirements for COLs.
Proposed § 53.1410 is equivalent to
§ 52.71. Proposed § 53.1413 would
address general information
requirements for the content of
applications for COLs and is equivalent
to § 52.77, which references § 50.33.
Most of the provisions from § 50.33 are
restated in proposed § 53.1109. Some
requirements in § 50.33 related to
financial qualifications and construction
timelines are addressed in other
sections of part 53.
Proposed § 53.1416 would address the
technical content to be included in an
FSAR for an application for a COL and
is equivalent to § 52.79 except as
modified to reflect the technical
requirements in part 53 and with one
addition. Proposed § 53.1416 includes
the statement that the Commission will
require, before issuance of a COL, that
engineering documents, such as
analyses, drawings, procurement
specifications, or construction and
installation specifications, be completed
and available for audit if the more
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detailed information is necessary for the
Commission to verify the information in
the application and make its safety
determination. This statement is
equivalent to DC application
requirements in § 52.47 and is included
in proposed § 53.1416 for clarity.
Similar to the proposed requirements
for the content of OL applications,
proposed § 53.1416(a)(1) would
reference the requirements for the
content of an ESP application to address
application requirements related to
siting. Section 53.1416(a)(2) would
reference the requirements for the
content of a DC application to address
some of the application requirements
related to design of the commercial
nuclear plant. The remaining items
under proposed § 53.1416(a) are
likewise similar to the required content
for OL applications under proposed
§ 53.1369(a). Proposed § 53.1416(b)
would require COL applicants to
provide a report documenting the
resolution of any safety questions for
SSCs for which research and
development was necessary to confirm
the adequacy of their design and is
equivalent to § 50.34(b)(5). Proposed
§ 53.1416(c) would provide
requirements for COL applicants to
describe how the performance of each
safety design feature has been
demonstrated capable of fulfilling
functional design criteria considering
interdependent effects through either
analysis, appropriate test programs,
prototype testing, operating experience,
or a combination thereof to support
proposed § 53.440(a). It is largely
equivalent to §§ 52.79(a)(24) and
50.43(e). Proposed § 53.1416(d) would
address the content of COL applications
referencing an ESP. Proposed
§ 53.1416(e) would address the content
of COL applications referencing a
standard design approval. Proposed
§ 53.1416(f) would address the content
of COL applications referencing a
standard DC. Proposed § 53.1416(g)
would address the content of COL
applications referencing an ML.
Proposed § 53.1419 would address
other application content for COLs and
is equivalent to § 52.80. Proposed
§ 53.1419(a)(2) is new and would
require the inclusion of a description of
availability controls that are not
required to be included in the FSAR.
Proposed § 53.1422 would address
standards for review of applications and
the administrative review of
applications, including hearings, and is
equivalent to §§ 52.81 and 52.85. The
NRC has removed part 54 from the list
of standards in proposed § 53.1422(a).
Proposed part 53 does not include
requirements related to renewal of
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licenses, in relation to proposed
§§ 53.1422 and 53.1595.
Proposed § 53.1425 would address the
finality of NRC approvals referenced in
a COL application and is equivalent to
§ 52.83(a). Proposed § 53.1431 would
address the referral of COL applications
to the ACRS for review and is
equivalent to § 52.87. Proposed
§ 53.1434 would address the
authorization to conduct LWA activities
and is equivalent to § 52.91. Proposed
§ 53.1437 would address exemptions,
departures, and variances and is
equivalent to § 52.93. Proposed
§ 53.1440 would address issuance of
COLs and is equivalent to § 52.97.
Proposed § 53.1443 would address
finality of COLs and is equivalent to
§ 52.98.
Proposed § 53.1449 would address
inspection during construction and is
equivalent to § 52.99. Proposed
§ 53.1452 would address operation
under a COL and is equivalent to
§ 52.103. Paragraph (a) of proposed
§ 53.1452 would include footnotes to
provide that, for licensees installing
fueled manufactured reactors under a
COL, (1) the COL holder would notify
the NRC of its scheduled date for
initiating the physical removal of any
one of the independent physical
mechanisms to prevent criticality
required under § 53.620(d)(1) rather
than its scheduled date for the initial
loading of fuel, and (2) the NRC would
time its publication of the notice of
intended operation based on the COL
holder’s schedule for initiating the
physical removal of any one of the
independent physical mechanisms to
prevent criticality required under
§ 53.620(d)(1) rather than the COL
holder’s scheduled date for the initial
loading of fuel. These footnotes are
consistent with the provisions of
proposed § 53.620(d)(1)(iv), which
would state that, upon initiating the
physical removal of any one of the
independent physical mechanisms to
prevent criticality in the manufactured
reactor’s place of operation, the fueled
manufactured reactor has commenced
operation. For reactors without the
independent physical mechanisms to
preclude criticality under proposed
§ 53.620(d)(1), operation begins with
initial fuel load. In both cases, removal
of the physical features to prevent
criticality (for reactors with such
features) and initial fuel load (for
reactors without such features) put a
fully constructed utilization facility in a
position to sustain a nuclear chain
reaction, and in both cases, the
utilization facility cannot sustain a
nuclear chain reaction (for lack of
sufficient reactivity) until the action
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takes place. Therefore, the NRC
proposes that initiating the physical
removal of any one of the independent
physical mechanisms to prevent
criticality is the best analogue to initial
loading of fuel for reactors without such
features.
The proposed footnote in § 53.1452(a)
regarding timing of the notice of
intended operation for fueled
manufactured reactors with
independent physical mechanisms to
prevent criticality also addresses the
requirements of Section 189a.(1)(B)(i) of
the Act. This section requires, in part,
that ‘‘[n]ot less than 180 days before the
date scheduled for initial loading of fuel
into a plant by a licensee that has been
issued a combined construction permit
and operating license under section
185b., the Commission shall publish in
the Federal Register notice of intended
operation.’’ That section further requires
that this notice provide a 60-day period
in which to request a hearing ‘‘on
whether the facility as constructed
complies, or on completion will
comply, with the acceptance criteria of
the license.’’ In the case where a fueled
manufactured reactor arrives at the site
where it is to be operated by a COL
holder, the manufacturer would have
loaded fuel at the factory under its part
70 license. Therefore, at the site of
operation, there would not be ‘‘initial
loading of fuel into a plant by a licensee
that has been issued a combined
construction permit and operating
license’’ (emphasis added). Under a
literal reading of the entry condition in
Act Section 189a.(1)(B)(i), this situation
would not trigger its requirements.
However, the purpose of the provision
is to offer the hearing opportunity at
least 180 days prior to when the fuel is
loaded and ready for use at its
authorized location. It would be
contrary to that purpose if, in this
situation, the Commission did not
publish the notice of intended operation
and opportunity for the public to
request a hearing on conformance with
the acceptance criteria in the COL for
the site of operation. To fulfill the
underlying purpose of the law, the NRC
proposes to time the notice of intended
operation based on the COL holder’s
schedule for initiating the physical
removal of any one of the independent
physical mechanisms to prevent
criticality required under § 53.620(d)(1).
This action by the COL holder would be
the best analogue to initial fuel load by
the COL holder for the reasons stated
previously. This analogue is adopted in
other sections of the proposed part 53
and related sections in parts 50 and 73
that use initial fuel loading to identify
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a transition point for the applicability of
regulatory requirements. To address the
possible loading of fuel into a
manufactured reactor for subsequent
transport to and use at a commercial
nuclear plant, multiple sections that
determine the applicability of
regulations have been drafted or revised
to allow for either initial fuel load or
initiating the physical removal of any
one of the independent physical
mechanisms to prevent criticality
required under § 53.620(d)(1) for a
fueled manufactured reactor to
determine the applicability of the
requirement, as appropriate.
Proposed § 53.1455 would address
duration of COL and is equivalent to
§ 52.104. Proposed § 53.1456 would
address the transfer of a COL and is
equivalent to § 52.105. Proposed
§ 53.1458 would address application for
renewal and is equivalent to § 52.107.
Proposed § 53.1461 would address
continuation of COL and is equivalent
to § 52.109.
Proposed § 53.1470 would address
standardization of commercial nuclear
plant designs and licenses to construct
and operate commercial power reactors
of identical design at multiple sites and
is equivalent to appendix N of part 52.
This section would set out the particular
requirements and provisions applicable
to situations in which applications for
CPs and subsequent OLs, or COLs,
under this part are filed by one or more
applicants for licenses to construct and
operate nuclear power reactors of
identical design (‘‘common design’’) to
be located at multiple sites. Additional
information related to this proposed
section is provided in the final rule to
revise part 52 (72 FR 49352; August 28,
2007).
Subpart I—Maintaining and Revising
Licensing-Basis Information
Part 53 would establish requirements
for the maintenance of licensing-basis
information in subpart I.
Section 53.1500 would describe the
purpose of the subpart in terms of the
definition of licensing-basis information
in subpart A. Subpart I would be closely
tied to the requirements in subpart H,
which would provide the requirements
for contents of applications for the
various types of licenses issued under
part 53. Subpart I would generally be
organized into sections dealing with: (1)
licensing-basis information that
licensees are not authorized to change
without NRC approval (e.g., licenses,
regulations); and (2) licensing-basis
documents that licensees may change
provided specified criteria are satisfied
(e.g., FSAR, program descriptions). The
subpart would also capture certain
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general conditions on licenses and
changes to the licenses related to the
transfer and termination of licenses.
Section 53.1502 would define specific
terms and conditions of licenses. These
terms and conditions would be
equivalent to the regulations in: (1)
§ 50.54(h) stating that each license is
subject to the provisions of the Act and
requirements issued by the Commission;
(2) § 50.54(s) stating the actions the
Commission would take if it makes a
finding that there is not reasonable
assurance that adequate protective
measures can and will be taken in the
event of a radiological emergency; (3)
§ 50.54(aa) stating that each license is
subject to the specified sections of the
Federal Water Pollution Control Act;
and (4) § 50.54(dd) stating that a holder
of an OL or COL may take reasonable
actions that depart from the license in
a national security emergency.
Section 53.1505(a) would serve as an
introduction to and overview of the
sections that follow on changes to
licensing-basis information requiring
prior NRC approval, namely the
elements of licensing-basis information
defined by licenses, orders, and
regulations. The related sections within
these subparts would primarily deal
with the process of how a licensee
requests and the NRC issues an
amendment to a license or issues an
order that modifies a license. Another
important element of licensing-basis
information that a part 53 licensee
would not be able to change or deviate
from without NRC approval would be
the NRC regulations themselves. Section
53.1505(b) would refer to § 53.080 in
subpart A that would provide the
criteria for a licensee or other party to
satisfy when requesting an exemption
from NRC regulations.
Section 53.1510 would be equivalent
to § 50.90 and would require that a
licensee submit an application to
request an amendment to a license. The
required assessments that would be
included within an application to
amend a license under part 53 would
need to address the safety criteria and
analysis requirements of subparts B and
C. As with parts 50 and 52, licensees
would be required to include in their
applications to amend a license an
analysis of whether the amendment
involves no significant hazards
consideration using the standards in
§ 53.1520, which would be equivalent to
the standards in § 50.92. Although this
rulemaking provided an opportunity to
revise the terminology related to no
significant hazards consideration
determinations, which dates to the early
1960s when applications were
supported by final hazard summary
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reports, the NRC is proposing to
maintain the same terminology used in
part 50 to minimize the need for
associated changes in other regulations,
guidance, and public notices.
Section 53.1515 would establish
requirements for public notices and
state consultations associated with the
NRC’s processing of a license
amendment request. This section would
be equivalent to § 50.91 for the NRC’s
processes related to applications to
amend an OL or COL. Section 50.91(b)
stipulates that the Commission will
make available to the licensee the name
of the appropriate State official
designated to receive such amendments.
While the Commission intends to
continue following this practice, the
Commission has not included this
administrative matter in proposed part
53. Proposed § 53.1515(b)(3) contains
some modifications compared to
§ 50.91(b)(3) for clarity; these revisions
are not intended to revise the substance
of the provisions in part 53 compared to
part 50.
Section 53.1520 would be based on
§ 50.92. The section would continue to
use the criteria in § 50.92 for
determining that a proposed
amendment involves no significant
hazards consideration. Although more
specific terms such as event sequence
are used throughout part 53, § 53.1520
would use the term ‘‘accident’’ to
maintain consistency with the long
history of making no significant hazards
consideration determinations under part
50.
Section 53.1525 would provide
requirements for holders of an OL or
COL requesting to revise information
from a DC rule that was referenced in
the initial license application and
included in or incorporated by reference
into the facility FSAR. In keeping with
the current requirements in part 52, the
portion of the part 53 facility licensingbasis information obtained from the
certified design would be divided into
two categories. The most significant
design information and the ITAAC
would be certified by rule and
designated as ‘‘certification
information.’’ The remaining
information, which makes up the
majority of the design information
approved as part of the DC, would not
be certified by rule and is not
considered ‘‘certification information.’’
Part 52 refers to these categories of
information as Tier 1 and Tier 2
information, respectively, and refers to
a change made to that information on a
plant-specific basis as a departure.
Under part 52, a departure from Tier 1
information requires an exemption and,
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for information incorporated into the
license, a license amendment.
Part 53 would dispense with the Tier
1 and Tier 2 terminology. Rather,
§ 53.1525 would use the term
‘‘certification information’’ in place of
Tier 1, and a plant-specific departure
from the certification information would
require both a request for an exemption
from the associated DC rule and, for
information such as ITAAC
incorporated into the license, a license
amendment. However, as would be
provided in § 53.1525(c), a plantspecific departure from the information
approved by the NRC as part of the DC
rule but which is not certification
information (i.e., Tier 2 information
under part 52) would be assessed using
the process and criteria defined in
§ 53.1550 for changes to a FSAR. An
applicant or licensee would need to
identify such a change as a departure
from the referenced standard design in
the updated FSAR. The process for
making a generic change to a certified
design would be described in the
associated section in subpart H.
Section 53.1530 would not allow the
holder of an ML or the holder of a COL
using a manufactured reactor to make
changes to the design of the
manufactured reactor without
requesting a license amendment from
the NRC. This section would provide
the equivalent requirements as those in
§§ 52.98 and 52.171.
Section 53.1535 would establish
requirements for license amendments
during construction. The section would
provide the equivalent options and
requirements for the holders of a CP as
those in § 50.35(b). The regulations
would allow but do not require the
holder of a CP or LWA to request an
amendment under § 53.1510 if the
licensee desires to obtain NRC approval
of a specific design feature or
specification. The requirements for
obtaining an amendment to a COL to
address changes during construction
would also be provided in § 53.1535.
The proposed process would differ from
the current requirements in part 52 by
adopting a requirement that would
explicitly support a change process like
that described in RG 1.237, ‘‘Guidance
for Changes During Construction for
New Nuclear Power Plants Being
Constructed Under a Combined License
Referencing a Certified Design Under 10
CFR part 52.’’
The proposed regulation would allow
the holder of a COL to proceed at its
own risk in making a change during the
construction process and would require
that licensee to submit a license
amendment request no later than 45
days from the date the licensee begins
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to implement the change or departure
requiring NRC approval.
Section 53.1540 would serve as an
introduction to the sections that follow
on changes to licensing-basis
information that are primarily under the
control of a licensee but for which
evaluations are made to determine if a
submittal to the NRC requesting
approval would be required. The section
would also include definitions that
would be applicable when using the
processes in §§ 53.1545 through
53.1565. The definitions would be
largely equivalent to those in § 50.59(a)
but include some revision to reflect the
structure and terminology in other
subparts in part 53. For example, the
definition of ‘‘Change’’ in § 53.1540(b)
would address a ‘‘design feature or
related functional design criteria’’ rather
than a ‘‘design function,’’ because the
former are defined terms in part 53.
Similarly, in § 53.1540(b), the phrase
‘‘design basis’’ from § 50.59(a)(2) would
be replaced with functional design
criteria for SR SSCs.
Section 53.1545 would provide the
proposed requirements for updating of
FSARs. While the process-related
requirements proposed under § 53.1545
would be largely the same as those in
§ 50.71, the specifics of information to
be updated would differ due to the role
of PRA in satisfying the requirements in
subparts B and C. Additionally, the use
of the risk-informed approach in subpart
C would result in some but not all PRA
information being in the FSAR or
another licensing basis document and
therefore a separate PRA update
requirement similar to § 50.71(h) is not
included in proposed subpart I.
Proposed § 53.1239(a)(18) in subpart
H and the related references to this
proposed requirement for the holders of
OLs and COLs would require a
description of the PRA required by
§ 53.450(a) and its results to be included
in FSARs. However, guidance
documents are planned to clarify the
division of PRA-related information that
would need to be in the FSAR, in other
possible licensing basis documents, and
controlled as plant records subject to
inspections and audits. At a minimum,
the information from the PRA that
would be needed to show compliance
with subpart C would be included in the
FSAR (e.g., PRA summary and
analytical results for LBEs). The
submittal of voluminous PRA
information was initially required under
part 52, but that proved to be
impractical and was revised in the 2007
revision of part 52. Guidance is being
developed to ensure sufficient
information is submitted to the NRC to
support the licensing process and the
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NRC’s regulatory findings under part 53
or similar applications using the LMP
under parts 50 or 52.
The NRC has posed a question in
section VI, ‘‘Specific Requests for
Comments,’’ of this document that asks
about the appropriate level of detail for
PRA-related information in an FSAR
and whether other licensing basis
documents might be more appropriate
to both provide information to the NRC
and ensure the PRA is maintained and
updated as proposed in subpart C. The
program document would provide more
detail than the summaries in the FSAR
but still be a much-condensed source of
information in comparison to the
documentation of the PRA.
Section 53.1545(a)(3) and (4) would
be based on the inclusion of at least a
summary of PRA results and the related
margins to safety criteria in the FSAR
and would require updates to that
information. The routine reporting of
these margins would also inform
application of the criteria for allowing
changes without an amendment in the
following section (§ 53.1550) in subpart
I.
Section 53.1550 would establish
requirements for evaluating changes to a
facility as described in its FSAR. This
proposed section would provide the
equivalent of the requirements in
§ 50.59 for evaluating changes to an
FSAR (as updated) and determining if a
license amendment is required to
implement a change to a facility or
procedures. The evaluation criteria
proposed in § 53.1550 would reflect the
role of the PRA in the safety analyses
under part 53 and would include
several measures related to the changes
in plant risk resulting from a change in
the plant design or plant procedures.
Examples would include criteria that
rely on the identification of risksignificant event sequences in
accordance with the analysis
requirements of § 53.450; exceeding the
LBE evaluation criteria as defined in
§ 53.450; the consideration of potential
reductions in margin between the
estimated comprehensive risk metrics
and associated risk performance
objectives in the safety criteria in
§ 53.220; changes to the safety
classification of SSCs; and consideration
of reductions in defense in depth.
Section 53.1550 would include a
criterion related to a departure from a
method of evaluation used in the safety
analyses. The NRC has not yet
developed draft guidance for use in
applying proposed § 53.1550 but
anticipates that the NRC and
stakeholders will assess the potential
need for such guidance and that such
guidance would, if needed, be
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developed as part of ongoing or future
activities.
Section 53.1550 would include
certain concepts taken from existing
guidance for § 50.59 in the proposed
criteria related to DBAs. Specifically,
criterion (iv) for changes made to a
method of evaluation of DBAs under
§ 53.450(f) would be equivalent to a
change in a method of evaluation under
§ 50.59, and criterion (viii) on assessing
if a change creates a possibility for an
accident of a different type than
previously analyzed in the FSAR would
be similar to the § 50.59 criterion (v).
Guidance documents will be prepared
to address the content of applications
for PRA-related information under
proposed part 53, and this guidance will
also influence how potential changes in
the evaluation of LBEs other than DBAs
analyzed under § 53.450(e) are
evaluated and reported under the
proposed criterion (iv).
Section 53.1550(a)(2)(x) would
require evaluating plant changes to
ensure they would not prevent
satisfying the design requirements in
§ 53.440(j) related to the impact of a
large commercial aircraft. The inclusion
of a proposed requirement under
§ 53.1550 related to design features for
protecting against aircraft impact would
reflect the proposed design requirement
in subpart C and related proposed
requirements in subpart H to address
the proposed design requirement in
FSARs.
Sections 53.1560 through 53.1565 in
subpart I would define the processes for
a licensee to evaluate changes to the
program documents included in the
licensing-basis information submitted to
the NRC and to modify such programs
without NRC prior approval.
Section 53.1560 would include the
proposed requirements for updating
program documents included in
licensing-basis information and would
provide the equivalent of FSAR updates
for key program documents. The
proposed requirements in these sections
would provide a uniform approach for
updating program documents, which
correspond to the programs required
under subpart F.
The proposed § 53.1565 would
provide a process for licensees to make
changes to program documents included
in licensing-basis information without
obtaining prior NRC approval. The
proposed requirements would include
several generic criteria that, if not
satisfied, would prompt the need for
NRC approval of a change to a program
document. These generic criteria would
include whether a change would
comply with TS and NRC regulations.
Another proposed criterion for
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evaluating changes to program
documents would be conforming with
program-specific requirements,
including NRC-approved program
documents with more specific criteria
for a particular program, regulations,
administrative controls sections of TS,
and NRC-approved program documents.
Proposed § 53.1565(d) would include
specific criteria for evaluating changes
to several program documents that have
well established change processes and
guidance for licensees under parts 50
and 52. The program documents
specifically addressed in the proposed
section would include quality assurance
programs that would be equivalent to
§ 50.54(a), an emergency preparedness
program that would be equivalent to
§ 50.54(q), and the security program that
would be equivalent to § 50.54(p).
The proposed § 53.1570 would
establish requirements for the transfer of
commercial nuclear plant licenses by
providing the equivalent requirements
of § 50.80 for the possible transfer of an
ESP, CP, OL, or COL. Likewise, the
proposed § 53.1575 would establish
requirements for the termination of an
OL or COL by providing the equivalent
requirements of § 50.82. Other proposed
requirements related to
decommissioning and license
termination would be included in
subpart G.
Section 53.1580 would establish
requirements for information requests
the NRC could send to the various types
of licensees and would provide
requirements that would be equivalent
to requirements in § 50.54(f). The
proposed § 53.1585 would provide the
requirements that would be equivalent
to requirements in § 50.100 to address
revocation, suspension, modification of
licenses, and approvals for cause.
Section 53.1590 would propose to
address backfitting requirements by
providing requirements that would be
equivalent to those in § 50.109.
Proposed § 53.1595 would address
license renewals under part 53 with
simple statements that licenses may be
renewed. This section would be
expanded through future rulemakings to
more fully describe or reference the
processes related to requesting and
processing applications to renew ESPs,
OLs, and COLs issued under part 53 (if
finalized).
Subpart J—Reporting and Other
Administrative Requirements
Part 53 would address various
reporting and administrative
requirements in subpart J.
Section 53.1600 would explain the
organization of the various sections
within the subpart related to providing
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unfettered access to NRC inspectors;
maintaining certain records and
reporting specified events or conditions;
demonstrating compliance with
financial qualification requirements and
providing specified financial reports;
and maintaining financial protections to
address potential accidents.
Section 53.1610 would establish
requirements for the provision of
facilities and unfettered access for
inspections. These requirements would
be equivalent to § 50.70 with only minor
changes proposed to provide additional
flexibilities and address possible
differences related to reactors licensed
under part 53 and the possibility that
some commercial nuclear plants may
not be assigned resident inspectors.
Section 53.1620 would provide for
maintenance of records and the making
of various reports to the NRC. These
requirements would be largely
equivalent to § 50.71. This section is not
intended to reflect all provisions in
§ 50.71; several important requirements
in § 50.71 would be captured in other
sections of part 53. For example,
§ 53.1545 within subpart I would
provide requirements that would be
equivalent to § 50.71(e), updating
FSARs, and § 53.1680, ‘‘Annual
financial reports,’’ would provide the
equivalent of § 50.71(b), which covers
financial reports. A reporting
requirement related to completion of
power ascension testing would be added
to § 53.1620 to support the assessment
of annual fees under 10 CFR part 171,
‘‘Annual Fees for Reactor Licenses and
Materials Licenses, Including Holders of
Certificates of Compliance,
Registrations, and Quality Assurance
Program Approvals and Government
Agencies Licensed by the NRC,’’ which
normally commence upon completion
of those testing activities.
Section 53.1630 would establish
requirements for immediate notification
requirements for operating commercial
nuclear plants. These requirements
would be equivalent to § 50.72 with
minor changes proposed to make the
reporting criteria technology inclusive.
In addition, a new version of NRC Form
361 (NRC Form 361S) would be created
for use by part 53 licensees, but without
LWR-specific terminology to ensure
technology inclusiveness. A separate
rulemaking activity, ‘‘Reporting
Requirements for Nonemergency Events
at Nuclear Power Plants,’’ has been
initiated to consider possible changes to
the requirements in § 50.72. At a future
date, the NRC may consider reconciling
future changes to § 50.72 with the
requirements proposed in part 53,
which have been taken or derived from
the current reporting requirements.
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Section 53.1640 would address the
licensee event report system. These
requirements would be equivalent to
§ 50.73 with minor changes proposed to
make the requirements inclusive of
various reactor technologies and to
reflect appropriate internal references to
other sections in part 53. In addition,
NRC Forms 366, 366A, and 366B would
be revised to include corresponding
check boxes for part 53 licensees.
Section 53.1645 would require
periodic reporting of the quantity of
radionuclides released to unrestricted
areas in liquid and gaseous effluents,
doses to members of the public, and the
results of environmental monitoring.
These reporting requirements in the
proposed part 53 would be largely
equivalent to those in the TSs required
by § 50.36a, ‘‘Technical specifications
on effluents from nuclear power
reactors.’’ The only difference would be
that a § 50.36a requirement to
specifically address conditions where
the dose to the maximally exposed
individual could be significantly above
design objectives would refer to a design
objective of 10 mrem/year total effective
dose equivalent, instead of referring to
the design objectives in appendix I to
part 50. The proposed section would
also include an equivalent to the
reporting requirement in section IV of
appendix I to part 50 if the radiation
exposure to a member of the public in
any calendar quarter exceeds one-half of
the annual ALARA design objective.
Section 53.1650 would include a
reporting requirement to support
safeguards agreements between the
United States and the International
Atomic Energy Agency (IAEA) and
would be equivalent to § 50.78.
Section 53.1660 through 53.1700
would address financial requirements
and would be largely similar to existing
regulations in parts 50 and 52. Section
53.1670 would be entitled ‘‘Financial
qualifications’’ and would require
applicants other than electric utilities to
possess or have reasonable assurance of
obtaining funds for the activities for
which the license is being sought. The
NRC is seeking feedback on these
sections and their ramifications for
merchant plants 3 in section VI,
‘‘Specific Requests for Comments,’’ of
this document. The remaining financial
reports in part 53 would be equivalent
to § 50.71(b) for annual financial
reports, § 50.76 for a change of status,
§ 50.54(cc) for the filing of a petition for
3 A ‘‘merchant plant’’ is a plant licensed to a nonrate-regulated entity (e.g., a nonutility) that engages
in the business of production, manufacturing,
generating, buying, aggregating, marketing, or
brokering electricity for sale at wholesale or for
retail sale to the public.
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bankruptcy, and § 50.81 for creditor
regulations.
Sections 53.1710 through 53.1730
would address financial protection
requirements. Section 53.1720 would
require insurance to stabilize and
decontaminate a plant following an
accident. These requirements would be
taken from § 50.54(w) with the only
notable change being the addition of a
provision allowing plant-specific
estimates of costs to stabilize and
decontaminate a plant as an alternative
to the $1.06 billion minimum coverage
in § 50.54(w). Section 53.1730 is
equivalent to § 50.57(a)(5) and would
refer to the requirements in 10 CFR part
140, ‘‘Financial Protection
Requirements and Indemnity,’’ related
to financial protection requirements and
indemnity agreements, including the
financial protection requirements of the
Price-Anderson Act.
The NRC is proposing a technologyinclusive, risk-informed, and
performance-based approach for the
application of drug and alcohol testing
and fatigue management requirements
for facilities licensed under part 53. The
proposed requirements applicable to
these applicants, licensees, and other
entities would be commensurate with
the radiological consequences presented
by the applicants’ facilities and the
operation of these facilities.4 The
proposed FFD framework would consist
of a two-tiered graded approach similar
to that currently in part 26 and an
optional third tier for part 53
commercial nuclear plants that perform
an analysis that demonstrates the
facility and its operation would satisfy
the criterion in proposed § 26.603(c),
which refers to § 53.860(a). This
proposed FFD framework would be
established in subpart M, ‘‘Fitness for
Duty Programs for Facilities Licensed
Under Part 53,’’ of part 26.
The NRC used operating experience to
provide regulatory flexibility in the
proposed subpart M of part 26
framework to help support a licensee’s
or other entity’s response to changes in
societal drug use, drug testing
technologies and processes, and FFD
program performance. The flexibility
would also help in FFD program
implementation because of the wide
variety of staff sizes anticipated at
commercial nuclear plants licensed
under part 53 and the geographically
remote locations in which commercial
nuclear plants may be sited.
The proposed first-tier FFD program
requirements would apply to part 53
licensees and other entities of
commercial nuclear plants under
construction who satisfy the criterion in
§ 26.603(c) but elect not to implement
proposed § 26.604, ‘‘FFD program
requirements for facilities that satisfy
the § 26.603(c) criterion,’’ or who do not
satisfy the criterion in § 26.603(c), and
to holders of MLs who are assembling
or testing manufactured reactors. These
requirements would be provided in
proposed § 26.605(a) and would be
essentially equivalent to those
requirements in subpart K, ‘‘FFD
Program for Construction,’’ of part 26 as
supplemented by select requirements
from subparts E, ‘‘Collecting Specimens
for Testing,’’ and I, ‘‘Managing Fatigue,’’
of part 26, and the requirements in
subparts A, ‘‘Administrative
Provisions,’’ and O, ‘‘Inspection,
Violations, and Penalties,’’ of part 26.
The first-tier requirements would
involve policies, procedures, behavioral
observation, fatigue management, drug
and alcohol testing, determinations of
fitness, appeals, training, sanctions,
auditing, change control, performance
monitoring, recordkeeping, and
reporting. These requirements would
help deter individuals subject to this
section from illicit drug and/or alcohol
use and from being impaired from any
cause including fatigue. These proposed
requirements would also help licensees
4 The NRC uses the term ‘‘operation’’ in its part
26 discussion to focus on human performance,
namely the necessity of individuals to operate,
maintain, surveil, and protect the facility and
respond to operational transients and unlikely
event sequences.
Subpart M—Enforcement
Subpart M would contain two
provisions, § 53.9000 and § 53.9010,
which are analogous to provisions
contained in other parts of 10 CFR
Chapter I imposing requirements on
regulated entities. Section 53.9000
would provide notice of the
Commission’s authority under the Act
to obtain injunctions or other court
orders for the enumerated violations.
Paragraph (a) of § 53.9010 would
provide notice to all persons and
entities subject to part 53 that they are
subject to criminal sanctions for willful
violations, attempted violations, or
conspiracy to violate certain regulations
under part 53. Criminal sanctions would
not apply to the regulations listed in
paragraph (b). The regulations for which
criminal penalties would apply are
limited to those that establish either a
regulatory obligation or prohibition.
V. Changes to Other Parts of 10 CFR
Chapter I
10 CFR Part 26
A. Introduction
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and other entities identify individuals
as users of impairing substances and
demonstrate compliance with § 26.23,
‘‘Performance objectives.’’
The proposed second tier would
include all the proposed first-tier
requirements, plus the more
comprehensive set of FFD program
requirements in current subparts C,
‘‘Granting and Maintaining
Authorization,’’ D, ‘‘Management
Actions and Sanctions to be Imposed,’’
H, ‘‘Determining Fitness-for-Duty Policy
Violations and Determining Fitness,’’
and N, ‘‘Recordkeeping and Reporting
Requirements,’’ of part 26. These
requirements would be provided in
proposed § 26.605(b) and would be
applicable to licensees and other
entities satisfying the § 26.603(c)
criterion, at their discretion. These
requirements would also apply to
licensees or other entities not satisfying
the § 26.603(c) criterion that implement
an FFD program under subpart M of part
26, before the loading of fuel onsite into
a reactor vessel; before receiving a
manufactured reactor; or before
operating, testing, performing
maintenance of, or directing the
maintenance or surveillance of securityrelated equipment or equipment that a
risk-informed evaluation process has
shown to be significant to public health
and safety.
The second-tier requirements are
based on the additional risk presented
by nuclear reactor assembly, testing,
fueling, and operation and the necessity
for human actions in certain event
sequences. The inclusion of the current
part 26 requirements would align
proposed part 53 FFD and AA program
requirements with the current FFD and
AA programs required for facilities
licensed under parts 50 and 52. This
approach would ensure effective and
consistent AA and FFD program
implementation across the commercial
nuclear power industry, thereby
ensuring uniform requirements for
individuals who may perform roles and
responsibilities for multiple facilities
regardless of facility licensure.
Proposed § 26.604 would offer an
alternate option for an applicant
implementing an FFD program under
subpart M of part 26. If the applicant
demonstrates that the criterion in
proposed § 26.603(c) is met, then the
applicant (and the subsequent licensee
or other entity) must still implement an
FFD program described in subpart M of
part 26; however, drug and alcohol
testing would not be required unless
FFD performance declines or the
applicant, licensee, or other entity elects
to implement drug and alcohol testing.
The proposed § 26.604 requirements are
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equivalent to those proposed in
§ 26.605(a) except for required drug and
alcohol testing. This proposed
framework would focus on the human
performance of individuals while they
are performing those duties and
responsibilities that make them subject
to the FFD program. This performance
would be verified through behavioral
observation, evaluation of any FFD
concerns, performance monitoring,
fatigue management, and
determinations of fitness. Applicants
that do not satisfy the criterion in
proposed § 26.603(c), or elect not to
perform the analysis required to
demonstrate that the criterion in
§ 26.603(c) is met, would be subject to
an FFD program described in § 26.605,
‘‘FFD program requirements for
facilities that do not implement
§ 26.604,’’ or an FFD program that
implements all part 26 requirements,
except for those requirements in
subparts K and M of part 26.
In establishing the minimum FFD
program requirements in § 26.604, the
NRC reviewed current advanced reactor
designs against that of a non-power
production or utilization facility (NPUF)
that is not required to implement an
FFD program for those individuals who
have unescorted access to the controlled
access area (and vital area for some
facilities), including NRC-licensed
operators.5 This review was performed
because commercial nuclear plants
licensed under part 53 could be
designed with similar power levels and
radiological consequences as the
currently licensed NPUFs. From this
review, three principal considerations
supported the minimum set of
requirements for the § 26.604 FFD
program.
First, the radiological consequences
presented by a part 53 licensed facility
and its operation that satisfy the
criterion in § 26.603(c) may present a
greater potential radiological
consequence to workers and the public
in the vicinity of the facility than does
an NPUF. Second, the operating
characteristics of a part 53 licensed
facility are unlike that of an NPUF
because there may be a higher reliance
on individuals at the part 53 site to
safely and competently operate,
maintain, surveil, and secure SSCs that
may not be required at an NPUF, such
as systems that provide secondary heat
transfer, reactor coolant flow, pressure
control, and at-power core refueling.
Differences in operating characteristics
could include, for example: long-term,
full power operation with automated
5 Controlled access area and vital area are defined
in § 73.2, ‘‘Definitions.’’
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reactivity control systems for loadfollowing; active and passive safety and
security systems; innovative non-lightwater heat transfer systems; and energy
storage and hazardous chemical
systems. The individuals at part 53
facilities may also be required to
communicate to individuals both onsite
and offsite, such as electrical load
dispatchers, any conditions adverse to
safety, security, or quality. Third, part
53 licensed facilities may be sited in
geographically remote locations that
may not have a physically available
administrative or corporate support
team to provide face-to-face oversight,
engineering expertise, and maintenance
support like that at NPUFs. This places
a higher reliance on those individuals
required at a part 53 facility being fit for
duty and trustworthy and reliable
because a replacement individual may
not be readily available.
The NRC proposes to exclude drug
and alcohol testing from the proposed
§ 26.604 framework for five reasons: (1)
the § 26.23 performance objectives can
be met through effective
implementation of the defense-in-depth
regulatory framework established by
behavioral observation, reporting of
legal actions, the proposed performance
monitoring and review program (PMRP),
FFD training, and requirements from the
physical protection, AA, cyber
protection, and licensed operator
programs; (2) the PMRP would require
the licensee or other entity to monitor
its FFD program performance (both
qualitatively and quantitatively) against
its historical site performance, fleetlevel performance, if applicable, and
industry performance. The licensee or
other entity would be required to
implement corrective actions if site FFD
performance meets a licensee- or other
entity-established threshold or to
resolve a finding resulting from a
qualitative review or audit in a manner
that restores performance and corrects
root causes, contributing causes, or
both; (3) the requirements in proposed
§ 26.609, ‘‘Behavioral observation,’’ are
more robust than those in § 26.407,
‘‘Behavioral observation,’’ of subpart K
of part 26 and are proposed to
synchronize with and reinforce the AA
behavioral observation requirements in
§ 73.56, ‘‘Personnel access authorization
requirements for nuclear power plants,’’
or the proposed requirements under
§ 73.120, ‘‘Access authorization program
for commercial nuclear plants’’; (4) a
part 53 commercial nuclear plant that
satisfies the § 26.603(c) criterion will be
designed, operated, and secured with a
radiological risk profile that is lower
than that described in § 53.860(a)(2) and
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perhaps will approach the radiological
risk profile of an NPUF (which does not
implement an FFD program); and (5) the
NRC is aware that a part 53 commercial
nuclear plant could be designed and
constructed in such a manner to reduce
reliance on an onsite security force to
protect SSCs, NRC-licensed materials,
and sensitive information, with
enhanced capabilities for the detection,
assessment, and delay of a DBT
adversary.
Regarding fatigue management
requirements, work hour controls would
be required for personnel at utilization
and manufacturing facilities in
accordance with the existing scoping
criteria in § 26.4, ‘‘FFD program
applicability to categories of
individuals,’’ as revised in this
proposed rule. The amended § 26.4 also
would be used to determine whether an
individual would be subject to drug and
alcohol testing. The applicability of
these scoping criteria for certain
individuals (such as operators and
maintenance personnel) would be
determined by the licensee or other
entity through its risk-informed
evaluation process performed to assess
the risk significance of the SSC upon
which work is being performed or
directed by the individual. These
requirements also would be scaled
based on the potential radiological
consequences presented by the facility.
However, fatigue management would be
applied to all individuals subject to the
FFD program, similar to FFD program
implementation by the current fleet of
commercial nuclear plants because
fatigue management is a proactive
requirement designed to help prevent
on-shift impairment through work hour
scheduling and time off. The behavioral
observation program (BOP) would be
the principal requirement to provide
reasonable assurance that individuals
on shift are not mentally or physically
impaired due to fatigue, which in any
way could adversely affect their ability
to safely and competently perform their
duties.
The NRC is proposing subpart M of
part 26 for facilities licensed under part
53, in lieu of subjecting all part 53
licensees to the same part 26
requirements that apply to facilities
licensed under part 50 or 52, for four
principal reasons. First, subpart M of
part 26 would apply FFD requirements
in a risk-informed manner
commensurate with the radiological
consequences presented by facilities
licensed under part 53. This regulatory
strategy is consistent with the current
part 26, which provides a
comprehensive set of deterministic
requirements for licensees and other
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entities at facilities that are operating.
This approach is also consistent with
the current subpart K of part 26, which
provides a more flexible framework for
nuclear power reactors under
construction, where the probabilities of
serious radiological accidents are lower
and consequences from such accidents
are less severe than at operating plants.
Second, subpart M of part 26 would
enable a part 53 licensee or other entity
to implement innovative drug testing
technologies and behavior observation
techniques while continuing to
demonstrate compliance with the part
26 performance objective in § 26.23(b) of
providing reasonable assurance that
individuals are not under the influence
of any substance or mentally or
physically impaired from any cause,
which in any way adversely affects their
ability to safely and competently
perform assigned duties. These
technologies include drug and alcohol
testing using oral fluid, urine, and hair
specimens; screening using point of
collection testing and assessment
(POCTA) devices; and monitoring using
passive drug and alcohol detection
instrumentation. Part of the basis to
enable the use of innovative drug and
alcohol testing technologies is to
maintain FFD program effectiveness
should the staff size at a part 53
commercial nuclear plant be small and
challenge the effective implementation
of the behavioral observation and drug
and alcohol testing programs. Also, a
commercial nuclear plant that is sited at
a geographically remote location may
present additional challenges to
behavioral observation and drug and
alcohol testing that are not presented by
traditional LWR facilities licensed
under part 50 or 52, such as: efficiency
of postal services for shipping and
controlling biological specimens;
proximity to drug and alcohol collection
facilities that are reasonably equivalent
to that described in subpart E of part 26;
availability of internet and cellular
services to enable same-time
discussions among the Medical Review
Officer (MRO), donor, and laboratory;
accessibility to substance abuse
treatment services described in subpart
H of part 26; and proximity to an MRO
(or management and clinical staff) to
evaluate potential impairment caused
by fatigue and/or substance use or
abuse, for-cause and post-event
occurrences, and the individual’s
potential to return to duty.
A part 53 commercial nuclear plant
that is sited in a geographically remote
location and has a small staff size may
present implementation challenges and
the potential for small group dynamics
to impact FFD program effectiveness.
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Particularly in isolated environments,
psychological phenomena known as
‘‘groupthink’’ may take effect and could
impact the effectiveness of BOPs and
the ability to effectively manage safety
culture. For example, in circumstances
where small staffs are drawn from the
same small town and thereby have a
potentially narrow experience base, it
could be challenging to maintain a
safety conscious work environment in
which personnel feel free to raise safety
concerns without fear of retaliation,
intimidation, harassment, or
discrimination, and organizations may
resultingly experience groupthink-like
effects. Groupthink is particularly
prevalent among cohesive and insulated
groups that experience high levels of
decisional stress.6 Small staffs at part 53
commercial nuclear plants may
therefore be more susceptible to
groupthink if they are working in an
isolated environment where decisionmaking pressures may be high.
Groupthink could have adverse effects
on workplace safety culture, as studies
show that individuals will be more
hesitant to speak out against practices
they deem unsafe for fear of deviating
from group norms.7 Individuals may
also be unaware of systematic biases in
the group decision-making process and
may then be less likely to scrutinize the
potential risks of the group’s decision or
sufficiently contemplate alternative
paths of action.8 Furthermore, the
literature indicates that groups make
riskier decisions than individuals acting
alone due to the diffusion of
responsibility among group members.9
6 See e.g., Irene W#r<, Ragnar Rosness, and Stine
Skaufel Kilska, ‘‘Human performance and safety in
Arctic environments,’’ SINTEF (2018).
7 See e.g., Russell Mannion and Carl Thompson,
‘‘Systematic biases in group decision-making:
implications for patient safety,’’ International
Journal for Quality I Health Care, Vol. 26, No. 6
(2014): 606–612 (arguing that small group dynamics
in healthcare teams produce systematic biases in
group decision-making because healthcare
professionals may be reticent to vocalize concerns
they have about quality of care).
8 See e.g., W#r<, Rosness, and Kilska (arguing
that groupthink leads teams to ‘‘develop shared
rationalizations that bolster a proposed choice,
rather than examining alternative options and
identifying the risks associated with the proposed
choice’’). See also David Hofmann and Adam
Stetzer, ‘‘A Cross-Level Investigation of Factors
Influencing Unsafe Behaviors and Accidents,’’
Personnel Psychology, Vol. 49 (1996) (finding that
in a study of fatal accidents involving offshore oil
rigs, in the absence of standard operating
procedures, workers ‘‘equated normal work
methods (i.e. what everyone else does) with safe
and/or ideal work methods,’’ revealing that the
groupthink phenomena will further cement modes
of work that do not reflect safety protocols in small
groups that lack strong norms around workplace
safety and tacitly reward short-cuts that prioritize
efficiency over safety).
9 Mannion and Thompson, ‘‘Systematic biases in
group decision-making: implications for patient
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This phenomenon, known as ‘‘the risky
shift,’’ also runs counter to a safety
culture. Accordingly, ‘‘groupthink’’ and
‘‘the risky shift’’ may lead to group
behaviors that render behavioral
observation less effective. As such,
alternative approaches to behavior
observation programs, such as the
utilization of video-based surveillance
by individuals separate from the onsite
work unit, could serve to mitigate
potential issues associated with
groupthink. The incorporation of remote
observation, performed by individuals
physically separate from the site, could
help to bring in independent and
objective perspectives and help to break
patterns of thought and communication
that may result in groupthink.
Even without the influence of small
group dynamics, there are other
practical constraints to implementing
FFD requirements, such as random drug
and alcohol testing, among small staffs.
Random testing is less effective when
applied to small staff sizes because it
may be easier for staff to communicate
and predict when individuals will be
subject to drug and alcohol testing.
Furthermore, if a facility is sited in a
remote location, program
implementation could be challenged by
the following factors: limited mail
services to laboratories certified by the
U.S. Department of Health and Human
Services (HHS), availability of local
clinical or medical options for treatment
and determinations of fitness by an
MRO or Substance Abuse Expert, and
use of offsite drug and alcohol
collection facilities.
The increased potential for small staff
sizes to impact FFD policy compliance
warrants an approach to FFD that
emphasizes performance over
prescriptive requirements that may be
ineffective or infeasible at these
facilities. Therefore, the NRC proposes
the subpart M of part 26 framework to
provide a performance-based approach
to FFD. For example, proposed
§ 26.603(d) would use existing part 26
auditing requirements and the reporting
requirement in § 26.717, ‘‘Fitness-forduty program performance data,’’ and
clarify how FFD performance data
would be used to maintain or improve,
if necessary, FFD program effectiveness.
Specifically, § 26.603(d) would require
each licensee and other entity that elects
to implement subpart M of part 26 to
monitor and assess their site-specific
performance against the preceding
year’s site performance, the licensee’s
most recent fleet-level performance, and
the most recent industry performance.
safety,’’ International Journal for Quality I Health
Care, Vol. 26, No. 6 (2014): 606–612.
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Licensees and other entities would use
these datapoints to develop performance
measures, which would be qualitative
descriptions of the specific FFD
program elements, and threshold values
for each performance measure that, if
exceeded, would indicate a performance
deficiency. Each licensee and other
entity would compare its site’s current
performance data against the
performance measures and, if a
threshold is exceeded, the licensee or
other entity would be required to take
corrective actions to restore
performance. Also, the NRC proposes a
change control requirement to allow a
licensee or other entity to change its
subpart M of part 26 FFD program while
ensuring that FFD program effectiveness
is maintained.
Lastly, subpart M of part 26 would
consolidate the applicable FFD
requirements by placing in one subpart
all proposed part 26 requirements
(either new requirements or crossreferences to existing part 26
requirements) for part 53 licensees and
other entities. This should help
licensees and other entities implement
the requirements because it would
enable easy cross-reference to similar
requirements in other subparts that are
being implemented by non-part 53
licensees and entities subject to part 26.
Understanding how other licensees or
other entities implement similar FFD
requirements may facilitate the sharing
of operating experience in program
implementation.
The use of innovative technologies
and a risk-informed performance-based
framework parallels the considerations
presented in the Advanced Reactor
Policy Statement. As stated in the policy
statement, ‘‘[S]implified systems should
facilitate operator comprehension,
reliable system function, and more
straightforward engineering analysis.’’
Furthermore, these same attributes may
reduce potential radiation exposures,
help prevent the theft of nuclear
materials, and use technology and
design innovations. Should these
components and systems be designed,
implemented, and maintained to
minimize reliance on human actions
and leverage technology and innovation,
then the robust and prescriptive FFD
requirements in, for example, subparts
B, ‘‘Program Elements,’’ and E of part 26
could be scaled to the part 53-licensed
facility and its operation. This strategy
would be implemented in the subpart M
of part 26 framework.
Even though current subpart K of part
26, provides for FFD requirements
commensurate with the radiological
consequences presented by a nuclear
power plant construction site, proposed
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subpart M of part 26 would not allow
part 53 licensees and other entities to
implement the requirements in subpart
K. The principal reasons are that
(without significant changes to subpart
K that would be outside the scope of
this rulemaking): (1) subpart K does not
apply to holders of MLs who assemble
or test a reactor; (2) subpart K only
applies during construction, whereas
subpart M would apply during
construction, operation, and
decommissioning through
implementation of the insider
mitigation program (IMP) required by
§ 73.55 or proposed § 73.100; (3) subpart
K does not address training,
authorization as defined in § 26.5, and
MRO performance; (4) subpart K does
not expressly authorize the use of
innovative drug and alcohol testing
technologies; (5) subpart K does not
describe the use of time-dependent
alcohol limits or special analysis testing
of dilute urine specimens; and (6)
subpart K has less rigor in the protection
of worker rights and sensitive
information than that proposed in
subpart M.
Despite the differences between
subparts K and M of part 26, the
requirements in subpart M would be
essentially equivalent to many in
subpart K that were implemented by the
licensees of Vogtle Nuclear Station and
V.C. Summer Nuclear Station when they
were constructing four commercial
nuclear power reactors and NRC
inspection and operating experience
evaluation determined that the use of
subpart K contributed to adequately
protecting the public health and safety
and the common defense and security.
Further, given the risk profile posed by
facilities licensed under part 53 and the
proposed additional requirements in
subpart M of part 26 that were
developed from operating experience
and other part 26 subparts (but are not
included in subpart K of part 26), the
NRC concludes that if licensees and
other entities effectively implement the
proposed requirements in subpart M of
part 26, then individuals subject to the
rule should be fit for duty and
trustworthy and reliable.
B. Proposed Changes to Part 26,
Subparts A Through E and I
Section 26.3(d) is the applicability
paragraph for contractor/vendors (C/Vs)
who implement FFD programs or
program elements, to the extent that the
licensees and other entities specified in
§ 26.3(a) through (c) rely on those C/V
FFD programs or program elements to
meet the requirements of part 26.
Section 26.3(d) would be amended to
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address part 53 licensees and other
entities in proposed § 26.3(f).
Proposed § 26.3(f) would place part 53
licensees or other entities within the
scope of part 26. For licensees and other
entities of a part 53 commercial nuclear
plant, except a holder of an ML, the FFD
program would be required to be
implemented no later than the start of
construction activities. The holder of an
ML would need to implement its FFD
program before commencing activities
that assemble a reactor.
Current § 26.4 describes FFD program
applicability to categories of
individuals. These categories are based
on the duties, responsibilities, and the
types of access an individual may
possess. The NRC proposes to amend
§ 26.4 to include licensees and other
entities described in § 26.3(f). The NRC
expects that not all categories of
individuals described in current § 26.4
would be applicable to all part 53
facilities. The NRC is proposing
regulatory guidance in DG–5073,
‘‘Fitness-of-Duty Programs for
Commercial Nuclear Plants and
Manufacturing Facilities Licensed
Under 10 CFR part 53,’’ and DG–5078,
‘‘Fatigue Management for Nuclear
Power Plant Personnel at Commercial
Nuclear Plants Licensed Under 10 CFR
part 53,’’ to help address program
applicability to certain individuals.
Section 26.4(a)(1) and (a)(4) would be
amended to account for the possibility
that certain individuals may perform or
direct the performance of operational
and maintenance activities from a
remote facility (for example, a remotecontrol station) for licensees or other
entities licensed under part 53.
The framework of the current part 26
does not account for individuals who
perform operating and maintenance
duties at remote facilities. Although
current § 26.4(a)(1) does not limit the
operating of applicable SSCs to onsite
operating, § 26.5 limits the definition of
‘‘Maintenance,’’ for the purposes of
§ 26.4(a)(4), to include only ‘‘onsite
maintenance activities.’’ In the 2008
part 26 final rule preamble, the NRC
explained that the work hour
requirements apply to those individuals
who perform maintenance activities
within the licensee’s owner-controlled
area. Furthermore, regarding the
direction of applicable operations and
maintenance activities, current
§ 26.4(a)(1) and (4) address only
individuals who perform ‘‘onsite
direction.’’
Under the proposed amendments to
part 26, the limitation of ‘‘onsite’’
activities to those performed within the
owner-controlled area would still apply
to facilities licensed under part 50 or 52.
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However, for licensees and other
entities described in § 26.3(f), the NRC
would remove the ‘‘onsite’’ limitation to
include activities performed both within
the owner-controlled area as well as
operations and maintenance duties
performed at remote facilities where
safety-significant systems and
components are expected to be operated
within the design basis of the
commercial nuclear plant.
In the 2008 part 26 final rule, the
purpose of limiting ‘‘directing’’
activities to those ‘‘directing’’ activities
that are conducted onsite was to avoid
requiring work hour controls for
individuals performing incidental
duties, consistent with § 26.205(b)(5),
from an offsite location in instances
where those duties might be considered
to be ‘‘directive’’ in nature. Under the
proposed amendments to part 26, the
exclusion of incidental duties while
calculating work hours would still be
applicable for licensees and other
entities licensed under part 53.
However, for these licensees and other
entities, beyond instances of incidental
duties, the direction of operations and
maintenance activities associated with
safety-significant SSCs, when performed
at remote facilities, would be considered
in an equivalent fashion as direction
performed at non-remote facilities, for
the purposes of administering work
hour controls.
Proposed § 26.4(b) would include in
an FFD program individuals who are
granted unescorted access to the
protected area of a facility licensed
under part 53 and do not perform or
direct the performance of the duties
described in § 26.4(a). This requirement
would contribute to the defense-indepth regulatory framework that helps
provide that individuals who have
unescorted access are fit for duty,
trustworthy, and reliable. For example,
the NRC is proposing amendments to
part 73 to require a part 53 licensee to
subject individuals to a series of reviews
to help determine whether those
individuals are trustworthy and reliable
before granting them unescorted access
to the facility’s protected area.
The NRC would amend § 26.4(c) to
include in an FFD program individuals
who are assigned to physically report to
the part 53 licensee’s emergency
response facility (or facilities) or
participate remotely in emergency
response activities, and individuals
without unescorted access to the part 53
facility who, remotely or otherwise,
make decisions and/or direct actions
regarding plant safety or security. Part
53 commercial nuclear plants may be
licensed for and rely upon offsite
facilities to fulfill the role of a Technical
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Support Center or Emergency
Operations Facility. Therefore, the
proposed rule would account for such
offsite facilities or remotely performed
activities. Further, the use of personnel
to operate systems and components,
maintain and surveil SSCs, and respond
to plant conditions and security events
may be different than those included in
the Technical Support Center or
Emergency Operations Facility team for
power reactors currently licensed under
part 50 or part 52.
For the individuals whose duties for
the licensees and other entities in
§ 26.3(c) require the individuals to have
the types of access or perform the
activities listed in § 26.4(e)(1) through
(6) at the location where the commercial
nuclear plant will be constructed and
operated, current § 26.4(e) requires them
to be subject to an FFD program that
satisfies all the requirements of part 26
except subparts I and K. The NRC
would amend § 26.4(e) to except subpart
M as well as subparts I and K. The NRC
would also amend § 26.4(e) to include
in an FFD program the individuals
whose duties for the licensees and other
entities in § 26.3(f) require the
individuals to have the types of access
or perform the activities listed in
§ 26.4(e)(1) through (6) or perform
construction activities as defined in
§ 26.5.
Section 26.4(e)(4) would be revised to
include in an FFD program individuals
who witness or determine inspections,
tests, and analyses certifications
required under part 53 because current
§ 26.4(e)(4) includes the individuals
who perform the same duties under part
52.
The proposed rule would amend
§ 26.4(f) to require individuals who
construct or direct the construction of
safety- or security-related SSCs at
facilities licensed under part 53 to be
subject to an FFD program under
subpart M of part 26 or an FFD program
that demonstrates compliance with all
of the requirements of part 26 except for
subparts I, K, and M of part 26.
Section 26.4(g) is the applicability
paragraph for FFD program personnel
(e.g., the FFD manager, MRO, and
technicians) and persons who perform
AA determinations (e.g., the licensee- or
other entity-designated Reviewing
Official). This section would be
amended to address part 53 licensed
facilities. Specifically, a part 53 licensee
or other entity would use FFD program
personnel to implement its FFD
program as well as other assigned
individuals who are not involved in the
day-to-day operations of the program to
implement specific elements of its FFD
program, such as the collection of a
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specimen for drug or alcohol testing.
These individuals would be held
accountable for program
implementation, including consistent
implementation of protections afforded
to all individuals subject to the FFD
program.
Section 26.4(h) would be amended to
include subpart M of part 26.
The NRC proposes to include several
new definitions in § 26.5, ‘‘Definitions,’’
and amend some existing definitions.
The NRC is proposing to add a
definition for ‘‘Biological marker.’’ The
proposed definition would be consistent
with ‘‘Biomarker’’ defined by the HHS
in its Mandatory Guidelines for Federal
Workplace Drug Testing (HHS
Guidelines) using oral fluid as the
biological specimen to be tested (84 FR
57554; October 25, 2019). However, the
proposed definition for § 26.5 would
add that the endogenous substance used
to validate that the biological specimen
‘‘was produced by the donor’’ because
subpart M of part 26 proposes to have
the MRO evaluate any discrepant
biological marker identified in a
biological specimen collected from a
donor.
The NRC is proposing a definition for
the word ‘‘Change’’ as used in the
proposed § 26.603(e), ‘‘FFD program
change control,’’ process. The proposed
definition would be consistent with the
definition of ‘‘Change’’ for a part 50 or
52 licensee’s emergency plans in
§ 50.54(q)(1)(i).
The NRC proposes to revise the
definition of ‘‘Constructing or
construction activities’’ to clarify that
for licensees or other entities in
§ 26.3(f), the definition of
‘‘Construction’’ would be that as
proposed in § 53.020.
The definitions of ‘‘Contractor/
vendor’’ (C/V) and ‘‘Other entity’’ would
be revised to make them applicable to
part 53 licensees. A holder of an ML
under part 53 could be a C/V under the
proposed C/V definition.
The NRC is proposing a definition for
‘‘Illicit substance’’ because this phrase is
used in subpart M of part 26 and would
address substances that cause
impairment and possible addiction but
are not an ‘‘illegal drug’’ as defined in
§ 26.5. This proposal is based on
operating experience where individuals
have admitted to using common
household, non-drug substances to
achieve a high or satisfy an addiction.
These common household items
include, but are not limited to nitrous
oxide, butane, propane, glue, paint
vapors, lighter fluid, nail polish
remover, degreasers, permanent
markers, and methyl alcohol (which is
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found in hand sanitizer and
mouthwash).
The definition of ‘‘Questionable
validity’’ would be revised to make it
applicable to an FFD program
implemented under subpart M of part
26, which would include all biological
specimens.
The NRC is proposing a definition for
‘‘Reduction in FFD program
effectiveness’’ because this phrase,
similar to the proposed definition for
‘‘Change,’’ is used in proposed
§ 26.603(e). The proposed definition is
generally consistent with the definition
of ‘‘Reduction in effectiveness’’
provided for emergency plans in
§ 50.54(q)(1)(iv).
The proposed rule would make the
current definition of ‘‘Reviewing
official’’ applicable to those licenses and
other entities in § 26.3(f).
The current part 26 definition of
‘‘Safety-related structures, systems, and
components’’ would be amended to use
the NRC’s proposed definition in
§ 53.020 for the part 53 licensees and
other entities described in § 26.3(d) and
(f).
The NRC would amend the definition
of ‘‘Security-related SSCs’’ in § 26.5 to
make it applicable to a licensee or other
entity described in § 26.3(d) and (f).
The NRC proposes a definition for
‘‘Special Nuclear Material’’ that would
refer to the definition in § 70.4,
‘‘Definitions,’’ of part 70 to ensure
consistency.
The NRC is proposing a revision of
the definition of ‘‘Unit outage’’ to
account for the potential use of
commercial nuclear plants for purposes
other than electricity generation.
Section 26.21, an applicability
statement for part 26 FFD programs,
would be amended to include licensees
and other entities described in § 26.3(f)
that choose to implement an FFD
program that implements all part 26
requirements, except those in subparts
K and M of part 26.
Section 26.51, ‘‘Applicability,’’ would
be amended to apply to licensees and
other entities described § 26.3(f) that
elect not to implement the requirements
in subpart M of part 26 for the categories
of individuals in § 26.4 and those
licensees and other entities that elect to
implement the requirements in § 26.605.
Section 26.53(e), (e)(1) and (3), and (g)
through (i), which are general
provisions for granting and maintaining
authorization, would be amended to
apply to licensees and other entities
described § 26.3(f).
Section 26.63(d), a suitable inquiry
requirement, would be amended to
apply to licensees and other entities
described § 26.3(f).
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Section 26.73, the applicability
statement for subpart D of part 26,
would be amended to apply to licensees
and other entities described § 26.3(f)
that elect not to implement the
requirements in subpart M of part 26 for
the categories of individuals in § 26.4
and those licensees and other entities
that elect to implement the
requirements in § 26.605(b).
Section 26.81, the purpose and
applicability statement for subpart E of
part 26, would be amended to apply to
licensees and other entities described in
§ 26.3(f) that elect not to implement the
requirements in subpart M of part 26 for
the categories of individuals in § 26.4
and those licensees and other entities
that implement proposed § 26.605(a) or
(b). The subpart E requirements to be
implemented are listed in proposed
§ 26.607(c)(2)(i) and (c)(2)(ii) and (c)(3).
Section 26.201, the applicability
statement for subpart I of part 26 would
be amended to apply to licensees and
other entities described in § 26.3(f).
Also, the applicability statement would
be divided into two paragraphs for
clarity.
The NRC proposes to add § 26.202,
‘‘General provisions for facilities
licensed under part 53,’’ for licensees or
other entities described in proposed
§ 26.3(f) that elect to implement the
requirements in subpart I of part 26 in
accordance with § 26.604 and § 26.605.
Section 26.202 would establish
requirements equivalent to those in
current § 26.203, ‘‘General provisions,’’
which is applicable to part 50 and 52
licensees. The NRC would add the
separate § 26.202 because § 26.203 refers
to various requirements under subpart B
of part 26, which would not be
applicable to facilities licensed under
part 53 that implement subpart M of
part 26.
Additionally, § 26.202(c), ‘‘Training
and assessments,’’ unlike § 26.203(c),
‘‘Training and examinations,’’ would
not include a comprehensive
examination requirement because
trainee assessment is conducted as part
of a SAT that would be required as
proposed under the FFD program
training requirements in § 26.608.
Proposed changes in §§ 26.205,
26.207, and 26.211 would add
references to new requirements in
subparts I and M of part 26 that would
be applicable specifically to licensees
and other entities in § 26.3(f). The NRC
would not change the specific
provisions for work hour requirements
in current § 26.205(d). However, as
addressed in the discussion of proposed
changes to § 26.4(a), whether a licensee
or other entity under part 26 would
need to implement work hour controls
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for certain individuals or groups would
be dependent, in part, on
determinations reached by that
licensee’s risk-informed evaluation
process.
Proposed changes to
§§ 26.207(a)(1)(ii) and 26.211(b) would
allow licensees and other entities in
§ 26.3(f) to perform face-to-face
assessments to support the approval of
work hour control waivers and the
conduct of fatigue assessments,
respectively, using electronic
communications. These proposals
would allow supervisors to conduct
such assessments from a remote location
under appropriate circumstances. Such
remotely conducted assessments would
need to be supported by someone who
is present in-person with the individual
being assessed and who is trained in
accordance with the requirements of
either § 26.29 and § 26.203(c) or § 26.608
and § 26.202(c). The reasoning for these
proposals and the associated need for
in-person support to augment electronic
communications is addressed further in
the preamble discussion of proposed
§ 26.619.
C. Proposed Requirements for Part 26,
Subpart M
The proposed rule would add a new
subpart M to part 26 that would provide
alternative FFD requirements for part 53
licensees and other entities.
Proposed § 26.601 would make
subpart M of part 26 applicable to part
53 licensees and other entities, at their
discretion. If a licensee or other entity
in § 26.3(f) does not elect to implement
an FFD program that demonstrates
compliance with the requirements of
subpart M, then the individuals
specified in § 26.4 would be subject to
an FFD program that demonstrates
compliance with all part 26
requirements, except for those
requirements in subparts K and M.
Proposed § 26.603(a) would require an
applicant to provide a description of its
FFD program and its implementation
within its application for a license. This
requirement is equivalent to the existing
requirements in §§ 26.401(b) and
52.79(a)(44). The entities that would be
required to submit these FFD program
descriptions are certain applicants that
would comply with the part 53
application requirements in subpart H.
In subpart H, § 53.1309(a)(6) would
require an applicant for a CP to provide
a description of its FFD program in its
PSAR. Under §§ 53.1279(b)(4),
53.1369(x), and 53.1416(a)(24), an
applicant for an ML, OL, and COL,
respectively, would be required to
provide a description of its FFD
program in its FSAR.
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Unlike an application for a license, a
description of an FFD program does not
receive NRC review for possible
approval. The applicant provides the
NRC with information about the
applicant’s proposed FFD program to
inform the NRC’s inspection program
and to demonstrate that the FFD
program will be effectively
implemented before a licensee or other
entity commences any activity making
individuals at the NRC-licensed facility
subject to the FFD program.
Proposed § 26.603(a)(1) would require
a summary description of the analysis
described in § 26.603(c), if performed.
The analysis should describe the
operation of the facility. This would
include informing the Commission of:
(1) the principal individuals assigned by
job title (work category) and a summary
description of the human actions (e.g.,
monitoring, operating, responding,
surveillance, oversight, etc.) that they
perform to maintain the facility in a safe
operating or shutdown condition; (2) the
principal individuals by job title and a
summarized description of the human
actions to secure and protect the facility
(without providing sensitive
information); (3) the estimated total
population of individuals subject to the
FFD program and per shift by job
description; and (4) references to
supporting documentation. The purpose
of these descriptions is to enable an
NRC assessment of the licensee’s or
other entity’s analysis and the required
human actions to operate, monitor,
surveil, maintain, and secure the facility
within its design and licensing basis so
that if an operational or security-related
event were to occur, the facility would
respond as designed and licensed and
the calculated radiological dose
consequences would not exceed the
consequences described in
§ 53.860(a)(2). This is important because
facilities that implement § 26.604 are
expected to have very small staff sizes
and may be sited in geographically
remote locations, both of which could
challenge effective implementation of
the FFD program.
Proposed § 26.603(a)(2) would require
the applicant to state what FFD program
it plans to implement.
Proposed § 26.603(a)(3) would require
a discussion that informs the NRC of the
applicability of the applicant’s FFD
program to individuals who perform
safety- or security-significant activities.
This description should summarize any
key differences between the staff at the
site and any remote facility and the
categories of individuals in § 26.4. The
principal purpose of providing this
description would be to inform the NRC
of any substantial differences in the
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applicability of the FFD program to the
categories of individuals in § 26.4.
Proposed § 26.603(a)(4) would require
a description of the drug and alcohol
testing and fitness determination
process to be implemented through the
licensee’s or other entity’s procedures,
including the collection and testing
facilities to be used, biological
specimens to be collected, and sanctions
to be imposed upon a confirmed FFD
policy violation. This process includes
how individuals who test positive for a
drug or alcohol will be evaluated before
being afforded unescorted access to the
protected area to perform or direct those
duties or responsibilities making them
subject to the FFD program. The
principal purpose of describing this
return-to-duty process is to inform the
NRC of the behavioral observation
strategy (for those facilities that
implement § 26.604) and/or drug
screening and testing strategy.
Proposed § 26.603(a)(5) would require
a summary description of the
applicant’s planned PMRP. This
description must provide the
performance measures and thresholds
that the applicant intends to use.
Proposed § 26.603(b) would establish
when the FFD program must be
implemented and the longevity of the
FFD program. This proposal is
equivalent to the current § 26.3, which
states, in part, when licensees and other
entities must begin implementing their
FFD programs. Unlike the current part
26 regulations, proposed § 26.603(b)
would expressly state that an FFD
program would not be applicable during
decommissioning of a part 53 facility for
licensees and other entities specified in
§ 26.3(f). However, licensees of facilities
licensed to operate a reactor should be
aware that the physical protection
program under § 73.55, ‘‘Requirements
for physical protection of licensed
activities in nuclear power reactors
against radiological sabotage,’’ and
under proposed § 73.100 include a
requirement for the implementation of
an IMP, even during decommissioning.
Proposed § 26.603(b) would also
require the holder of an ML to
implement its FFD program no later
than the start of activities that assemble
a reactor. The holder of the ML should
establish in its procedures when reactor
assembly commences and what
constitutes assembly. For example, the
FFD program would not need to be
implemented for the receipt, storage,
inspection, and staging of components
and systems used to assemble (i.e., build
or fabricate) the reactor because this is
not a current requirement for LWR
facilities licensed under part 50 or 52.
Furthermore, the NRC currently does
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not require that an FFD program be
applied to the assembly or
manufacturing of components (or basic
components as defined in § 21.3), or
systems that were fabricated or
assembled outside the footprint of a
commercial power reactor, and this
regulatory position would also apply to
a manufacturing facility.
Proposed § 26.603(c) would require
the applicant, licensee, or other entity
seeking to implement an FFD program
under § 26.604 to perform a site-specific
analysis to determine whether the
facility and its operation satisfy the
criterion in § 53.860(a)(2). If the analysis
is performed and demonstrates that the
radiological consequences presented by
the facility and its operation satisfy the
criterion, then the licensee or other
entity could implement the FFD
program detailed in § 26.604. If the
analysis does not demonstrate that the
facility and its operation satisfy the
criterion, then the licensee or other
entity must implement the FFD program
described in either § 26.605 or subparts
A through I, N, and O of part 26.
Proposed § 26.603(c) would also
require licensees and other entities that
implement proposed § 26.604 to update
the technical analysis used to justify
compliance with the criterion in
§ 53.860(a)(2). This analysis would be
updated to reflect changes made to the
staffing, FFD programs, or offsite
support resources described in the
analysis to show that the facility and its
operation continue to satisfy the
criterion. This is important because
facility, operation, or staffing changes
outside FFD program implementation
(e.g., changes in the facility safety
analysis, physical protection strategies,
or the security plan, implementing
procedures, or contingency response
strategies) could adversely impact the
licensee’s or other entity’s documented
analysis demonstrating that the facility
and its operation satisfy the criterion if
event sequences require human action.
Proposed § 26.603(d) would require
the establishment of a PMRP. The
concept of a PMRP is not new. This
requirement would consolidate for part
53 the requirements in current §§ 26.41,
‘‘Audits and corrective actions’’; 26.415,
‘‘Audits’’; 26.717, ‘‘Fitness-for-duty
program performance data’’; and
26.183(c), which describes MRO
responsibilities. The proposal would
state that the licensee or other entity
must monitor the effectiveness of its
FFD program by comparing performance
data against performance measures and
thresholds. The development of
quantitative thresholds would be new,
but this is born from licensees and other
entities with facilities licensed under
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parts 50 or 52 already collecting,
reviewing, and reporting FFD
performance data. Additionally, the
benefit of quantitatively measuring FFD
program performance against
established thresholds benefits a
licensee’s and other entity’s
determination of whether they are
maintaining FFD program performance
in a manner that demonstrates
compliance with the performance
objectives in § 26.23.
The NRC is proposing the PMRP
because the subpart M of part 26
requirements would enable a high
degree of flexibility in FFD program
implementation (e.g., drug testing). A
licensee or other entity would not only
have options in the type of FFD program
they may implement under part 26, but
they would have options in the types of
biological specimens they may test for
drugs, where to collect the biological
specimens (e.g., at the NRC-licensed
facility or offsite at a local hospital or
clinic), and the use of collection and
assessment devices to screen
individuals for drugs and alcohol. These
FFD program flexibilities could cause
FFD programs under subpart M of part
26 to become very site-specific,
necessitating performance measures to
enable the licensee or other entity to
maintain the effectiveness of its FFD
program.
Fitness-for-duty program effectiveness
would be determined by comparing
actual performance against the
performance measures and thresholds.
The result of that comparison would
inform licensee or other entity decisions
whether to change FFD program
elements to address a performance
deficiency. Also, the thresholds would
have sufficient margin, based on
operating experience, before conditions
adverse to safety and security may occur
should an individual be identified as
impaired or not trustworthy and
reliable. The potential of a humanrelated failure causing a condition
adverse to safety and security is
dependent on the duties and
responsibilities of the individual and
the defense-in-depth designed to
prevent or mitigate an adverse
consequence. The PMRP would account
for this by requiring the review of FFD
performance data, in part, by work
category, C/V, and individuals
employed by the licensee who are not
a C/V as defined in § 26.5 (i.e., a
licensee employee).
Proposed § 26.603(d)(1) would require
the licensee or other entity to document
and maintain its PMRP. Proposed
§ 26.603(d)(1)(i) would require that the
performance measures be identified and
designed to monitor FFD program
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performance. Proposed
§ 26.603(d)(1)(i)(A) would require the
FFD program of a licensee or other
entity subject to the requirements of
§ 26.604 to include monitoring of the
BOP. The purpose of this monitoring is
to help ensure that individuals subject
to the FFD program are observing the
behaviors of others, are being observed
themselves, and are reporting FFD
concerns to licensee- or other entitydesignated individuals. The other
performance measures would include
occurrence of FFD policy violations
evaluated by licensee employee, C/V,
and labor category, and occurrence of
individuals with potentially
disqualifying information or who
possessed an FFD prohibited item.
Proposed § 26.603(d)(1)(i)(B) would
require the FFD program of a licensee or
other entity that is either subject to the
requirements of § 26.604 and has
implemented a drug testing program at
its discretion, or is subject to the
requirements of § 26.605, to include the
performance measures identified in
§ 26.603(d)(1)(i)(A) and those necessary
to monitor the effectiveness of the drug
and alcohol testing program. The drug
and alcohol measures would include the
monitoring of FFD performance data for
pre-access and random testing and
subversion attempts by the categories of
licensee employee, C/V, and labor
category.
Proposed § 26.603(d)(1)(ii) would
require the licensee or other entity to
establish thresholds for each
performance measure. Initial thresholds
must be based on FFD performance data
from comparable facilities subject to
part 26, the licensee’s or other entity’s
fleet-level program performance if
applicable, and industry FFD
performance data. This provision
introduces the requirement to ‘‘maintain
FFD program effectiveness.’’ This
terminology describes a performancebased regulatory strategy in which the
licensee or other entity must initially
establish a level of performance that is
representative of other facilities in the
licensee’s fleet of facilities subject to
part 26, if applicable, and the FFD
performance of comparable facilities
subject to part 26.
Proposed § 26.603(d)(1)(iii) would
require that the licensee or other entity
evaluate FFD data as it is received to
determine whether a threshold has been
exceeded. Historical FFD performance
data for the current LWR fleet indicates
that, for particular work categories and
employment types, few FFD policy
violations occur per year. Therefore, for
work categories that may be significant
to worker safety (e.g., radiation
protection technicians), physical
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protection (i.e., security personnel), or
safety (i.e., NRC-licensed operators and
individuals who perform or direct the
performance of activities that a riskinformed evaluation process has shown
to be significant to public health and
safety), a single FFD policy violation
could be a significant occurrence and
warrant corrective actions. Based on
licensee-submitted FFD-related reports
under §§ 26.417, 26.419, 26.717, and
26.719, licensees and other entities with
facilities licensed under parts 50 or 52
implement some form of corrective
action that is typically scaled to the
significance of the violation. These
corrective actions have included
counseling, follow-up drug and/or
alcohol testing, remedial training,
generic announcements to the
workforce, and reviews of recently
performed or directed work by the
individual suspected of being impaired.
Proposed § 26.603(d)(1)(iii) would
require that the PMRP include a year-toyear comparison of FFD performance
data to help provide assurance that an
adverse trend in FFD program
performance would be identified if
occurring. This proposed requirement
was developed from the annual FFD
performance data reporting
requirements in §§ 26.417(b)(2) and
26.717. In particular, the proposed yearto-year comparison of FFD performance
data is equivalent to § 26.717(c), which
requires, in part, licensees and other
entities to analyze their performance
data at least annually and take
appropriate actions to correct any
identified program weaknesses.
Proposed § 26.603(d)(1)(iv) would
require the licensee or other entity to
perform and document quantitative and
qualitative reviews. These reviews
would be performed in three program
areas: protections afforded to
individuals subject to the FFD program,
laboratory test results and MRO
performance, and change control. The
purpose of these reviews would be to
specifically target performance within
the three program areas to assess
whether the outcomes resulting from the
implementation of procedure
requirements are contributing to FFD
program effectiveness. The proposed
reviews would not require the
establishment of measures and
thresholds because the reviews are
expected to result in qualitative findings
regarding program effectiveness.
Qualitative findings and observations
could still result in the consideration of
corrective actions in the targeted
program areas.
Proposed § 26.603(d)(1)(iv)(A) would
require the licensee or other entity to
monitor whether its FFD program is
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affording appropriate protections to
individuals subject to the FFD program.
The review of these protections would
include, in part, assessing the licensee’s
or other entity’s protection of the
following: privacy during the specimen
collection process; specimen integrity,
custody, and control; information
gathered from FFD program
implementation; and due process during
appeals of FFD policy violations.
Proposed § 26.603(d)(1)(iv)(B) would
require, in part, a review of laboratory
test results and MRO performance.
Effective performance by the laboratory
(e.g., obtaining and communicating
accurate test results) and MRO (e.g.,
correct evaluation of the laboratory test
results based on § 26.185 or HHS
Guidelines) would result in three
significant outcomes: (1) protection of
the donor from an inaccurate FFD
policy violation determination; (2)
protection of the donor, other
individuals, and the facility from
potential harm should the donor be
impaired or not trustworthy and
reliable; and (3) a performance-based
assessment of both the laboratory and
MRO. This last outcome could facilitate
actions to improve laboratory
performance, MRO training under
§ 26.607(m), or both. Proposed
§ 26.603(d)(1)(iv)(B) would also require
a comparative analysis between the
POCTA screening result(s) and the
corresponding specimen test results
obtained from the HHS-certified
laboratory if the POCTA indicated a
positive, adulterated, substituted, or
invalid screening result or discrepant
biological marker, to assess the
effectiveness of the POCTA and to
inform MRO decisions under § 26.185 or
§ 26.607(m)(6). The results of this
biennial review could also inform the
conduct of laboratory audits.
Proposed § 26.603(d)(1)(iv)(C) would
require that the change control
requirement in proposed § 26.603(e) be
included in the biennial program review
to help ensure that changes
implemented over the life of the facility
do not result in a reduction in program
effectiveness even if a mitigating action
was implemented for the specific
change. This requirement was
developed from §§ 26.137(f) and
26.713(d). This part of the review would
require an assessment of all changes
since the last review and their potential
aggregated impact on FFD program
effectiveness. For example, if last year
the licensee elected to contract with a
different MRO and this year the licensee
implemented a new type of POCTA
device, each of those program changes
probably would not have resulted in a
recognizable reduction in FFD program
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effectiveness. But, if the drug testing
positivity rate (or FFD policy violations)
for C/Vs decreased markedly during a
future maintenance outage that required
many C/Vs, then the reduction could
indicate, for example, that the POCTA
device was not as effective as
determined by a forensic toxicologist
review under §§ 26.603(e) and 26.607(h)
or that the new MRO was improperly
crediting prescription medication for
laboratory-confirmed positive test
results.
Proposed § 26.603(d)(2) would state
when the licensee or other entity must
implement corrective actions. This
requirement would be equivalent to the
requirement in current § 26.415(b) and
was developed from requirements
contained in §§ 26.41(a) and (f),
26.127(e), 26.129(b)(1)(i), 26.137(f)(3)
through (5), 26.155(a)(6), 26.157(e),
26.159(b)(1)(i), and 26.203(e)(2).
Corrective actions must be implemented
to correct root causes, contributing
causes, or both. There is margin built
into the FFD performance thresholds
and qualitative factors (e.g., to account
for potential changes in drug and
alcohol testing performance data when
there is a large influx of C/Vs to perform
maintenance) that may influence a
licensee or other entity’s causal
determination for an occurrence. Thus,
generalized or qualitative corrective
actions may be implemented like
informing management and placing a
sufficiently descriptive summary of the
occurrence in a corrective action
program for future monitoring to assess
recurrence.
However, should the occurrence
challenge safety or security or
significantly exceed a performance
threshold even when considering
qualitative factors and margin, the
licensee or other entity should
implement more robust corrective
actions to resolve the cause. An example
of a challenge to safety or security
would be the situation when an NRClicensed operator or maintenance
professional had operated, surveilled, or
maintained safety-significant SSCs and
was determined to have been impaired
by behavioral observation or potentially
under the influence of a narcotic as
determined by an alcohol or drug test or
screening result. Immediate corrective
actions could include, but would not be
limited to, a licensee or other entity
assessment of the duties and
responsibilities recently performed by
the individual. Operating experience
within the LWR operating reactor
community demonstrates few FFD
policy violations per year per site have
been caused by individuals who
perform or direct the performance of
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safety or security-significant activities.
Therefore, any such violations of the
FFD policy in a particular work category
in one year could be a significant
performance deficiency. These
violations could be even more
significant at part 53 facilities that have
a very small workforce subject to part
26.
Proposed § 26.603(d)(3) would require
the licensee or other entity to biennially
assess and document its FFD
performance monitoring program; this
requirement was developed from
§ 26.41(b). This documented review
would demonstrate that the
performance measures and thresholds
are appropriate based on site- and
licensee’s fleet-level program
performance, if applicable, and industry
performance and adjusted to maintain
FFD program effectiveness. Also, as a
result of this effort, the licensee or other
entity would be in possession of lessons
learned from fleet-level performance, if
applicable, and industry performance
that could contribute to their own
performance assessment to maintain
program effectiveness.
Under proposed § 26.603(d)(3)(i), the
identified program weaknesses and
corrective actions resulting from the
biennial review would be required to be
summarized in the licensee’s or other
entity’s annual report to the NRC in
compliance with either § 26.417(b)(2) or
§ 26.717, as applicable. This information
would inform the NRC of FFD program
weaknesses to facilitate regulatory
oversight and enable the NRC to
aggregate industry data for use in a
licensee or other entity PMRP.
Proposed § 26.603(d)(3)(ii) would
establish when the biennial PMRP
review must be completed and when
corrective actions from the review must
be implemented. The NRC selected the
May 15th date of odd-numbered years to
help ensure that all FFD programs will
maintain their previously determined
performance measures and thresholds or
reset them based on FFD program
performance early in the year in which
the biennial review was conducted. This
would assist in obtaining quality FFD
performance data over two annual
reporting cycles and evaluating whether
previous corrective actions were
effective.
In proposed § 26.603(e), the NRC
proposes a change control requirement
for subpart M of part 26 FFD programs.
Requiring licensees and other entities to
demonstrate compliance with certain
requirements before implementing
changes to their FFD programs would be
necessary for two primary reasons. First,
proposed changes to a licensee’s or
other entity’s FFD program could affect
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the analysis performed by the licensee
or other entity under proposed
§ 26.603(c), which helps determine the
FFD program requirements that must be
implemented. If this analysis changes,
then the licensee’s or other entity’s FFD
program requirements might change.
Second, the requirements in subpart M
of part 26 are performance based.
Therefore, FFD program implementation
may change periodically in response to
societal changes in substance abuse or
from PMRP implementation. Change
control therefore relies on the licensee
or other entity maintaining its
procedures in a manner that details how
its FFD program is to be implemented
while incorporating changes, with
documentation that justifies the changes
to support the PMRP, audits, and NRC
inspection.
Proposed § 26.603(e)(1) would permit
the licensee or other entity to
implement changes to its FFD program
if it performs and retains an analysis
demonstrating that the change does not
reduce the effectiveness of the FFD
program or the change was necessitated
or justified by a change to part 26,
laboratory processes, or guidance issued
by the HHS or NRC. The proposed
change control requirement would
enable flexibility in program
implementation should the NRC or HHS
change its drug testing procedures (as
implemented by the licensee or other
entity through its procedures) in
response to changes in societal
substance abuse or drug testing
technologies.
The proposed change control
requirement was developed from the
change control requirements in
§ 50.54(p) and (q)—the change control
requirements for security and
emergency plans, respectively.
However, unlike these two
requirements, the NRC does not review
and approve a licensee’s or other
entity’s FFD program or its
implementing procedures, and the FFD
program is not licensing-basis
information as described in § 53.1300.
Proposed § 26.603(e)(2) would require
that if a change reduces FFD program
effectiveness, then the licensee must
implement a mitigating strategy so the
FFD program, as revised, will continue
to demonstrate compliance with the
performance objectives in § 26.23 and
not result in a reduction in program
effectiveness.
Proposed § 26.603(e)(3) would
prohibit, with one exception, the use of
the change control process to reduce the
minimum panel of drugs to be tested
and would reference the drugs listed in
proposed § 26.607(c)(1). Proposed
§ 26.607(c)(1) would reference current
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§ 26.31(d)(1), which states that, at a
minimum, licensees and other entities
shall test for marijuana metabolite,
cocaine metabolite, opioids (codeine,
morphine, 6-acetylmorphine,
hydrocodone, hydromorphone,
oxycodone, and oxymorphone),
amphetamines (amphetamine,
methamphetamine,
methylenedioxymethamphetamine, and
methylenedioxyamphetamine),
phencyclidine, and alcohol. The testing
of these drugs and drug metabolites,
except phencyclidine, and alcohol is
necessary for the FFD program to
remain effective. Also, there is no
proposed subpart M of part 26
requirement stating that this panel of
drugs and drug metabolites needs to
consist of only scheduled drugs.10 This
flexibility would account for the
situation where an impairing substance
becomes prevalent in society and a
licensee or other entity elects to add the
substance to their panel of substances to
be tested prior to it being scheduled by
the Drug Enforcement Administration.
The exception in proposed
§ 26.603(e)(3) would be that, should
HHS elect to remove phencyclidine
from the panel of drugs and drug
metabolites to be tested, a licensee or
other entity could make this change in
its FFD program without resulting in a
reduction in FFD program effectiveness.
This outcome would be justified based
on the very infrequent occurrence rate
of FFD policy violations due to
phencyclidine use since 2010. However,
if HHS proposes to remove a class of
drugs from the panel of drugs to be
tested that is listed in § 26.31(d)(1),
except for phencyclidine, then a
licensee or other entity may not make a
similar change to its panel of drugs to
be tested, because this change would be
a reduction in FFD program
effectiveness even with a mitigative
strategy implemented.
Changes in the HHS panel of drugs
and drug metabolites to be tested may
also shift from one metabolite to a
10 The Drug Enforcement Administration
classifies drugs, substances, and certain chemicals
used to make drugs into five (5) distinct categories,
depending upon the drug’s acceptable medical use
and the drug’s abuse or dependency potential.
These categories appear as Schedules I through V
of section 202 of the Controlled Substances Act (21
U.S.C. 812). Schedule I drugs have a high potential
for abuse, have no currently accepted medical uses
in treatment in the United States, and lack accepted
safety for use under medical supervision. At the
other end of the classification scheme, Schedule V
drugs have the least potential for abuse among the
five categories of drugs, have a currently accepted
medical use in treatment in the United States, and
abuse of the drug may lead to limited physical
dependence or psychological dependence. For more
information, see https://www.dea.gov/druginformation/drug-scheduling.
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different metabolite for the same drug
class (e.g., amphetamines, opioids) to be
tested. Should HHS issue such a change
to its panel, this would not be expected
to result in a reduction in FFD program
effectiveness because HHS would be
targeting a more prevalent or effective
metabolite in its drug testing program.
This situation could occur as HHS
gathers more operating experience from
Federal Government implementation of
its HHS Guidelines, or data generated by
drug testing laboratories and federally
mandated drug testing programs
required by Federal agencies such as the
NRC and U.S. Departments of
Transportation, Energy, and Defense.
Proposed § 26.603(e)(4) would require
that change control records be
maintained for a 5-year record retention
period based on the current NRC
practice to conduct triennial inspections
of licensees’ and other entities’ FFD
programs. This would afford the NRC an
opportunity to review the licensee’s or
other entity’s determination that FFD
program changes have not reduced the
effectiveness of their FFD program.
Licensees and other entities would also
be required to summarize each change
made under proposed § 26.603(e) in
their annual FFD performance reports
required by § 26.617(b)(2) or § 26.717, as
applicable.
Proposed § 26.604 would establish the
minimum set of FFD program
requirements for licensees and other
entities who have a documented
analysis that demonstrates that the
facility and its operation satisfy the
criterion in § 53.860(a)(2). For these
licensees, compliance with the
performance objectives in § 26.23 would
be ensured through the BOP; defense-indepth measures proposed in subpart M
of part 26 like the PMRP, change
control, and audits; and other
requirements, such as those for AA,
physical protection, and licensed
operators. The adequacy of these
measures in satisfying the performance
objectives is supported by operating
experience, which demonstrates margin
between an FFD-related occurrence and
a condition adverse to safety or security,
as illustrated by for-cause, post-event,
and random testing data. A facility that
satisfies the criterion in proposed
§ 53.860(a)(2) would present a smaller
potential radiological consequence than
a facility that does not satisfy the
criterion, so the requirements in
proposed § 26.604 are scaled to the
lower risk presented consistent with the
Commission’s Advanced Reactor Policy
Statement.
The disadvantages of implementing
the FFD program described in proposed
§ 26.604 would be few. Since drug and
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alcohol testing would not be required,
behavioral observation would be the
keystone requirement in this
performance-based framework to
provide that individuals are fit for duty,
trustworthy, and reliable, and can safely
and competently perform the duties and
responsibilities making them subject to
the FFD program. If not, the individuals
would be assessed in accordance with
the licensee’s or other entity’s
procedures similar in manner to that
required by subpart K of part 26, and the
proposed PMRP would require
corrective actions should a threshold be
exceeded.
If a licensee or other entity elects not
to perform the analysis in proposed
§ 26.603(c) to determine whether it
satisfies the criterion in proposed
§ 53.860(a)(2); performs the analysis and
finds that the facility and its operation
does not satisfy the criterion in
proposed § 26.603(c); or is a holder of an
ML, the licensee or other entity could
not implement the FFD program
described in § 26.604. Instead, the
licensee or other entity would
implement either the program described
in proposed § 26.605 or an FFD program
that demonstrates compliance with all
the requirements in current subparts A
through I, N, and O of part 26.
Proposed § 26.605 would establish
requirements in a graded manner
similar to the regulatory framework
established by the requirements in
subparts A through I, N, O, and K of part
26. This existing graded approach
consists of an FFD program for
construction of a commercial nuclear
plant and a more robust program that
must be implemented before reactor
operation. The former is the FFD
program in proposed § 26.605(a), and
the latter is proposed § 26.605(b). Like
that for an FFD program under § 26.604,
the FFD program under § 26.605 would
include FFD program elements similar
to those in subpart B of part 26, but the
proposed requirements are less
prescriptive, enabling more flexibility in
program implementation like that
offered in subpart K of part 26. For
example, the requirements in subpart B
of part 26 are explicit requirements for,
in part, the collection and analysis of
urine specimens. Subpart B of part 26
does not enable the use of oral fluid for
drug testing or screening, except under
very limited situations as described in
subpart E of part 26, or the use of hair
specimens, unlike proposed § 26.605.
Proposed § 26.605 would require drug
and alcohol testing based on either the
requirements in part 26 or the HHS
Guidelines. The principal benefit of the
proposed § 26.605 FFD program is that
it would provide a regulatory framework
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that is consistent with the radiological
consequences for a facility that does not
satisfy the criterion in proposed
§ 53.860(a)(2) while affording
flexibilities in the conduct of drug and
alcohol testing.
Proposed § 26.605(a) would apply to
licensees and other entities who
perform the § 26.603(c) analysis and
satisfy the criterion in § 53.860(a)(2) but
decide not to implement the FFD
program described in proposed § 26.604,
licensees and other entities who do not
perform the § 26.603(c) analysis, and
licensees and other entities who
perform the analysis but their analysis
does not demonstrate that their facility
and its operation satisfy the criterion in
§ 53.860(a)(2). These entities must
establish, implement, and maintain an
FFD program under § 26.605(a) either
during construction activities as defined
in § 26.5, or during activities performed
under an ML that allows the assembly,
testing, or both, of a manufactured
reactor. This FFD program implements
all the FFD program requirements in
§ 26.604 plus drug and alcohol testing.
The timing element of the proposed
applicability statement of § 26.605(a) is
equivalent to that for an LWR licensee
or other entity who is performing those
same activities at a facility licensed
under part 50 or 52 and helps provide
assurance that those individuals who
assemble, test, or perform construction
activities as defined in § 26.5 or direct
these activities are fit for duty and
trustworthy and reliable. This is
important because assembly and testing
a manufactured reactor and the
construction and testing of SSCs
required for facility operation require, in
part, adherence to procedures, possible
implementation of unique and precise
assembly techniques, and quality
assurance and controls. Additionally,
SSCs within a manufactured reactor
may not be accessible, testable, or
available for quality assurance and
verification after the reactor is
assembled. This requirement is also
proposed to address solo-assembly
activities that may cause latent failures
and passive SSCs located internal to a
reactor (for example, a fusible link
designed to melt at a particular
temperature to trigger an actuation
mechanism) that are relied upon for safe
operation but cannot be inspected or
tested for proper installation,
configuration, or operation after
installation. A § 26.605(a) FFD program
for these types of activities is equivalent
to the FFD program applicable to the
assembly of the reactor vessel internals
and testing of the SSCs internal to the
reactor at an LWR licensed under part
50 or 52.
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Proposed § 26.605(b) would apply to
the same licensees and other entities as
in proposed § 26.605(a) but before the
loading of fuel onsite into a reactor
vessel; before receiving a manufactured
reactor; or before individuals subject to
part 26 operate, test, perform
maintenance of, or direct the
maintenance or surveillance of securityrelated equipment or equipment that a
risk-informed evaluation process has
shown to be significant to public health
and safety. These entities must
establish, implement, and maintain an
FFD program that implements all the
requirements in § 26.605(a), except
proposed §§ 26.610, ‘‘Sanctions’’;
26.617, ‘‘Recordkeeping and reporting’’;
and 26.619, ‘‘Suitability and fitness
determinations’’; plus additional
requirements due to the increased
radiological consequences presented by
a part 53 commercial nuclear plant as
the licensee readies it for operation.
These additional requirements include
those in subparts C, D, H, and N of part
26, some of which would replace
§§ 26.610, 26.617, and 26.619.
Proposed § 26.605(b) would also
enable the licensee or other entity to
better integrate its facility into the LWR
fleet and Category I fuel cycle facilities
because subparts C, D, and H of part 26
would be required. These subparts
would be required, in part, because it is
expected that: (1) individuals will be
able to work at any part 50, 52, or 53
commercial nuclear plant and will
possess a nuclear safety culture and
desirable qualifications, skills,
expertise, or services; and (2) licensees
and other entities of facilities licensed
under parts 50, 52, and 70 may venture
to construct or operate a facility
licensed under part 53. Therefore, the
implementation of these subparts would
help ensure that all individuals subject
to part 26, except those individuals
subject to an FFD program under
§ 26.604, § 26.605(a), or subpart K of
part 26, would be subject to FFD
programs that provide reasonable
assurance that the individuals are fit for
duty, trustworthy, and reliable.
Proposed § 26.606, ‘‘Written policy
and procedures,’’ would require
licensees and other entities to
implement and maintain an FFD policy
and procedures for their FFD programs.
This section would establish
requirements equivalent to those in
current § 26.403, ‘‘Written policy and
procedures,’’ of subpart K. However, a
principal difference is that proposed
§ 26.606 is written to enable the use of
urine, oral fluid, and hair for drug
testing and screening.
Proposed § 26.606(a)(1) would require
each licensee and other entity to
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provide a written FFD policy statement
to individuals subject to the FFD
program before the individuals are
subjected to behavioral observation and
any FFD program drug and alcohol test.
This would be a protection measure
afforded to individuals subject to the
FFD program to help ensure that they
know what is expected of them before
being subject to the FFD program and
potential consequences should they
violate the FFD policy or procedures.
This requirement would also contribute
to safety and security because
understanding FFD program
responsibilities may enhance an
individual’s safety culture or the
individual may self-select out of the
licensee’s or other entity’s hiring
process.
Proposed § 26.606(a)(2) would require
that the FFD policy statement describe
the performance objectives in § 26.23,
which are the same FFD program
performance objectives required for
facilities licensed under parts 50, 52, or
70. Having a standard performance
outcome based on a licensee or other
entity satisfying the § 26.23 performance
objectives would enhance consistency
in FFD program implementation across
all entities subject to part 26. It would
also generate confidence that
individuals subject to part 26 will safely
and competently perform their duties
and responsibilities and use NRClicensed materials in a manner that will
protect the public health and safety and
common defense and security.
Proposed § 26.606(a)(3) would require
that the FFD policy statement describe
the minimum days off requirements in
§ 26.205(d)(3) or maximum average
work hours requirements in
§ 26.205(d)(7).
Proposed § 26.606(a)(4) would require
the FFD policy statement be written in
sufficient detail to provide affected
individuals with information on what is
expected of them and what
consequences may result from a lack of
adherence to the policy, including those
elements described in § 26.603(b), part
26-required sanctions, and required
medical/clinical treatment and followup testing for FFD policy violations.
This requirement is equivalent to
§ 26.403(a) of subpart K but includes an
additional description of what the
policy statement must include. For
example, the policy would describe the
NRC-required sanctions to help deter
substance abuse and required medical/
clinical treatment and follow-up testing
for FFD policy violations. This
provision would provide a protection
measure by helping the individual get
the assistance they need and help
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ensure that the individual refrains from
substance abuse.
Proposed § 26.606(a)(5) would require
that the FFD policy statement describes
the individual’s responsibilities to
report for work in a physiological and
psychological condition that enables the
safe and competent performance of
assigned duties and responsibilities and
inform a licensee- or other entitydesignated representative when the
individual determines that this cannot
be accomplished.
Proposed § 26.606(b) would require
licensees and other entities
implementing a FFD program in
accordance with subpart M of part 26 to
establish, implement, and maintain
written procedures for their FFD
programs. This requirement would be
equivalent to that in § 26.403(b) of
subpart K.
Proposed § 26.606(b)(1) would
establish requirements for a subpart M
of part 26 FFD program in which the
licensee or other entity implements a
drug and alcohol testing program. This
provision would be equivalent to the
requirements in current § 26.403(b)(1) of
subpart K, but § 26.606(b)(1)(i) through
(iv) proposes additional clarity and
specificity that licensees and other
entities must detail in their procedures
to address new testing methods in
subpart M of part 26 that are not
permitted under the current part 26
framework. Clarity and specificity in
procedural instructions would support
consistent program implementation,
which protects all individuals subject to
the program.
Proposed § 26.606(b)(1)(iv) would
require that if the licensee or other
entity elects to use the HHS Guidelines
for the conduct of drug testing, the FFD
program procedures must include the
name of the specific HHS Guideline and
revision being implemented by the
licensee or other entity and a
description of the specific sections in
the guideline that are being
implemented, including specimen
collections, drug testing, laboratory
procedures, and evaluation of test
results. This requirement would help
ensure the following: the validity and
accuracy of drug testing because the
specimens would be subject to
laboratory testing that has been certified
by the HHS; protection of worker rights
equivalent to the privacy, information,
and due process protections afforded to
Federal workers under the HHS
Guidelines because the HHS Guidelines
are used in the Federally mandated drug
testing programs; consistency in
program implementation because all
individuals subject to the FFD program
would be subject to the same collection,
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testing, and evaluation processes; and
FFD program effectiveness because the
effectiveness of the HHS Guidelines
have been verified by HHS’s National
Laboratory Certification Program
(NLCP). Detailed procedures would
enhance MRO and FFD program
personnel reviews of individual test
results because instructions would be
provided for, in part, the evaluation of
specific test results (e.g., positive,
negative, biological markers), the
conduct of additional testing for invalid
or dilute specimens, and the assessment
of subversion attempts (e.g., adulterated
or substituted). This would benefit FFD
program effectiveness and help prevent
misunderstanding of program
requirements and processes.
Proposed § 26.606(b)(2) would require
licensees and other entities to include in
their written procedures the immediate
and follow-up actions that would be
taken, and the procedures that would be
used, in certain situations specified in
proposed § 26.606(b)(2)(i) through (vi).
Proposed § 26.606(b)(2) would be
equivalent to the requirements in
current § 26.403(b)(2), which provides
the same requirement under an FFD
program for construction for part 50 or
52 licensees and other entities. This
would help ensure the effectiveness of
the FFD program and its consistent
implementation, because part 53
licensed facilities would be
implementing procedures to address the
same requirements and with individuals
who would understand what is
expected of them no matter what part 53
facility they were assigned.
The situation specified in proposed
§ 26.606(b)(2)(i) would arise when
individuals subject to the FFD program
have been involved in the use, sale, or
possession of illegal substances, illegal
drugs, or illicit substances. This
provision would be equivalent to
current § 26.403(b)(2)(i), except that the
phrase ‘‘illegal drugs’’ would be
replaced with ‘‘illegal substances, illegal
drugs, or illicit substances.’’ Illegal
substances would include legal
substances used in a manner
inconsistent with Federal or State law.
The situation specified in proposed
§ 26.606(b)(2)(ii) would arise when
individuals who are subject to the FFD
program are impaired by any substance
or the consumption of alcohol as
determined by behavioral observation or
a test that measures blood alcohol
concentration, as defined in § 26.5.
Except for a few differences, this
provision would be equivalent to
current § 26.403(b)(2)(ii) of subpart K.
The NRC would not include the phrases
‘‘to excess’’ and ‘‘accurately’’ in
proposed § 26.606(b)(2)(ii). Subpart M of
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part 26 is a performance-based
framework that focuses on impaired
human performance, and for alcohol,
impairment is determined by behavioral
observation or by blood alcohol
concentrations exceeding the limits in
§ 26.103, ‘‘Determining a confirmed
positive test result for alcohol,’’ using an
evidentiary breath testing (EBT) device
for alcohol (not whether an individual
drank ‘‘to excess’’). If impairment is
determined by an individual’s behavior,
it must be based on physiological
indications of alcohol impairment.
These indications are well established
in medical, clinical, and law
enforcement organizations, and could be
used by the licensee or other entity
through its procedures and training.11
The NRC would include the phrase
‘‘illegal substances, illegal drugs, and
illicit substances’’ in proposed
§ 26.606(b)(2)(ii) based on operating
experience and the terminology in
current § 26.23(b). There are far more
substances that may cause impairment
than just drugs, drug metabolites, and
alcohol. The phrase ‘‘before or while
constructing or directing construction of
safety- or security-related SSCs’’ in
current § 26.403(b)(2)(ii) would not be
included in proposed § 26.606(b)(2)(ii)
because proposed § 26.606 would apply
during construction, operation, and
decommissioning, if applicable. The
NRC would include the term
‘‘behavioral observation’’ in proposed
§ 26.606(b)(2)(ii) because impairment
can be visibly or audibly observed in an
individual, and individuals subject to
subpart M of part 26 would be trained
in behavioral observation under
proposed § 26.608.
The situation specified in proposed
§ 26.606(b)(2)(iii) would arise when
individuals who are subject to an FFD
program that includes drug and alcohol
testing attempt to subvert the testing
process by adulterating or diluting
specimens (in vivo or in vitro),
substituting specimens, or by any other
means. Except for one difference, this
provision would be equivalent to
current § 26.403(b)(2)(iii). The NRC
would include the phrase ‘‘if drug and
alcohol testing is conducted’’ to address
the licensee or other entity who
implements § 26.604, which does not
require drug and alcohol testing. The
purpose underlying this requirement
has increased in significance since
issuance of the 2008 part 26 final rule
11 By ‘‘well established’’ the NRC means that
there are Federal, State, and non-governmental
organizations with reputable and scientifically
based resources available for a licensee or other
entity to use in its procedures or training to inform
individuals of the physiological indications of
alcohol impairment or intoxication.
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because subversion attempts have
accounted for about one-third of all FFD
policy violations every year since 2016.
The situation specified in proposed
§ 26.606(b)(2)(iv) would arise when
individuals, who are subject to an FFD
program that includes drug and alcohol
testing, refuse to provide a specimen for
analysis or refuse to follow instructions
provided by FFD program personnel.
Except for two differences, this
provision would be equivalent to
current § 26.403(b)(2)(iv). As with
proposed § 26.606(b)(2)(iii), the NRC
would include the phrase, ‘‘if drug or
alcohol testing is conducted,’’ to
account for an FFD program
implemented under § 26.604. The NRC
would include the phrase ‘‘or follow the
instructions provided by FFD program
personnel’’ based on an existing
requirement in § 26.89(c) that the
collector must inform the donor that if
the donor refuses to cooperate in the
specimen collection process, then such
refusal will be considered a refusal to
test and sanctions for subverting the
testing process will be imposed.
The situation specified in proposed
§ 26.606(b)(2)(v) would arise when
individuals who are subject to an FFD
program had legal action taken relating
to drug or alcohol use. This requirement
would be equivalent to current
§ 26.403(b)(2)(v).
The situation specified in proposed
§ 26.606(b)(2)(vi) would be when
individuals subject to an FFD program
demonstrated character or actions
indicating that the individual cannot be
trusted or relied upon to perform those
duties and responsibilities or maintain
access to NRC-licensed facilities, SNM,
or sensitive information. This includes
character traits beyond those attributed
to drug or alcohol use. This proposal
would help ensure that the licensee or
other entity will implement an FFD
program designed to demonstrate
compliance with the § 26.23(c)
performance objective that FFD
programs must provide ‘‘reasonable
measures for the early detection of
individuals who are not fit to perform
the duties that require them to be
subject to the FFD program.’’ An
individual who is not trustworthy and
reliable is not fit to perform or direct the
performance of those duties and
responsibilities or be afforded those
types of access that make the individual
subject to an FFD program.
This proposed requirement also
would help to align the subpart M of
part 26 BOP with the BOP implemented
under § 73.56(f) and proposed § 73.120
and the purpose of the IMP as described
in § 73.55(b)(9) and proposed
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§ 73.100(b)(9).12 The demonstrated
character and actions of an individual
can indicate whether the individual can
be trusted and relied upon to safely and
competently perform assigned duties
and responsibilities or be afforded those
types of access making the individual
subject to the FFD program. This holds
true for any demonstrated adverse
character indication or action on- or
offsite.
The phrase ‘‘character or actions’’
would be used in proposed
§ 26.606(b)(2)(vi) to focus on observed
examples that indicate an individual
subject to subpart M of part 26 may not
be fit for duty or trustworthy and
reliable. Character traits include but are
not limited to personality, temperament,
honesty, carelessness, apathy,
psychosis, and commitment to safety
culture. Assessment of an individual’s
character should consider the potential
for changes in these traits when
compared to a previous baseline.
Actions would include a physical or
verbal demonstration of a character trait
that could call into question an
individual’s fitness, trustworthiness, or
reliability. For example, the individual
does something physically, verbally, or
in writing (e.g., falsifying records,
driving while impaired, or harming or
threatening to harm oneself, others, or
property) that compels another
individual to conclude that the observed
individual cannot be trusted or relied
upon. Unlike the background
investigation and reviews of ‘‘character
and reputation’’ in § 73.56(d)(6) and
(k)(1)(v) and proposed § 73.120, which
are principally retrospective reviews of
an individual and may be based on
third-party information (i.e.,
information from individuals not
subject to NRC requirements), the
‘‘character or action’’ focus of proposed
§ 26.606(b)(2)(vi) would be a present
observation of an individual subject to
the FFD program and performed by an
individual who is also subject to the
FFD program. Whether the information
would be received from an individual
subject to the FFD program or someone
who is not subject to the FFD program,
the licensee or other entity would need
to review this information (i.e.,
determine if the information and its
source are credible) to determine
whether the individual should maintain
authorization.
12 The IMP must monitor the initial and
continuing trustworthiness and reliability of
individuals granted or retaining unescorted AA to
a protected or vital area and implement defense-indepth methodologies to minimize the potential for
an insider to adversely affect, either directly or
indirectly, the licensee’s capability to protect
against radiological sabotage.
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Proposed § 26.606(b)(3) would require
licensees and other entities to address in
their procedures the process, including
the duties and responsibilities of FFD
program personnel, to be followed if an
individual’s behavior or condition raises
an FFD concern. This provision would
also require a process to be conducted
when credible information is received
by the licensee or other entity that the
individual is not fit for duty,
trustworthy, and reliable.
With a few exceptions, proposed
§ 26.606(b)(3) would be equivalent to
current § 26.403(b)(3). Instead of the
phrase ‘‘while constructing or directing
the construction of safety- or securityrelated SSCs’’ in current § 26.403(b)(3),
the NRC would use ‘‘on the NRClicensed facility’’ in proposed
§ 26.606(b)(3) because this provision
would apply during commercial nuclear
plant construction, operation, and
decommissioning, if applicable, in
addition to holders of an ML as
described in § 26.3(f). The requirement
that the roles and responsibilities of
FFD program personnel be described
was developed from current §§ 26.4(g)
and 26.31(b) and operating experience,
which has demonstrated that clear job
descriptions help ensure that
individuals know who is designated by
the licensee or other entity to make
decisions regarding FFD program
implementation and who can be
approached when physiological or
psychological help is needed. This is
principally a protection consideration
afforded to individuals subject to the
FFD program.
The proposed requirement would also
include two conditions not found in
current § 26.403(b) that would clarify
the initiation of the fitness
determination process should an
individual’s behavior or condition raise
an FFD concern. The phrase,
‘‘impairment from any cause that in any
way could adversely affect the
individual’s ability to safely and
competently perform the individual’s
duties,’’ would reflect the § 26.23(b)
performance objective. The condition,
‘‘the receipt of credible information
indicating that the individual cannot be
trusted or relied on to perform those
duties and responsibilities making the
individual subject to this part,’’ would
reflect the § 26.23(a) performance
objective. In either case, as required by
§ 26.23(c), the FFD program must
provide reasonable measures for the
early detection of individuals who are
not fit to perform the duties that require
them to be subject to the FFD program.
Proposed § 26.606(b)(4) would require
licensees and other entities to have
written procedures that address the
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operation and oversight of an onsite or
offsite collection facility. This
requirement would be equivalent to
current §§ 26.403(b) and 26.405(e) and
is developed from § 26.41(b), which
states that each licensee and other entity
who is subject to subpart B of part 26,
shall ensure that the entire FFD program
is audited, which is part of a licensee’s
or other entity’s oversight of the facility,
and § 26.87(a), which states that each
FFD program must have one or more
designated collection sites that have all
necessary personnel, materials,
equipment, facilities, and supervision to
collect specimens for drug testing and to
perform alcohol testing. Having
procedures for the operation and
oversight of the onsite or offsite
collection facility would enhance
consistency in program implementation,
protect individuals subject to testing,
and account for the flexibilities afforded
in the types of biological specimens
than may be collected under an FFD
program subject to subpart M of part 26.
Section 26.606(b)(4), when used with
the PMRP described in § 26.603(d) and
the proposed audit requirement in
§ 26.605(a), would help maintain FFD
program effectiveness and prevent
subversion attempts at facilities that
may not be under the direct day-to-day
oversight of FFD program personnel.
Proposed § 26.606(b)(5) would require
licensees and other entities to have
written procedures that address the
fatigue management requirements in
§ 26.202(b), ‘‘Procedures,’’ and either
§ 26.205(d)(3) or (d)(7).
Proposed § 26.606(b)(6) would require
licensees and other entities to have
written procedures that provide
measures to prevent subversion of drug
and alcohol tests conducted onsite and
offsite. This proposal was developed
from § 26.27(c)(1).
Proposed § 26.607, ‘‘Drug and alcohol
testing,’’ would establish drug and
alcohol testing requirements for
licensees and other entities
implementing proposed § 26.604, at
their discretion, and licensees and other
entities implementing proposed
§ 26.605. Except for a few differences,
proposed § 26.607 would be equivalent
to current § 26.405, which requires
licensees and other entities
implementing an FFD program under
subpart K of part 26 to have a drug and
alcohol testing program that
demonstrates compliance with the
requirements in § 26.405(b) through (g).
The differences are commensurate with
the risk consequences presented by a
part 53-licensed facility as compared to
a part 50 or 52 nuclear power plant.
These proposed requirements would
improve flexibility in the conduct of
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drug and alcohol testing while
maintaining protections afforded to
individuals subject to the FFD program.
Proposed § 26.607(a) would require
licensees and other entities to obtain a
split specimen for all drug tests using
oral fluid or urine for all test conditions
in § 26.607(b), (h) and (j). Neither
current subpart K nor current subparts
B or E of part 26 require a split
specimen. However, the majority of the
LWR fleet uses split specimens for drug
testing and commercially available drug
screening products use a split specimen
technique. Since publication of the 2008
part 26 final rule, the HHS has issued
guidelines for urine and oral fluid that
require split specimens, and the draft
proposed HHS Guidelines for hair
requires split specimens, as well.
The required use of a split specimen
process would protect the individual
because, upon a donor-alleged
discrepant or questionable test result,
the donor may provide permission to
test the split specimen (specimen B) in
an effort to refute the laboratory test
results for specimen A. The requirement
also would enable the MRO to direct
laboratory testing of specimen B if
specimen A were invalid; though the
NRC expects specimens becoming
invalid at the laboratory to be a rare
occurrence as testing would be
conducted in HHS-certified laboratories
with trained collectors. In the event that
a specimen is determined to be invalid,
then the occurrence would likely
warrant further investigation by the
MRO and laboratory to identify the
cause. This protocol would be
equivalent to the special analysis testing
in current § 26.163(a)(2) for dilute
specimens in that additional laboratory
analysis is performed because of a
questionable test result.
If a split specimen is tested by an
HHS-certified laboratory, then the test
result from specimen B must be used as
part of the determination for an FFD
policy violation as required by
§ 26.185(n), ‘‘Evaluating results from a
second laboratory.’’ However, this is not
to say that the test results from
specimen A should be discarded. Since
the HHS-certified laboratory should
report all test results from all specimens
tested to the MRO, like the information
described in § 26.169, ‘‘Reporting
results,’’ test result differences between
specimens A and B can be used to
inform the MRO as to what should be
reported to the licensee or other entity
to either facilitate medical or clinical
assistance for the individual, inform an
FFD-policy violation determination, or
both.
The proposed § 26.607(a) requirement
would also state that if the licensee or
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other entity elects to use a POCTA
device for screening during random
testing or portal area monitoring (e.g.,
pre-access screening), a split specimen
would not need to be taken. The reason
for this exception would be that the
requirements in § 26.607(h)(4) establish
the process to be implemented when a
screening test indicates a presumptive
positive, adulterant, or a discrepant
biological marker, if applicable. This
process includes collecting and testing a
specimen for analysis at an HHScertified laboratory.
Proposed § 26.607(b) would require
the licensee or other entity to subject
individuals identified in § 26.202 to
drug and alcohol testing under the five
conditions listed in § 26.607(b)(1)
through (5). Proposed § 26.607(b) would
be equivalent to current § 26.405(c).
Proposed § 26.607(b)(1) would require
pre-access testing similar to current
§ 26.405(c)(1), which requires testing
before assignment to construct or direct
the construction of safety- or securityrelated SSCs. Unlike current
§ 26.405(c)(1), the proposed requirement
would not include the phrase,
‘‘construct or direct the construction of
safety- or security-related SSCs,’’
because, for licensees or other entities
under part 53, the pre-access test
condition applies to construction,
operation, and decommissioning, if
applicable, to help inform a licensee’s or
other entity’s authorization
determination. The proposal also would
use ‘‘pre-access’’ instead of ‘‘preassignment,’’ which is used in current
§ 26.405(c)(1).
A pre-access test would require the
collection of an oral fluid or a urine
specimen no more than 14 days before
the individual is granted unescorted
access. Although this change has roots
in the 2008 part 26 final rule, which
reduced the period within which preaccess testing must be performed from
60 days to 30 days or less, the 14-day
proposal is based on three lessons
learned from operating experience.
First, the 14-day period would be a
large enough window of time to collect
the specimen and evaluate test results
because licensees or other entities
typically receive laboratory test results
within 5 business days of laboratory
receipt of the biological specimen. At
the same time, the 14-day period would
be small enough to help ensure that the
test results are representative of the
individual’s forensic toxicology before
being granted authorization.
Second, the 14-day window would
enable the licensee or other entity to
conduct an unannounced pre-access
drug and alcohol screening using a hair
specimen or a POCTA. This would help
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prevent an individual from attempting
to subvert the drug and alcohol test by
temporarily abstaining from drug or
alcohol abuse or adulterating or
substituting their specimen to obtain a
non-positive test result.
Third, the NRC does not expect
licensees and other entities licensed
under part 53 to have the large and
periodic influxes of individuals (either
licensee employees or C/Vs) that LWRs
have to support facility operation,
maintenance, engineering design
changes, or nuclear refueling. Therefore,
these licensees or other entities would
not be periodically challenged to in-take
a large workforce within the proposed
14-day pre-access testing window.
Proposed § 26.607(b)(2) would require
the licensee or other entity to conduct
random drug and alcohol testing of all
individuals subject to the FFD program.
With one exception, this proposed
requirement would be equivalent to
current § 26.405(b). Section 26.405(b)
gives licensees and other entities that
implement an FFD program subject to
subpart K of part 26 the option to
impose random drug and alcohol
testing. Proposed § 26.607(b)(2) would
not offer that option because subpart M
of part 26, unlike subpart K, would not
allow a licensee or other entity to
implement a fitness monitoring program
under current § 26.406 instead of a
random testing program. The principal
reasons for not allowing this flexibility
would be that no licensee or other entity
has ever implemented a fitness
monitoring program (i.e., there is no
operating or regulatory experience on
which to judge the effectiveness of a
fitness monitoring program) and the
proposed subpart M framework already
uses behavioral observation to help
ensure FFD program effectiveness.
Supplementing the proposed § 26.609
BOP with an additional observation
technique (i.e., the fitness monitoring
program) would not result in a level of
deterrence or detection equivalent to
that which would be obtained through
behavioral observation and random drug
and alcohol testing.
Proposed § 26.607(b)(2)(i) through (v)
would provide specific requirements for
the conduct of a random testing
program. These paragraphs would be
equivalent to § 26.405(b)(1) through (4),
although with a few differences. The
similar provisions would be proposed in
§ 26.607(b)(2)(i), (b)(2)(iii), and
(b)(2)(iv).
The differing provisions would
include proposed § 26.607(b)(2)(ii),
which would refer to an ‘‘FFD program
procedure’’ instead of the reference to
an ‘‘FFD program policy’’ in
§ 26.405(b)(2) because procedures
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contain the instructions that implement
FFD program requirements, but the FFD
policy need not contain specific
instructions. Section 26.607(b)(2)(ii)
would also require individuals who are
selected for random testing to report to
the onsite collection site, as opposed to
the collection site in § 26.405(b)(2)
because alcohol metabolism necessitates
a relatively timely alcohol test. This
change is also proposed because the
NRC expects that part 53 licensees and
other entities may use a combination of
onsite (for random, for-cause, and postevent testing) and offsite (for pre-access,
post-event, and follow-up testing)
collection facilities for drug and alcohol
testing and may have to afford
reasonable accommodation to certain
individuals, which would add
complexity in the licensee’s or other
entity’s procedurally determined time
period in which an individual must
report to the collection facility.
Another difference from § 26.405(b)
would be proposed § 26.607(b)(2)(v),
which would establish the random
testing rate for the population of
individuals subject to testing. Subpart K
of part 26 does not establish a random
testing rate. The proposed requirement
would be equivalent to current
§ 26.31(d)(2)(vii), which requires that
the sampling process used to select
individuals for random testing provides
that the number of random tests
performed annually is equal to at least
50 percent of the population that is
subject to the FFD program. The NRC
would revise that slightly for proposed
§ 26.607(b)(2)(v) to require a 50 percent
random testing rate for the licensee
employee population and a 50 percent
random testing rate for the C/V
population. The NRC proposes this
change for two reasons.
First, although operating experience
has demonstrated that § 26.31(d)(2)(vii)
helps provide reasonable assurance that
individuals are fit for duty and
trustworthy and reliable through the
detection and deterrence of substance
abuse, this same operating experience
demonstrates that, on many occasions,
the C/V population has been tested at a
rate lower than 50 percent, even though
this population results in the majority of
all FFD policy violations. This bias
occurs because C/Vs are available for
testing only during short periods of time
or periodically throughout the year,
whereas licensee employees are
essentially always available for a test.
A second reason why the NRC is
proposing a different 50 percent random
testing protocol than in the current part
26 requirements is that the flexibilities
afforded to part 53 licensees or other
entities in subpart M of part 26 are not
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afforded to licensees or other entities
that must implement an FFD program
under subparts A through I, N, and O of
part 26. These flexibilities include
enabling the use of a POCTA device to
screen individuals during the random
testing process and the use of offsite
collection facilities for pre-access
testing. The potential reduction in FFD
program effectiveness caused by
licensee or other entity implementation
of these options would be offset by
subpart M requirements that mitigate
possible challenges to the FFD program,
such as the 50 percent random testing
rate for the licensee employee
population and 50 percent random
testing rate for the C/V population.
Proposed § 26.607(b)(3) would require
for-cause testing equivalent to that used
in current FFD programs implementing
§ 26.405(c)(2). The NRC would require
for-cause testing, like random testing, to
be conducted onsite to ensure that the
test is conducted as soon as reasonably
practicable. This is an important
consideration when for-cause testing for
alcohol or using oral fluid for drug
screening or testing because human
metabolism continually lowers the
concentrations of the drugs, drug
metabolites, and alcohol perhaps to
concentrations lower than the initial or
confirmatory testing cutoffs.
Additionally, for facilities that are sited
in geographically remote locations, an
offsite collection facility might be too far
away or not readily accessible.
Proposed § 26.607(b)(4) would require
post-event testing in a manner
equivalent to current § 26.405(c)(3) with
a few adjustments. For part 53 licensees
or other entities, the NRC proposes postevent testing under two conditions:
events involving human errors that may
have caused or contributed to the events
(proposed § 26.607(b)(4)(i)), and events
not involving human error that result in
adverse health consequences or damage
to any safety- or security-related SSC
(proposed § 26.607(b)(4)(ii)). The word
‘‘significant’’ would not be used in
§ 26.607(b)(4)(ii)(A) to describe the
‘‘illness or personal injury’’ as used in
§ 26.405(c)(3)(i) because
§ 26.607(b)(4)(ii)(A) would describe
which illnesses or injuries are covered.
Proposed § 26.607(b)(4)(ii)(B), unlike
§ 26.405(c)(3)(ii), would not use the
word ‘‘significant’’ to describe the
damage to safety- or security-related
SSCs because any damage to safety- or
security-related SSCs would require
testing within four hours of the event
unless immediate medical intervention
precludes the conduct of the test on the
individual(s) who caused or contributed
to the event. Proposed
§ 26.607(b)(4)(ii)(B) also would not use
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the word ‘‘construction’’ as in
§ 26.405(c)(3)(ii) because § 26.607(b)(4)
would apply to construction, operation,
and decommissioning, if applicable.
Proposed § 26.607(b)(4)(i) would
require the licensee or other entity to
define in its procedures the terms
‘‘human error’’ and ‘‘event.’’ These
terms may take on various meanings
and they are not defined in the current
or proposed rule, so the licensee or
other entity would be required to
describe or define these terms to help
ensure consistent implementation of
subpart M of part 26 and that the postevent test condition would be
consistently applied to all individuals
subject to the FFD program. The
§ 26.405(c)(3)(i) requirement that ‘‘the
event is recordable under the
Department of Labor standards
contained in 29 CFR 1904.7, and
subsequent amendments thereto,’’
would not be carried over to proposed
§ 26.607(b)(4). Instead, the NRC
proposes to prescribe the post-event test
conditions in § 26.607(b)(4), in part so
they would not change unless the NRC
amends the requirement.
Proposed § 26.607(b)(5) would require
follow-up testing. This requirement
would be equivalent to current
§ 26.405(c)(4), although the proposed
§ 26.607(b)(5) would further describe
follow-up testing. The NRC proposes to
describe follow-up testing as part of a
series of tests for drugs, alcohol, or both,
which are performed after an individual
subject to part 26 has violated the FFD
policy on substance use or abuse, or the
sale, use, or possession of illegal drugs.
Follow-up testing would be used to
verify an individual’s continued
abstinence from substance abuse. The
NRC would not include a reference to a
follow-up plan as in § 26.405(c)(4)
because the intent of a follow-up plan
is to conduct a series of drug tests,
alcohol tests, or both, to verify
continuing abstinence from substance
abuse. Nevertheless, individuals who
violate an FFD policy on substance use
or abuse, or the sale, use, or possession
of illegal drugs, should have a follow-up
plan that includes a definition of
‘‘abstinence’’ from the medical
professional prescribing the plan.
Proposed § 26.607(c) would provide
additional testing requirements. This
proposed requirement would be
equivalent to § 26.405(d) and would
require implementation of select
requirements from current subpart E of
part 26. The proposed requirements
would govern directly observed
collections, shy bladder situations,
special analysis testing, and alcohol
testing. These requirements would be
necessary to maintain FFD program
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effectiveness equivalent to that
currently implemented by the LWR
fleet.
Proposed § 26.607(c)(1) would require
validity testing and establish the
minimum panel of drugs and drug
metabolites to be tested. This panel
would be the same as those in
§§ 26.31(d)(1) and 26.405(d) because,
based on operating experience from
LWR FFD program implementation, this
panel has been determined to contribute
to a licensee or other entity satisfying
the FFD performance objectives in
§ 26.23(a) through (d).
Proposed § 26.607(c)(1) would differ
from § 26.405(d) because it would
require testing of oral fluid and urine
specimens for validity, including at
least one biological marker (developed
from an HHS Guidelines provision) and
one adulterant (equivalent to current
validity testing for urine specimens in
part 26). Section 26.405(d) requires that
urine specimens collected for drug
testing be subject to validity testing. The
addition of oral fluid validity testing is
important because, just as there are
publicly available kits to subvert a urine
drug test, kits that may be used to
subvert a drug test that uses oral fluid
as a biological specimen are also readily
available.
Proposed § 26.607(c)(2) would
include requirements that already exist
in the part 26 framework that provide
protections for individuals subject to the
FFD program and contribute to testing
effectiveness when collecting and
assessing a urine specimen. Specifically,
current § 26.115, ‘‘Collecting a urine
specimen under direct observation,’’
describes the exclusive grounds for
performing a directly observed
collection and the process to be
followed to protect the privacy of the
individual. Section 26.119,
‘‘Determining ‘shy’ bladder,’’ establishes
the process to be followed when a donor
is not able to produce a sufficient
amount of urine for testing, and
§ 26.163(a)(2) requires special analysis
testing when a specimen is dilute to
help prevent a subversion attempt.
Proposed § 26.607(c)(3) would require
implementation of all the current
alcohol testing requirements in § 26.91,
‘‘Acceptable devices for conducting
initial and confirmatory tests for alcohol
and methods of use,’’ through § 26.103,
‘‘Determining a confirmed positive test
result for alcohol.’’ Using the same
alcohol testing framework for parts 50,
52, 70, and 53 licensees and other
entities would provide for regulatory
consistency, protections for individuals
subject to the FFD program (e.g., the
quality controls and verification applied
to the EBT device), and FFD program
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effectiveness (e.g., accuracy of test
results). For alcohol testing, unlike drug
testing, there is a preponderance of
evidence that correlates blood alcohol
concentrations to impairment and
intoxication. Furthermore, FFD
performance data has demonstrated that
the time-dependent alcohol cutoffs in
§ 26.103 have increased the detection of
individuals who are under the influence
of alcohol. For these reasons, the current
alcohol requirements in part 26 are
proposed for FFD programs under
subpart M.
Proposed § 26.607(c)(4) would
establish additional testing
requirements. This proposal would be
equivalent to current § 26.405(f) for
facilities licensed under part 53 for the
conduct of drug testing. Unlike
§ 26.405(f), proposed § 26.607(c)(4)
would not reference validity screening
and initial drug and validity tests at
licensee testing facilities as this would
be required in proposed § 26.607(c)(1).
Another minor difference between
§ 26.405(f) and proposed § 26.607(c)(4)
would reflect the requirement in subpart
M of part 26 to use an HHS-certified
laboratory for all biological specimens
collected and not just for urine
specimens.
Consistent with § 26.405(f), proposed
§ 26.607(c)(4) would require the use of
an HHS-certified laboratory for all test
conditions listed in § 26.607(b), MROdirected tests, and the testing of a split
specimen. Further, HHS-certified
laboratory test results using urine or oral
fluid would be required for the issuance
of an FFD policy violation and part 26required sanction.
All drug testing would need to be
performed at an HHS-certified
laboratory to help ensure FFD program
effectiveness and to protect the donor
from a false positive test result and an
unwarranted FFD policy violation. The
donor would be protected because
laboratory procedures for specimen
accessioning, testing, custody and
control, and evaluation of test results
and the training and qualification of
laboratory personnel are evaluated by
HHS as part of the NLCP. This provides
assurance that the drug testing results
are accurate and attributed to the donor.
Urine, oral fluid, and hair specimens
may also be screened and tested for
drugs and alcohol as described in
§ 26.607. Drug and alcohol screening
results obtained from urine and oral
fluid specimens collected and analyzed
using a POCTA device and screening
results obtained from a hair specimen or
a portal monitor may only be used as
potentially disqualifying information for
a licensee’s or other entity’s
authorization determination (i.e., used
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to assess the fitness, trustworthiness,
and reliability of the individual). These
screening results may not be used for
the administration of an FFD policy
violation and sanction, except as
proposed §§ 26.607(i)(3) and 26.610 for
subversions, as defined in § 26.5, of the
drug and alcohol screening process.
There are three phrases or
requirements in § 26.405(f) that the NRC
does not propose to use in
§ 26.607(c)(4). The first is the phrase,
‘‘consistent with its standards and
procedures for certification,’’ regarding
the operation of an HHS-certified
laboratory, because the laboratory
would not be HHS-certified if it were
not following ‘‘its standards and
procedures for certification.’’ The
second is the requirement that urine
specimens that yield positive,
adulterated, substituted, or invalid
initial validity or drug test results must
be subject to confirmatory testing by the
HHS-certified laboratory, except for
invalid specimens that cannot be tested.
This requirement would not be used
because, under subpart M of part 26,
licensees or other entities would be
required to use an HHS-certified
laboratory. For a laboratory to be HHScertified, it must follow the HHS
Guidelines and include procedures that
describe when a specimen cannot be
tested. Lastly, the § 26.405(f)
requirement that other specimens that
yield positive initial drug test results
must be subject to confirmatory testing
by a laboratory that demonstrates
compliance with stringent quality
control requirements that are
comparable to those required for
certification by the HHS, would not be
used because subpart M of part 26
would require the use of an HHScertified laboratory.
Proposed § 26.607(c)(4) would require
the licensee or other entity to contract
with a primary and backup HHScertified laboratory. This provision
would help ensure that specimens are
processed and tested to maintain FFD
program effectiveness should the
primary laboratory be unable to perform
specimen testing. This would help
maintain protections afforded to
individuals subject to the FFD program
(e.g., should the donor or MRO request
testing of the split specimen, a different
laboratory could be used). This
requirement also would state that the
primary and backup laboratories must
have a different certifying scientist.
Having a back-up HHS-certified
laboratory and a different certifying
scientist would benefit the program and
donor because the drug testing
instruments, technicians, and certifying
scientist would be independent of the
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primary laboratory testing and review
process. The back-up HHS-certified
laboratory may be of the same corporate
entity as the primary laboratory.
Proposed § 26.607(c)(4) would also
state that the laboratory would be
subject to inspection or audit by the
licensee or other entity and that records
and documents must be provided and/
or able to be photocopied and removed
from the premises to support the
inspection or audit. This requirement
would be equivalent to current
§ 26.41(d) except that laboratories
would not be able to limit the use and
dissemination of documents copied or
taken from the laboratory by a licensee
or other entity. This is necessary to
ensure the continuing effectiveness of
FFD programs, because NLCP findings
and audit results could adversely
impact FFD program effectiveness.
Pertinent information includes and
should not be limited to NLCPidentified weaknesses (e.g., custody and
control, accessioning, instrumentation,
procedures, training, supervision,
review of test results, and resolution of
previously identified corrective actions)
that may impact the effectiveness of
FFD programs.
Proposed § 26.607(d) would help
protect the donor from mistakes made
during the drug and alcohol testing
processes and help ensure FFD program
effectiveness. The rule would require
the licensee or other entity to protect the
individual’s privacy and the integrity of
the specimen and to implement quality
controls to ensure that test results are
valid and attributable to the correct
individual. This requirement would be
equivalent to the first sentence of
current § 26.405(e), except that the word
‘‘stringent’’ was removed from the
phrase ‘‘stringent quality controls,’’
because the word ‘‘stringent’’ is not
defined.
Proposed § 26.607(e) would describe
the requirements for licensees and other
entities that use offsite collection
facilities. Consistent with current
§ 26.405(e), a licensee or other entity
would be able to conduct specimen
collections and alcohol testing at a local
hospital or other facility. Unlike
§ 26.405(e), proposed § 26.607(e) would
not restrict licensees and other entities
to use hospitals and other facilities that
meet the requirements in 49 CFR part
40, ‘‘Procedures for Transportation
Workplace Drug and Alcohol Testing
Programs,’’ because subpart M of part 26
is intended to provide flexibilities
beyond those in the current part 26
framework. Licensees and other entities
may use these Department of
Transportation requirements to inform
their procedures under § 26.606(b)(1) as
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long as the procedures do not conflict
with the requirements in part 26 or the
HHS Guidelines.
Proposed § 26.607(e) would also
require licensees and other entities to
audit offsite collection facilities before
their use and biennially to confirm that
the facility procedures are comparable
to those described in subpart E of part
26 or the HHS Guidelines for urine and
oral fluid. This proposed requirement is
based on current § 26.41(a) and (b). The
§ 26.607(e) audit requirement would be
a program effectiveness consideration
because offsite collection facilities may
not require vigilance of their collectors
(e.g., identification of subversion
attempts), diligence in the protection of
worker rights (e.g., privacy and
specimen custody and control), or
procedural compliance.
The offsite facility used by a licensee
or other entity under proposed
§ 26.607(e) would have to be licensed to
conduct specimen collections and
perform alcohol testing, and be audited,
by the State or a State-designated entity.
This requirement would help provide
assurance of adequate collection facility
performance and may help reduce the
burden on the licensee or other entity
and the collection facility. Crediting a
State audit (or State licensure, oversight,
or regulation) is established in
§§ 26.4(i)(4) and (j), 26.91(e)(5),
26.153(f)(1), and 26.183(a).
Proposed § 26.607(f) would provide
the requirements for initial drug testing.
This provision would be equivalent to
§ 26.405(f) except to account for the
alternative biological specimens that
may be tested under subpart M of part
26. For the testing of all biological
specimens, the licensee or other entity
under part 53 would be required to use
a device that employs an immunoassay
screening technique, or an alternative
technology that the licensee or other
entity has incorporated into its FFD
program through the § 26.603(e) change
control process, that demonstrates
compliance with the requirements of the
U.S. Food and Drug Administration
(FDA) for commercial distribution.
Examples of alternative technologies
include liquid or gas chromatography
and mass spectrometry. Licensees and
other entities would use the § 26.603(e)
change control process to evaluate and
document a change to their collection
and analysis procedures to enable the
use of a better or perhaps more costeffective collection and/or testing
technology. Another difference from
§ 26.405(f) would be changing the word
‘‘urine’’ in § 26.405(f) to ‘‘biological
specimens’’ in § 26.607(f). Lastly,
proposed § 26.607(f) would include the
phrase ‘‘discrepant biological marker’’
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as a drug screening result that must be
analyzed by an HHS-certified laboratory
and evaluated by the MRO to help
inform the MRO’s determination of a
subversion attempt.
Proposed § 26.607(g) would enable a
part 53 licensee to use oral fluid as a
biological specimen for testing. This
requirement would be equivalent to
§ 26.31(d)(5), which enables the MRO to
conduct drug and alcohol testing using
alternative methods, and § 26.405,
which does not preclude the use of oral
fluid specimens for FFD programs that
implement subpart K of part 26
requirements. In order to provide
assurance that drug testing is effective
and protects the worker, § 26.607(g)
would require that the licensee’s or
other entity’s procedures incorporate
the HHS Guidelines or the requirements
in part 26 for the conduct of urine or
oral fluid testing.
The proposed § 26.607(g) requires that
the oral fluid collection device must
have received premarket approval from
the FDA and must not expire before
laboratory testing. Also, the drugs, drug
metabolites, initial and confirmatory
testing cutoffs, and biological markers, if
applicable, must be those established by
HHS for oral fluid drug testing and the
alcohol cutoffs in part 26. If they are not
established by HHS or the NRC for the
paneled drugs and drug metabolites,
then they would be determined and
documented by a forensic toxicologist
review. This forensic toxicologist review
would help ensure that the device
accurately tests for the drug, drug
metabolite, biological markers,
adulterants, and/or alcohol and that the
results from the device are comparable
to those established in the HHS
Guidelines for oral fluid testing.
Proposed § 26.607(h)(1) and (2) would
enable the use of a POCTA device
during the random and pre-access
testing processes. These requirements
are adopted from § 26.97, ‘‘Collecting
oral fluid specimens for alcohol and
drug testing,’’ and § 26.405(f), which
does not preclude the use of oral fluid
testing. To use a POCTA device for
urine, oral fluid, or other biological
indicators (breath, sweat, etc.), a
forensic toxicology review would be
required to ensure that the device is
forensically effective. If the POCTA
device is forensically effective, then the
donor would be reasonably protected
from a false positive test result, the
licensee or other entity would be
reasonably protected from false negative
test results, and the FFD program would
remain effective. For a POCTA device to
be forensically effective, the forensic
toxicologist would need to document an
evaluation that the performance of the
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POCTA device must be comparable to
the requirements in § 26.161(b) for a
urine specimen or the procedures in the
HHS Guidelines for urine or oral fluid,
as implemented by the licensee or other
entity through its procedures.
The use of POCTA for oral fluid and
urine specimens for the pre-access and
random testing processes would be
acceptable because individuals in the
pre-access process would be subject to
an oral fluid or urine specimen
collection and possible drug screening
using a hair specimen, which are both
required to be sent to an HHS-certified
laboratory. For random testing, the
individual would have also been
granted authorization under the AA and
FFD requirements and have been subject
to behavioral observation and physical
protection screening (e.g., verification of
identify, and screening for explosives
and contraband).
Proposed § 26.607(h)(3) would require
that procedures be developed that
ensure the effectiveness of the POCTA
collection process, assessment of the
screening results, and prevention of
subversion attempts. This requirement
would be equivalent to current
§ 26.403(b)(1) and would help ensure
protections afforded to individuals
subject to the FFD program and program
effectiveness. The subpart M of part 26
framework enables the use of POCTA
for random screening of individuals for
any part 53 facility, so the licensee or
other entity should exercise due
diligence and implement risk
management strategies to ensure the
efficacy of random screening and its
contribution to an effective FFD
program.
Proposed § 26.607(h)(4) would
provide that an individual donor who
screens positive (or whose specimen is
invalid or indicates a discrepant
biological marker or adulterant) is
removed from all duties and
responsibilities making the donor
subject to subpart M of part 26. Under
proposed § 26.607(h)(4)(i), the donor
then would be immediately subject to a
drug and alcohol test that provides
quantified confirmatory test results from
which an FFD policy violation may be
issued. Similar to other requirements for
specimen collections, except for
biological specimens analyzed by a
passive detection system, the licensee or
other entity would be required to
implement procedures that ensure that
all specimens collected are uniquely
assigned to the donor (i.e., procedures
that provide for custody and control of
the specimen). If the individual shows
signs of impairment during the POCTA
process, proposed § 26.607(h)(4)(ii)
would require the temporary removal of
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the individual’s authorization until the
MRO reviews the laboratory test
result(s), and interviews the individual,
and a determination of fitness finds that
authorization may be restored. Section
26.607(h)(4) is equivalent to § 26.77(b)
and was informed by the requirements
in §§ 26.419, 26.75(c) and (d), and
26.185(c).
Proposed § 26.607(i) would enable the
collection of hair specimens for drug
testing to supplement pre-access testing
that uses urine or oral fluid specimens.
Hair testing would be a new feature in
the part 26 framework. The NRC
proposes to permit the use of hair
testing for only Schedule I or II drugs or
their metabolites to inform a licensee’s
or other entity’s determination whether
the individual is trustworthy and
reliable. For example, if an individual
stated no prior use of illegal drugs or
potentially addictive habits, a hair
screening test could be performed
during the pre-access process to
ascertain the validity of the individual’s
statement. However, if the HHS-certified
laboratory communicates a laboratoryconfirmed positive test result, an FFD
policy violation may not be
administered. This laboratory
information must be treated as
potentially disqualifying FFD
information, unless the individual
subverts the screening process, in which
case a permanent denial of
authorization must be issued under
proposed § 26.610. To provide
assurance of testing effectiveness and
protections afforded to individuals
subject to the FFD program, proposed
§ 26.607(i) would require that an HHScertified laboratory must be used to
analyze the hair specimen, a forensic
toxicologist must review the licensee’s
or other entity’s hair screening process,
the test kit must be cleared by the FDA,
and hair screening must be conducted
in accordance with the HHS Guidelines.
The forensic toxicologist review would
be necessary if the panel of drug or drug
metabolites to be tested and their cutoffs
are not established by HHS or the NRC
for hair.
Proposed § 26.607(j) would allow the
use of portal area screening for drugs,
alcohol, or both. This provision would
result in a substantial contribution to a
licensee or other entity satisfying the
§ 26.23 performance objectives by
helping ensure that 100 percent of all
individuals who arrive at the NRClicensed facility to perform or direct
those duties and responsibilities or
maintain those types of access making
them subject to the FFD program are fit
for duty and deterred from arriving
onsite in a physiological condition that
may be adverse to safety and security.
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Additionally, screening could be
conducted when an individual exits the
NRC-licensed facility to provide
assurance that substance abuse had not
occurred on the site (see § 26.23(d)). The
screening device could be electronically
linked to temporarily prevent ingress or
egress and could automatically inform
licensee- or other entity-designated
officials of the portal area alarm. The
proposed requirement would enable the
licensee or other entity to use
innovative technologies to maintain
FFD program effectiveness when their
PMRP compels the licensee or other
entity to implement mitigative strategies
to maintain program effectiveness. The
use of portal screening technologies may
also represent cost savings because, for
NRC-licensed facilities that have small
staff sizes or are geographically remote,
passive drug and alcohol screening
technologies could be an innovative
alternative to a random testing program,
although the license or other entity
would need to request and receive an
exemption.
Proposed § 26.607(j) would also
provide that if the portal area screening
instrument detects a substance that
exceeds the instrument’s established
setpoint, the individual would be tested
with either a collection kit that must be
analyzed by an HHS-certified laboratory
or a POCTA. This situational screening
would be equivalent to a for-cause test.
The requirements would not allow an
individual to be rescreened by the portal
area screening instrument following an
initial screening detection that exceeded
an established setpoint in order to
prevent a subversion attempt. Similar to
other drug and alcohol testing
technologies enabled for use by subpart
M of part 26, a forensic toxicology
review would be required before using
passive screening technology to help
ensure the effectiveness of the
instrument by protecting against false
positive or negative screening results,
which would place an unwarranted
burden on the individual, licensee, or
other entity. These instruments and
alcohol screening devices, already in the
marketplace, may also be used to
determine true identity to facilitate
implementation of the FFD BOP, which
may be very practicable at facilities that
operate with small staff sizes.
Proposed § 26.607(k) would enable
the use of a blood specimen for drug,
alcohol, or other testing for certain
medical conditions as determined by
the licensee- or other entity-designated
MRO. This requirement would be
equivalent to current § 26.31(d)(5). The
use of a licensee- or other entitydesignated MRO and not one designated
by a third party, such as an MRO
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employed by an offsite specimen
collection facility, is important because
the MRO must be familiar with the
subpart M of part 26 requirements. To
help ensure testing effectiveness and
protect the worker, the blood test would
need to be conducted by a laboratory
that demonstrates compliance with
quality control requirements that are
comparable to those required for
certification by the HHS, such as a
hospital or clinic certified by the State,
Commonwealth, or territory.
Proposed § 26.607(l) would require
licensee and other entities to use a
Federal custody-and-control form (CCF)
approved by the OMB for the collection
and packaging of a hair, oral fluid, or
urine specimen. This proposed
requirement is based on the CCF
documentation requirements in current
subpart E of part 26 because subpart K
of part 26 does not require the use of a
CCF under § 26.117(e). Additionally,
when using a POCTA device, the
licensee or other entity would be
required to implement a licensee- or
other entity-approved and -maintained
procedure that ensures the reliability of
the tracking, handling, and storage of a
specimen from the point of specimen
collection to final disposition of the
specimen and the reliability of an
identification system to uniquely assign
the specimen to the donor. Both
requirements would help protect the
worker by helping ensure chain of
custody and by contributing to program
effectiveness.
Proposed § 26.607(m) would establish
requirements for the licensee- or other
entity-designated MRO. Section
26.607(m)(1) would be equivalent to
§ 26.405(g), however, the word
‘‘designated’’ would be added to the
first sentence to clarify that the MRO
would be designated by the licensee or
other entity, and not by a third party. As
stated with regard to proposed
§ 26.607(k), this change would clarify
that it is the licensee’s or other entity’s
responsibility, through their designated
MRO, to determine whether an
individual is fit for duty and
trustworthy and reliable. This would be
consistent with the description of FFD
program personnel in current § 26.31(b)
and help provide FFD program
effectiveness and protections to
individuals subject to the FFD program.
The paragraph was also modified from
§ 26.405(g) to address the
determinations of FFD policy violations
and fitness required by subpart H for a
part 53 licensee or other entity that
implements the FFD program described
in § 26.605(b).
Proposed § 26.607(m)(2) would help
ensure that MRO reviews are consistent
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with those MRO reviews conducted at
other NRC-licensed facilities subject to
part 26 and that the MRO maintains
knowledge of drug collection, testing
processes and procedures, and
evaluation of testing results.
The NRC also proposes that if an
MRO performed the duties and
responsibilities in §§ 26.185 and 26.187
for at least three continuous years in the
last 10 years prior to being hired or
contracted by the licensee or other
entity, then the MRO would not need to
repeat the initial training and
examination requirements. The basis for
3 years is that the MRO would have
experienced three annual cycles of
evaluating drug and alcohol test results,
contributed to the FFD annual report to
the NRC, experienced a refueling or
maintenance outage, understood the
duties and responsibilities of
individuals subject to the FFD program
to make informed determinations of
fitness, demonstrated a safety culture
that helps ensure FFD program
effectiveness, and been subject to NRC
inspection. The basis for 10 years is the
relatively long periods between
significant changes to part 26 and the
HHS Guidelines.
Proposed § 26.607(m)(3) would
require that the MRO attend a medicalor clinical-based training session on a
triennial basis. This proposal was
developed from Section 13.1 of the HHS
Guidelines for urine and oral fluid with
two substantial differences: the HHS
Guidelines state that ‘‘requalification
training,’’ including an exam, must be
conducted ‘‘at least every 5 years from
initial certification,’’ whereas the
proposed § 26.607(m)(3) would require a
training session every three years. The
proposed requirements are justified
because changes in societal drug use or
forensic toxicology could occur more
frequently than every 5 years, which
could compel MROs to attend training
in areas of forensic toxicology,
determinations of fitness, or other part
26 technical areas on a more frequent
periodicity than every 5 years to
improve their knowledge and expertise.
Proposed § 26.607(m)(4) would
require the MRO to evaluate drug testing
results by implementing the
requirements in § 26.185 or the HHS
Guidelines through the licensee’s or
other entity’s procedures. This
requirement would help ensure FFD
program effectiveness and enhance
consistency across the commercial
nuclear industry for the evaluation of
drug testing results. This also would
help protect individuals because they
would be subject to the same evaluation
criteria. If § 26.185 provides insufficient
information for an MRO to make a
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determination on a drug testing result
(including adulterant and discrepant
biological markers), the guidance issued
by a State agency in the state in which
the NRC-licensed facility is located,
Federal agency, or nationally recognized
MRO training and certification
organization may be used to inform an
MRO determination. This provision
would ensure that the MRO has the
flexibility to inform their evaluation of
the drug testing results and fitness
determination, if necessary, considering
the drug- and alcohol-related
flexibilities afforded in subpart M of
part 26.
The proposed requirement would also
state that an MRO need not review a
confirmed alcohol positive test result
determined by an EBT device under
§ 26.607(c)(3)(vi) and (vii), which are
equivalent to the current requirements
in §§ 26.101 and 26.103, respectively.
The results of an EBT device are precise
and accurate enough to support the
issuance of an FFD policy violation
without an MRO review of an EBT test
result if the instrument demonstrates
compliance with the requirements in
§ 26.91. The NRC acknowledges that
there are physiological conditions that
may cause an abnormally high blood
alcohol concentration, such as diabetes,
acid reflux, gastroesophageal reflux
disease, and perhaps certain diets (high
protein and low carbohydrates).
However, operating experience has not
demonstrated a compelling need to
require an MRO review of all EBT test
results. For consistency, a licensee or
other entity may elect to require its
MRO to review all EBT test results when
a donor communicates a testing concern
or physiological condition. If the donor
has a testing concern, the occurrence
could be appealed under the proposed
§ 26.613. If the donor presents a
physical condition to the MRO that may
have caused an elevated EBT test result,
the MRO may direct an alternative
testing process (see § 26.607(m)(5))
should it be medically necessary.
Proposed § 26.607(m)(5) would
require the licensee- or other entitydesignated MRO to determine and
approve the use of oral fluid or urine as
an alternative biological specimen when
the donor cannot provide a requested
specimen for testing. This proposed
requirement is equivalent to
§ 26.31(d)(5), which enables the use of
an alternative specimen collection if a
medical condition makes the collection
of the biological specimen difficult. This
determination and the retest must be
completed as soon as reasonably
practicable and documented to support
recordkeeping, auditing, and NRC
inspection.
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Proposed § 26.607(m)(6) would
require that the MRO review all
specimens screened or tested associated
with a drug-related FFD policy
violation. This includes POCTA, split
specimens, and all specimens taken to
resolve a discrepant condition, such as
a possible subversion attempt,
impairment without a known cause, or
a donor-requested or MRO-directed
retest. To resolve a discrepant
condition, the MRO is authorized to test
a specimen for a biological marker,
adulterants, or additional drugs. The
broad scope of this MRO evaluation
would be necessary because of the
variety of different screening and testing
methods that may have been associated
with the FFD policy violation. All
information learned from the conduct of
part 26 drug and alcohol screening and
testing should be used in the evaluation
of an individual’s trustworthiness and
reliability, issuance of a sanction, and
development of a follow-up treatment
and testing plan, if administered.
Proposed § 26.607(n) is equivalent to
current § 26.31(d)(6) and would
establish limits on the screening and
testing of biological specimens. This is
a protection consideration afforded to
individuals subject to the FFD program
and was not provided in subpart K of
part 26. This requirement states that
specimens collected under NRC
regulations may only be designated or
approved for screening and testing as
described in this part and may not be
used to conduct any other analysis or
test without the written permission of
the donor. Analyses and tests that may
not be conducted include, but are not
limited to, deoxyribonucleic acid (i.e.,
DNA) testing, serological typing, or any
other medical or genetic test used for
diagnostic or specimen identification
purposes.
The NRC proposes to require that no
biological specimens may be passively
sampled and analyzed in a manner
different than described in subpart M of
part 26 to ensure workers are protected
from non-consensual passive screening.
The subpart M framework enables
passive detection of drugs and alcohol,
whereas passive detection is not
afforded in subparts A through I, N, and
O of part 26.
Proposed § 26.607(o) is equivalent to
current §§ 26.31(b)(1)(iii)(A) and 26.89
and would require that all specimen
collections be conducted by a licenseeor other entity-designated and -trained
individual. For subpart M of part 26,
this would include onsite specimen
collections, except a collection by a
portal area screening instrument in
§ 26.607(j).
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Proposed § 26.608 would require
licensees and other entities to provide
FFD program training to individuals
subject to the FFD program. The
proposed performance-based § 26.608
requirement was developed from the
prescriptive training requirements in
current § 26.29 and modeled on current
§ 50.120 and the proposed requirements
in §§ 53.725 and 53.830 because there is
no training requirement in subpart K of
part 26.
Proposed § 26.608(a)(1) would require
an FFD training program that includes
the licensee’s or other entity’s FFD
policies and procedures, including
fatigue management, and the
individuals’ FFD program
responsibilities. Individuals who collect
specimens for testing or screening must
also be trained in specimen collector
duties and responsibilities, including, at
a minimum, specimen collection,
custody and control, identification and
response to subversion attempts, and
privacy. The fatigue management
training must include the knowledge
and abilities described in § 26.202(c).
For individuals specified in § 26.4, a
licensee or other entity of a commercial
nuclear plant would be required to use
a SAT as defined in proposed in
§ 53.725. These requirements are based
on requirements in § 26.29(a)(2), (3), (9),
and (10).
Proposed § 26.608(a)(2) would require
training on the BOP. This requirement
would be based on §§ 26.29(a)(8), (9),
and (10) and 26.33. The proposal would
require individuals to be trained in the
detection of behaviors or conditions
related to not only illegal drugs, as in
the current § 26.33 BOP requirements,
but also illicit drugs and substance
abuse onsite and offsite. Also, in
reference to impairment from fatigue or
any cause if left unattended, the phrase
in § 26.33, ‘‘may constitute a risk to
public health and safety or the common
defense and security,’’ would be
replaced in § 26.608(a)(2)(iii) with
‘‘could result in inattentiveness or
human errors,’’ because subpart M of
part 26 is focused, in part, on ensuring
individuals are fit for duty to safely and
competently perform or direct the
performance of assigned duties and
responsibilities.
Proposed § 26.608(a)(2)(iv) focuses on
training to inform individuals that they
are responsible for their own conduct,
as well as observing others. Specifically,
individuals would be trained to
recognize when they feel unable to
safely and competently perform
assigned duties and responsibilities or
act in a trustworthy and reliable
manner. The proposed training
requirement and the proposed reporting
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requirement in § 26.606(a)(5) are in the
interest of safety and security because
the individual is proactively
announcing that assistance may be
necessary. This would be consistent
with the performance objectives in
§ 26.23(b) and (c) where certain
behavior or stress conditions may be
indicative of an individual not being fit
for duty, trustworthy, and reliable.
Proposed § 26.608(a)(3) would help
ensure that individuals subject to the
FFD program understand that FFD
policy violations would result in an FFD
program sanction and that program
information learned or generated by
FFD program implementation would be
used to aide licensee or other entity
authorization determinations and be
shared, as requested, with other
licensees or other entities subject to
parts 26, 53, and 73. This proposed
requirement is equivalent to
§ 26.29(a)(1). Proposed § 26.608(a)(3)
would be a protection measure afforded
to individuals subject to the FFD
program because they would
understand that licensees and other
entities subject to parts 26, 53, and 73
would be informed of, in part, an
individual’s character, reputation, and
ability to follow policies, procedures,
and instructions to safely and
competently perform assigned duties
and responsibilities in a trustworthy
and reliable manner. Fitness-for-dutyrelated information would include drug
and alcohol testing results (not
quantitative testing values), issuance of
any sanctions, FFD-determinations
regarding trustworthiness and
reliability, testing programs, treatment,
and other remedial or corrective action.
Proposed § 26.608(b) would require
individuals be trained and receive a
trainee assessment before pre-access
testing and that refresher training and
trainee assessments be conducted
periodically thereafter. These
requirements would be equivalent to
§ 26.29(c)(1). However, § 26.608(b) was
developed from the SAT-based training
requirements in § 50.120 and training
elements from the annual training
requirements in § 26.29(c)(2). The term
‘‘systems approach to training’’ would
have the meaning in proposed
§ 53.725(c). A trainee assessment would
be the same as in currently required
SAT-based training programs.
Proposed § 26.608(c) would require
licensees and other entities to
periodically evaluate their FFD training
programs and revise them as
appropriate. This training focus is not
required by subpart K of part 26 or
§ 26.29 but is proposed to address the
flexibilities afforded in subpart M of
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part 26. This section would be
equivalent to § 50.120(b)(3).
Proposed § 26.609 would require the
implementation of a BOP. The proposed
requirement would be equivalent to that
in §§ 26.33 and 26.407, ‘‘Behavioral
observation,’’ and would apply during
construction, operation, and
decommissioning, if applicable. Because
subpart M of part 26 would apply
during decommissioning through a
licensee’s IMP, proposed § 26.609(a) and
(b) were developed, in part, from
proposed § 73.100(b)(9) and current
§§ 73.55(b)(9) and 73.56(f) to help
ensure consistency in the conduct of
behavioral observation whether
conducted for FFD or security purposes.
Under the FFD program, the purpose
of the BOP would be to help ensure that
individuals subject to the FFD program
are fit for duty and trustworthy and
reliable to perform or direct those duties
and responsibilities and maintain those
types of access that make the individual
subject to the FFD program. This
assurance is accomplished by requiring
each individual subject to subpart M of
part 26 to be subject to behavioral
observation, and by requiring all
individuals to perform behavioral
observation of others and report FFD
concerns to the licensee- or other entitydesignated representative(s). The intent
of the BOP requirement is not to require
that all individuals be observed at all
times by others; NRC-licensed operators,
maintenance professionals, security
officers, and others routinely perform
solo operations periodically throughout
the day. However, individuals must be
subject to observation while they are
performing or directing the performance
of duties and responsibilities or
maintaining the types of access making
them subject to the FFD program.
Observing behavior only at the
beginning of a work shift is not
sufficient to ascertain whether an
individual is fit for duty, trustworthy,
and reliable. Controlled substances may
have a delayed effect between use (e.g.,
ingestion) and the onset of physiological
or psychological effects, and fatigue
accumulates with time. Behavior must
be continually observed throughout the
work shift to detect any changes from
baseline human performance
characteristics, including mental or
physical health and mannerisms, or any
activities that may indicate that the
individual is not trustworthy and
reliable.
Proposed § 26.609(a) would differ
from §§ 26.33 and 26.407 in that it
would place the responsibility for
performing behavioral observation on
‘‘all individuals subject to this subpart,’’
rather than only those ‘‘individuals
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specified in § 26.4(f) [who] are
constructing or directing the
construction of safety- or securityrelated SSCs’’ in § 26.407 or
‘‘individuals who are trained under
§ 26.29 to detect behaviors’’ in § 26.33 to
improve clarity.
Proposed § 26.609(b) would require
all individuals subject to the FFD
program to report to the licensee- or
other entity-designated representative
any onsite or offsite behaviors or
activities by individuals subject to this
part that may constitute an
unreasonable risk to the safety or
security of the NRC-licensed facility or
SNM or may cause harm to others. The
NRC proposes this description of
reportable conduct because an
individual’s activities (e.g., use of illegal
substances) and communications (e.g.,
hate speech or threats of violence)
offsite are a direct indication of the
individual’s fitness, trustworthiness,
and reliability and must be evaluated as
to whether authorization should be
granted or maintained. Proposed
§ 26.609(b) would include a description
of this conduct instead of the § 26.33
undefined phrase, ‘‘FFD concerns,’’ to
enhance the clarity of the requirement.
This proposed BOP reporting
requirement would include any
information relating to character or
reputation of the individual indicating
that the individual cannot be trusted or
relied upon to perform those duties and
responsibilities or maintain access to
NRC-licensed facilities, SNM, or
sensitive information. This would better
align with the proposed § 73.120 BOP
requirement, which states that each
person subject to behavioral observation
must communicate to the licensee or
applicant observed behaviors or
activities of individuals that may
constitute an unreasonable risk to the
health and safety of the public and
common defense and security. Proposed
§ 26.609(a) and (b) were written broadly
to include offsite conduct that the
reporting individual considers serious
enough to call into question the
character or reputation of the subject
individual.
Proposed § 26.609(c) would require
that licensees and other entities perform
behavioral observation visually, inperson, and, when necessary, remotely
by live video and audible streaming and
capture. This requirement was
developed from the security observation
requirements in § 73.55(e)(7)(i)(B) and
(C), (h)(2)(v), and (i)(2) and (i)(5)(ii).
Conducting an in-person observation of
another individual is the preferred
method to ascertain whether the
observed individual can safely and
competently perform assigned duties
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and responsibilities. When in-person
observations are not feasible (e.g.,
during solo operations), the proposed
requirement would enable the use of
video monitoring. This is addressed, for
example, in proposed § 26.609(d)
regarding NRC-licensed operator
manipulation of reactor controls.
Additionally, certain duties (such as
maintenance activities performed by a
single worker outside of a control room)
may not present an opportunity for
video monitoring; in these situations,
behavioral observation should be
conducted on a sampling basis (i.e., a
planned observation of the work
activity) as outlined in a licensee’s or
other entity’s FFD program.
In situations involving small staff
sizes, facilities sited in geographically
remote locations, or both, additional
observers would enhance the
effectiveness of a BOP. Technological
developments in automated safety and
security systems may enable licensees
or other entities to reduce staff sizes to
10 to 40 percent of the staff size of an
LWR facility licensed under part 50 or
52. Smaller staff sizes may translate into
more solo operations, less teamwork,
fewer peer checks, or infrequent
management oversight of field activities,
leading to fewer behavioral
observations. Therefore, a licensee or
other entity would have fewer
opportunities to observe whether
individuals are fit for duty. Enabling
video and audible streaming and
capture to enhance the BOP would be
consistent with the security-related
behavioral observation requirement in
proposed § 73.120(c)(2)(ii), which
would also enable video conferencing or
other acceptable electronic means
promoting face-to-face interaction for
those individuals working remotely.
Proposed § 26.609(d) would require
that licensees or other entities perform
behavioral observation of NRC-licensed
operators who manipulate the controls
of any commercial nuclear plant
licensed under part 53, remotely by live
video and audible streaming capture for
those part 53 facilities where individual
task loading does not allow for the
effective conduct of behavior
observation in addition to assigned
operational tasks. The purpose of this
paragraph would be similar to that of
proposed § 26.609(c), where the
possibility of in-person observation is
significantly diminished because of solo
operations or because the facility may
only require a minimum staff size
onsite.
Proposed § 26.610 would be
equivalent to § 26.409, ‘‘Sanctions,’’ and
would require the licensee or other
entity to establish sanctions for FFD
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policy violations that, at a minimum,
prohibit the individuals specified in
§ 26.4 from being assigned to perform or
direct those duties and responsibilities
or maintaining authorization making
them subject to subpart M of part 26. To
be consistent with § 26.75, ‘‘Sanctions,’’
the severity of the sanction as described
in § 26.610 would escalate with the
number of occurrences and severity of
the FFD policy violation. The sanction
would be long enough to help deter
future FFD policy violations and
facilitate counseling and treatment
before the licensee reinstates the
individual’s access to the facility. The
NRC proposes this requirement because
the 14-day denial described in § 26.75
may not allow sufficient time for
counseling and treatment based on the
particular FFD policy violation.
Equivalent to § 26.75(c), proposed
§ 26.610 would also require a minimum
5-year denial of access to the NRClicensed facility for certain violations of
the FFD policy within the protected area
of a commercial nuclear plant and by an
individual or individuals who are the
operators of the conveyance to transport
or use formula quantities of strategic
SNM. Equivalent to § 26.75(b), proposed
§ 26.610 would require a permanent
denial of authorization be issued for any
subversion attempt.
Proposed § 26.611 would protect
information collected from FFD program
implementation and would be
equivalent to current § 26.411,
‘‘Protection of information.’’ The
protected information would include,
but not be limited to, privacy and
medical information. Section 26.611
would not include the § 26.411
requirement that FFD programs must
maintain and use the personal
information with the highest regard for
individual privacy because such a
requirement would be unnecessary in
light of the proposed § 26.611(a)
requirement that licensees and other
entities must establish and maintain a
system of files and procedures to
prevent unauthorized disclosure.
Proposed § 26.611(b), although
equivalent to § 26.411(b), would require
licensees and other entities to have all
individuals sign a consent to be subject
to the FFD program before subjecting
the individual to the FFD program (e.g.,
before being subject to a pre-access test
in § 26.607(b)(1), unlike § 26.411(b)).
The purpose of this proposal would be
to enhance protections afforded to
individuals subject to the FFD program
and their knowledge of, in part, why
they are subject to drug and alcohol
testing, behavioral observation,
information collection, MRO reviews,
and other FFD program elements. Like
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the consent required by § 26.411(b), the
consent would authorize disclosure of
the collected information. Consent
would not be needed for disclosures to
the individuals and entities specified in
§ 26.37(b)(1) through (b)(6), (b)(8), and
persons deciding matters under review
in proposed § 26.613, ‘‘Appeals
process.’’
Proposed § 26.613 would be
equivalent to § 26.413, ‘‘Review
process.’’ The proposed title was
changed to an appeal process to clarify
that § 26.613 would be the process
implemented when an individual elects
to appeal a licensee or other entity
determination that the individual had
violated the FFD policy. The proposal
would also require that the process
include a schedule for the completion of
the review of the determination that the
individual had violated the FFD policy.
The NRC proposes this requirement
because operating experience
demonstrates that workers may not be
protected from a continuous review
process that does not result in an
outcome.
Proposed § 26.615 would require
licensees and other entities to perform
audits of the FFD program. The
proposed section would be equivalent to
§ 26.415, ‘‘Audits.’’ Under proposed
§ 26.615(a), audits would be performed
at a frequency that ensures the FFD
program’s continuing effectiveness. This
would be particularly important for FFD
program elements that are not part of
the FFD PMRP required by § 26.603(d).
Corrective actions would be taken as
soon as reasonably practicable to resolve
any problems identified and preclude
recurrence. Proposed § 26.615(b) would
require the subject matter, scope, and
frequency of audits be revised as
necessary to improve or maintain
program performance based on findings
resulting from licensee or other entity
implementation of its FFD PMRP. These
requirements were developed from
appendix B to part 50, ‘‘Quality
Assurance Criteria for Nuclear Power
Plants and Fuel Reprocessing Plants’’;
criterion X, ‘‘Inspection’’; and criterion
XVIII, ‘‘Audits.’’
Proposed § 26.615(c) would be
equivalent to § 26.415(b) and would
enable licensees and other entities to
conduct joint audits or accept audits of
C/Vs so long as the audit addresses the
relevant services of the C/Vs.
Proposed § 26.615(d) would be
equivalent to § 26.415(c) by establishing
requirements for the auditing of HHScertified laboratories. Unlike § 26.415(c),
the proposal would not contain a
reference to the Department of
Transportation drug and alcohol testing
requirements. This would broaden the
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regulatory flexibility afforded to a
licensee or other entity in that they may
use an offsite collection or testing
facility that does not meet the
Department of Transportation
requirements.
Proposed § 26.615(d) would state that
licensees and other entities need not
audit an HHS-certified laboratory if the
licensee’s or other entity’s panel of
drugs and drug metabolites to be tested
is equivalent to the panel by which the
laboratory is certified by HHS or is
subject to the standards and procedures
for drug testing and evaluation used by
the laboratory under the HHS
Guidelines. The NRC would afford this
flexibility because the NRC is aware that
HHS desires to streamline changes in its
guidelines to its panel of drugs and drug
metabolites to be tested. Therefore, if a
licensee or other entity elects to
implement the HHS Guidelines in its
procedures and maintains the minimum
panel of drugs and drug metabolites to
be tested as required by subpart M of
part 26, a licensee or other entity may
still use (and not audit) the HHScertified laboratory because the
§ 26.603(e) change control process
would maintain FFD program
effectiveness.
To help ensure FFD program
effectiveness, § 26.615(d) would also
require that collection facility
procedures are comparable to those
required in subpart E of part 26,
including a proposed requirement that
the offsite facility’s specimen collection
and testing procedures are audited on a
biennial basis, which is also a
protection consideration afforded to
individuals subject to the FFD program.
Conducting this audit on a biennial
basis would be equivalent to that
required in § 26.41(b) and would help
ensure that the specimen collection
process at the facility remains effective.
Proposed § 26.617 would establish
recordkeeping and reporting
requirements equivalent to those in
current § 26.417. However, § 26.617
would require retention of records
pertaining to administration of the FFD
program and FFD performance data
required by § 26.717 until license
termination, which is based on current
§ 26.711(a) because § 26.417 does not
provide for a retention period.
Proposed § 26.617(b)(1) would be
identical to the reporting requirements
in § 26.417(b)(1) regarding the licensee’s
or other entity’s FFD program.
Proposed § 26.617(b)(2) would require
the reporting of annual (i.e., January
through December) program
performance information to the NRC
before March 1 of the following year.
This reporting would be equivalent to
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the annual program performance
requirement in § 26.417(b)(1), and the
March 1 due date is based on the
reporting deadline in § 26.717(e).
Licensees and other entities would be
required to report FFD performance
information using new NRC Forms 893,
‘‘Single FFD Policy Violation Form,’’
and 894, ‘‘10 CFR part 26, subpart M,
Annual Reporting Form for FFD
Performance Information.’’
Proposed § 26.617(c) would require
that FFD-related information be shared
within the commercial nuclear industry
when requested to support
authorization determinations. This
requirement would help individuals
seeking employment by another NRClicensed facility subject to subpart C of
part 26, complete their NRC-required
sanctions and licensee-administered or
-directed drug and/or alcohol abuse
treatment plans before the restoration of
authorization by a licensee or other
entity. Information sharing may also
enhance FFD program effectiveness
because FFD-related lessons learned
from, for example, substance testing,
subversion attempts, and laboratory and
MRO performance must be shared when
requested.
Proposed § 26.619 would require
licensees or other entities to establish a
process to evaluate individuals when
their fitness or trustworthiness and
reliability are in question. Section
26.619 would be equivalent to § 26.419,
‘‘Suitability and fitness
determinations,’’ but, unlike § 26.419,
would apply during the construction
and operation phases. Also, proposed
§ 26.619 would require that a suitability
or fitness determination conducted for
cause be conducted face-to-face. This
proposed requirement is based on
current § 26.189(c); however, unlike
§ 26.189(c), proposed § 26.619 would
not prohibit augmenting determinations
via electronic means of communication.
Instead, § 26.619 would explicitly
permit determinations to be performed
via electronic means, so long as those
determinations are supported by an
appropriately trained individual who is
present in-person with the individual
being assessed.
In considering the current restriction
on the use of electronic means of
communication for determinations of
fitness conducted for cause, the NRC
finds that since publication of the 2008
part 26 final rule, there have been
developments in using electronic means
of communication (i.e.,
‘‘videoconferencing’’) as an alternative
to conducting face-to-face interactions.
To address these considerations, the
NRC contracted the Pacific Northwest
National Laboratory (PNNL), DOE, to
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study whether a medical and mental
health assessment via electronic
communication could be an acceptable
alternative to an in-person, face-to-face
assessment.13 Based on this study, if
electronic means were to be used to
conduct a face-to-face assessment, an inperson element would still be integral to
the assessment process. However, under
certain circumstances, face-to-face
determinations and assessments
conducted as part of an FFD program for
an entity licensed under part 53 (i.e.,
those determinations and assessments
performed in accordance with § 26.619,
§ 26.207, or § 26.211) may be augmented
via electronic communications. Such
remotely conducted determinations and
assessments would be required to be
conducted with someone who is present
in-person with the individual being
assessed and who is trained in
accordance with the requirements of
either § 26.29 and § 26.203(c) or § 26.608
and § 26.202(c). Permitting the use of
electronic communications would help
ensure FFD program effectiveness,
especially in instances where the part
53 commercial nuclear plant is sited in
a geographically remote location or
when the facility has a small staff size.
D. Proposed Changes to Part 26, Subpart
N
Proposed § 26.709 would make the
recordkeeping and reporting
requirements in subpart N of part 26
applicable to licensees and other
entities of facilities licensed under part
53 that elect not to implement the
requirements in subpart M of part 26 or
elect to implement the requirements in
§ 26.605(b).
Proposed § 26.711(c) and (d) would be
amended to make these requirements
applicable to licensees or other entities
described in § 26.3(f). Section 26.711(c)
provides protection to individuals
subject to part 26 by enabling an
individual’s right to review FFD-related
information and correct any inaccurate
or incomplete information. Section
26.711(d) requires, in part, that any
FFD-related information shared with
other licensees or other entities is
correct and complete.
E. Proposed Changes to Part 26, Subpart
O
The vast majority of the proposed
changes to part 26 would be new or
revised substantive provisions that
would establish a regulatory obligation
or prohibition or would be conforming
edits to reflect the addition of part 53.
13 PNNL, Technical Letter Report, ‘‘The Use of
Electronic Communications to Perform
Determinations of Fitness,’’ dated August 2017.
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The only new provision that would not
be substantive, such that violation of it
would not result in a criminal penalty,
would be proposed § 26.601. Therefore,
the NRC proposes to add § 26.601 to the
list of regulations in § 26.825(b) to
which criminal sanctions do not apply.
10 CFR Part 50
A. Section 50.160: Emergency
Preparedness for Small Modular
Reactors, Non-Light-Water Reactors,
and Non-Power Production or
Utilization Facilities
This proposed rule would revise
§ 50.160(b)(3) and (c)(2) to make that
section applicable to applicants and
licensees under part 53. Section 50.160
provides an alternative to other part 50
emergency preparedness requirements
focused on large light-water reactors to
provide an optional emergency
preparedness framework specifically for
small modular reactors (SMRs) and
other new technologies. These
alternative emergency preparedness
requirements adopt a performancebased, technology-inclusive, riskinformed, and consequence-oriented
approach. Commercial nuclear reactor
applicants complying with § 50.160
would be required to submit as part of
the application the analysis used to
determine whether the criteria in
§ 53.1109(g)(2)(i)(A) and (B) are met
and, if they are met, the size of the
plume exposure pathway emergency
planning zone (EPZ). An EPZ bounds
the area surrounding a facility within
which detailed planning is needed to
implement predetermined, prompt
protective actions. The criterion in
proposed § 53.1109(g)(2)(i)(A) is that
public dose, as defined in § 20.1003, is
projected to exceed 10 mSv (1 rem)
TEDE over 96 hours from the release of
radioactive materials from the facility
considering accident likelihood and
source term, timing of the accident
sequence, and meteorology. The
criterion in proposed
§ 53.1109(g)(2)(i)(B) is that predetermined, prompt protective measures
are necessary. These are the same
criteria that are in § 50.33(g)(2)(i)(A) and
(B) and are used to assess the need for
and size of an EPZ in applications under
parts 50 and 52.
Applicants choosing to comply with
§ 50.160 must determine the
radiological releases from the facility
that are evaluated in the determination
of the plume exposure pathway EPZ.
Consistent with other Federal guidelines
such as the Federal Emergency
Management Agency ‘‘Radiological
Emergency Preparedness Program
Manual,’’ issued in 2023, and the
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Environmental Protection Agency ‘‘PAG
Manual: Protective Action Guides and
Planning Guidance for Radiological
Incidents,’’ issued in 2017, applicants
should consider quantitative and
qualitative information on the potential
radiological releases that make up the
spectrum of accidents used to develop
the basis for the applicant’s site-specific
EPZ. This information is derived from
the licensing basis. The NRC plans to
update the risk-informed approach in
RG 1.242 for part 53 while maintaining
its flexibility for using information
already developed and available in
licensing basis documents, including
PRA results, deterministic dose
quantities, accident timing, target set
analyses, mitigation capabilities, and
site-specific factors such as
meteorology.
In its safety analysis report, the
applicant would describe the LBEs
relevant to the facility and would
consider these LBEs as candidates for
the spectrum of accidents used to
develop the site-specific EPZ. The LBEs
assessed include a wide range of events
that are appropriate for considering in
the facility’s emergency preparedness
and response planning. In addition,
§ 50.160(b)(1)(iv)(A)(2) requires
licensees to be capable of implementing
their approved emergency response plan
in conjunction with their safeguards
contingency plan. Radiological sabotage
events are typically factored into EPZ
determinations by considering
consequences to be bounded by LBEs
and by crediting protection against the
DBT in reducing the likelihood of a
release.
The provisions in proposed
§ 53.860(a) provide an alternative to
applicants and licensees by not
requiring them to protect against the
DBT of radiological sabotage in
accordance with §§ 73.55 and 73.100 if
they can demonstrate that the
consequences from unmitigated
radiological sabotage events are below
the safety criteria in proposed § 53.210.
The deployment of some commercial
nuclear plants under part 53 may
involve new scenarios where the source
terms and consequences of sabotagerelated events are not bounded by the
consequences of the unlikely and very
unlikely event sequences analyzed
under subpart C. Accordingly, the NRC
plans to develop guidance for part 53
applicants and licensees choosing to
comply with the alternative emergency
preparedness requirements in § 50.160
to address this new class of reactors. In
Section VI of this document, the NRC is
asking for stakeholder feedback on the
clarity of the regulations and guidance
for various scenarios that might arise in
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implementing graded approaches for
security and emergency planning for
some commercial nuclear plant designs.
B. Appendix B to Part 50: Quality
Assurance Criteria for Nuclear Power
Plants and Fuel Reprocessing Plants
Appendix B to part 50 would be
amended to make it applicable to
applicants and licensees under part 53.
This results in the need for some
revisions to recognize differences in
terminology between parts 50 and 53.
Namely, the term ‘‘design bases,’’ which
is defined in § 50.2, is not used in part
53. For this reason, text is added in both
Section III, ‘‘Design Control,’’ and
Section IV, ‘‘Procurement Document
Control,’’ to refer to ‘‘functional design
criteria, as defined in § 53.020,’’ as the
part 53 equivalent of the term ‘‘design
bases.’’
10 CFR Part 73
A. Section 73.100: Technology-Inclusive
Requirements for Physical Protection of
Licensed Activities at Commercial
Nuclear Plants Against Radiological
Sabotage
Proposed § 73.100 would provide a
performance-based regulatory
framework for the design,
implementation, and maintenance of a
physical protection program and
security organization for certain
commercial nuclear plants licensed
under part 53. The current § 73.55
physical security requirements for
nuclear power reactors licensed under
part 50 and part 52 use a combination
of performance criteria (e.g.,
§ 73.55(b)(1) through (3)) and numerous
prescriptive requirements developed to
achieve performance objectives (e.g.,
§ 73.55(k)(5)(ii)). By contrast, in the
proposed performance-based approach
to physical security for part 53,
performance objectives and
requirements would be the primary
bases for regulatory decision-making,
giving the licensee the flexibility to
determine how to demonstrate
compliance with the established
performance criteria for an effective
physical protection program. This
proposed physical protection program
would provide an optional pathway for
licensees that elect not to demonstrate
compliance with the provisions in
§ 73.55 and do not satisfy the criterion
as described in proposed § 53.860(a)(2).
This proposed physical protection
program would provide that activities
involving SNM are not inimical to the
common defense and security and do
not constitute an unreasonable risk to
the public health and safety.
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Section 73.100(a) would require each
part 53 licensee that elects to
demonstrate compliance with this
section rather than § 73.55 to implement
the requirements therein through its
physical security plan, training and
qualification plan, safeguards
contingency plan, and cybersecurity
plan (referred to collectively hereafter as
‘‘security plans’’) prior to initial fuel
load into the reactor (or, for a fueled
manufactured reactor, before initiating
the physical removal of any one of the
independent physical mechanisms to
prevent criticality required under
§ 53.620(d)(1)). The security plans
would need to identify, describe, and
account for site-specific conditions that
affect the licensee’s capability to satisfy
the requirements of § 73.100. Based on
experience from recent new reactor
licensing reviews, the NRC recognizes
that licensees may seek to receive
unirradiated fuel onsite before carrying
out the security requirements in
§ 73.100. However, these security
requirements would have to be
implemented at some point before
reactor operation to address the
increased risk arising from irradiated
fuel onsite. This proposed rule would
make clear that part 53 applicants and
licensees using § 73.100 may bring
unirradiated nuclear fuel onsite and
protect it in accordance with the NRC’s
requirements for physical protection of
SNM of moderate and low strategic
significance under § 73.67 until initial
fuel load into the reactor (or, for a fueled
manufactured reactor, until initiating
the physical removal of any one of the
independent physical mechanisms to
prevent criticality required under
§ 53.620(d)(1)).
Section 73.100(b) would outline the
general performance objective and
design requirements of the licensee
physical protection program. A
licensee’s program would be required to
provide protection against any
deliberate act within the DBT of
radiological sabotage, including spent
fuel sabotage, which could directly or
indirectly endanger the public health
and safety by exposure to radiation. The
physical protection program is
supported by the AA program,
cybersecurity program, and IMP to
demonstrate compliance with the
general performance objective of
§ 73.100(b).
Section 73.100(b)(2) was developed,
in part, from § 73.55(b)(3). To satisfy the
general performance objective of
§ 73.100(b)(1), the physical protection
program would need to protect against
the DBT of radiological sabotage. The
existing fleet of LWR satisfies this
objective by preventing significant core
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damage and spent fuel sabotage. Some
non-LWR reactor licensees’ physical
protection programs may be designed to
prevent a significant release of
radionuclides from any source.
Therefore, the proposed performance
objective would focus on radiological
sabotage in general, rather than a
specific focus on core damage or spent
fuel sabotage, to be technology inclusive
and allow for flexibility for different
reactor technologies.
Under the proposed § 73.100(b)(2)(ii),
licensees must provide defense in depth
in achieving performance requirements
through the integration of engineered
systems, administrative controls, and
management measures. This
requirement would apply defense-indepth concepts as part of the physical
protection program to ensure the
capability to demonstrate compliance
with the performance objective of the
proposed § 73.100(b)(1) is maintained in
the changing threat environment. The
defense-in-depth philosophy applies to
measures against intentional acts as
required by § 73.100(b), and the designs
of physical security systems should
employ defense in depth through
systems diversity, independence, and
separation under § 73.100(b)(2). The
most common defense-in-depth
measures apply concepts of
redundancy, diversity, independence,
and safety margin to ensure systems
reliability and availability. The defensein-depth philosophy applies to the
design of a physical protection program,
which integrates engineered controls
and administrative controls, to provide
protection against the DBT for
radiological sabotage.
Section 73.100(b)(3) would require
the physical protection program to be
designed and implemented to achieve
and maintain the reliability and
availability of SSCs required for
demonstrating compliance with
specified performance requirements.
These physical protection performance
requirements were informed by
§ 73.55(b) and the Commission’s
Advanced Reactor Policy Statement.
The performance objective of
protecting against the DBT of
radiological sabotage is achieved by the
design and implementation of the
physical protection program,
maintained at all times, with the
following required performance
capabilities proposed in the provisions
in § 73.100(b)(3): intrusion detection,
intrusion assessment, security
communication, security response,
protecting against land and waterborne
vehicle bomb assaults, and access
control portals. The physical protection
program must maintain the reliability
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and availability of SSCs relied upon for
demonstrating compliance with the
performance requirements. The terms
‘‘reliability and availability’’ are
intended to describe defense in depth in
a performance-based manner and would
be critical elements for demonstrating
compliance with the proposed
requirement for protection against the
DBT of radiological sabotage as
described in the proposed
§ 73.100(b)(2).
The first element, ‘‘intrusion
detection,’’ would be provided through
the use of detection equipment, patrols,
access controls, and other program
elements and would provide
notification to the licensee that a
potential threat is present and where the
threat is located.
The second element, ‘‘intrusion
assessment,’’ would provide a
mechanism through which the licensee
would identify the nature of the threat
detected. This would be accomplished
through the use of video equipment,
patrols, and other program elements that
would provide the licensee with timely
information about the threat for use in
determining how to respond.
The third element, ‘‘security
communication,’’ would provide a
mechanism through which the licensee
would communicate the necessary
information to the response force to
ensure effectiveness of the physical
protection program. This would be
accomplished through the redundant,
independent, and diverse design of
physical security and/or plant SSCs
relied on for onsite and offsite security
communications. The continuity and
integrity of communications should
account for the DBT’s ability to affect
the reliability and availability of
communications.
The fourth element, ‘‘security
response,’’ would provide a mechanism
through which the licensee would be
capable of timely security response to
interdict and neutralize threats up to
and including the DBT of radiological
sabotage. The security response may
include the use of onsite armed
responders, law enforcement responders
(local, State, or Federal), or other offsite
armed responders (e.g., licensee
proprietary or contract security
personnel who are positioned offsite), or
a combination thereof, as appropriate.14
14 The NRC’s security regulations for commercial
nuclear power reactors have historically considered
onsite armed responders to be the only acceptable
method for interdicting and neutralizing threats up
to and including the DBT of radiological sabotage.
The proposed rule would permit advanced power
reactor licensees to use any interdiction and
neutralization method, which would be an
extension of the Commission’s position in SRM–
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The licensee must provide protection
against any element of the DBT, to
include those that do not rise to the full
capability of the DBT. Structures,
systems, and components relied on to
provide delay functions must be
designed to provide for timely response
to adversary attacks with adequate
defense in depth. Delay would allow the
licensee to take necessary actions to
counter any attempt by the threat to
advance towards the protected target or
target set element. The overall response
objective would be to place the threat in
a condition from which the threat no
longer has the potential for, or
capability of, doing harm to the
protected target.
The fifth element, ‘‘protecting against
land and waterborne vehicle bomb
assaults,’’ would provide a mechanism
through which the licensee would be
capable of protecting the plant against
the DBT vehicle bomb assault. The
methods that are relied on to protect
against a DBT land vehicle and
waterborne vehicle bomb assault must
be designed to protect the reactor
building, structures containing safety or
security related systems, and
components from explosive effects.
The sixth element, ‘‘access control
portals,’’ would provide a mechanism
through which the licensee would be
capable of detecting and denying
unauthorized access to persons and
pass-through of contraband materials
(e.g., weapons, incendiaries, explosives)
to protected areas. Integrity of the access
control system is maintained through
licensee oversight and ensures that
attempts to circumvent or bypass the
established process will be detected and
access denied.
The proposed performance
requirements would permit the
applicant or licensee to determine how
to design the physical protection
program to protect the plant against the
DBT of radiological sabotage without
SECY–17–0100, ‘‘Security Baseline Inspection
Program Assessment Results and Recommendations
for Program Efficiencies,’’ dated October 8, 2018,
and SRM–SECY–20–0070, ‘‘Technical Evaluation of
the Security Bounding Time Concept for Operating
Nuclear Power Plants,’’ dated June 6, 2024. Under
the proposed rule, a licensee would retain the
responsibility to detect, assess, interdict, and
neutralize threats up to and including the DBT of
radiological sabotage, but would be able to rely on
law enforcement or other offsite armed responders
as a method for fulfilling the required interdiction
and neutralization capabilities. For licensees that
choose to rely on law enforcement to fulfill these
capabilities, the proposed rule would not create any
NRC regulatory jurisdiction over, or requirements
for, law enforcement. In SRM–SECY–23–0021,
‘‘Proposed Rule: Risk-Informed, TechnologyInclusive Regulatory Framework for Advanced
Reactors (RIN 3150–AK31),’’ dated March 4, 2024,
the Commission approved a similar approach to
defend against radiological sabotage.
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prescriptive requirements such as those
currently found in § 73.55. DG–5076,
‘‘Guidance for Technology Inclusive
Requirements for Physical Protection of
Licensed Activities at Commercial
Nuclear Plants,’’ has been developed by
the NRC to describe one acceptable
approach to demonstrate compliance
with requirements proposed in § 73.100.
Section 73.100(b)(4) would require
the licensee to identify target sets in
accordance with § 73.55(f). For nonLWR and SMRs, target sets would be
defined in DG–5071, ‘‘Target Set
Identification and Development for
Nuclear Power Plants,’’ as the minimum
combination of equipment, operator
actions, and/or structures that, if all are
prevented from performing their
intended safety function or prevented
from being accomplished, barring
extraordinary actions by plant
operations, would likely result in a
significant release of radionuclides from
any source (e.g., a release to the
environment exceeding that analyzed in
the DBA licensing basis).
Section 73.100(b)(5) would require
that each licensee perform a site-specific
analysis for the purpose of identifying
and analyzing site-specific conditions
that affect the design of the onsite
physical protection program.
Section 73.100(b)(6) would require
licensees to implement a performance
evaluation program, which would
ensure that a licensee will periodically
test and evaluate the effectiveness of the
physical protection program to protect
against the DBT. This program would
ensure that licensees are able to
demonstrate that the physical protection
program satisfies the response
requirements of § 73.100 and that the
site’s protective strategy effectively
protects against the DBT. Licensee
performance evaluations would include
methods to assess, test, and challenge
the integration of the physical
protection programs functions and
demonstrate the effectiveness of security
plans, licensee protective strategy, and
implementing procedures in accordance
with § 73.100(g).
Section 73.100(b)(7) would require
licensees to implement an AA program
in accordance with § 73.56. Section
73.100(b)(8) would require licensees to
establish, maintain, and implement
protection against a cyberattack based
on either the proposed cybersecurity
program described in § 73.110 or the
program described in existing § 73.54.
Section 73.100(b)(9) would require an
IMP that monitors the initial and
continuing trustworthiness and
reliability of individuals granted or
retaining unescorted access or
unescorted AA to a protected or vital
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area. The IMP must also implement
defense-in-depth methodologies to
minimize the potential for an insider
(active, passive, or both) to adversely
affect the licensee’s capability to protect
against radiological sabotage. Because
no one element of the AA program, FFD
program, cybersecurity program, or
physical protection program, would, by
itself, provide the level of protection
against the insider necessary to
demonstrate compliance with the
performance objective of the proposed
§ 73.100(b), the effective integration of
these programs is a necessary
requirement to achieve defense in depth
against the potential insider.
Section 73.100(b)(10) would require
that the licensee have the capability to
track, trend, correct, and prevent
recurrence of failures and deficiencies
in the implementation of the
requirements of this section. Section
73.100(b)(11) would require the
coordination of the security plans and
associated procedures with other onsite
plans to manage the safety and security
interface during normal or emergency
operations.
Section 73.100(c) was developed from
§ 73.55(c)(7), ‘‘Security implementing
procedures,’’ and § 73.55(d), ‘‘Security
organization,’’ and would outline the
requirements for the composition,
equipping, and training of the security
organization. The purpose of the
security organization is to effectively
implement the physical protection
program. Individuals assigned to
perform physical protection or
contingency response duties must be
trained, equipped, and qualified to
perform assigned duties and
responsibilities.
Section 73.100(d) would establish a
performance requirement for searches of
personnel, vehicles, and materials for
the protection against radiological
sabotage. The requirement describes
broad categories of material (explosives,
firearms, incendiary devices, etc.) to be
detected and prevented from entry into
the protected area; specific items that
will be prohibited would not be
prescribed in the regulation but will be
stated in the licensee security plans
with detailed descriptions being
identified in implementation
procedures.
Section 73.100(e) would require a
training and qualification program,
described in the training and
qualification plan, that ensures
personnel are able to effectively perform
their assigned security-related job
duties. This high-level requirement
would allow flexibility in how the
licensee chooses to train its security
personnel. One method for
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accomplishing this requirement would
be to provide a training and
qualification program that would be
equivalent to appendix B to part 73.
Section 73.100(f) would require
periodic security reviews of the physical
protection program to ensure effective
implementation of the program by
independent individuals. The
evaluation process would provide a
systematized approach for assessing the
physical protection program as a basis
for further development and
improvement. Program reviews should
be designed to ensure that the physical
protection program maintains
effectiveness and demonstrates
compliance with NRC requirements.
Section 73.100(f)(1) was developed from
§ 73.55(m) and would require review of
each element of the physical protection
program. Section 73.100(f)(2) would
require licensees to perform selfassessments of physical protection
program functions to ensure that the
capability to detect, assess, interdict,
and neutralize the DBT of radiological
sabotage is maintained. Section
73.100(f)(3) would require an audit of
the effectiveness of the physical
protection program; security plans;
implementing procedures; cybersecurity
programs; management of the safety/
security interface activities; the testing,
maintenance, and calibration program;
and response commitments by local,
State, and Federal law enforcement
authorities. Section 73.100(f)(4) would
require that results and
recommendations, management
findings, and any actions taken be
documented and maintained to be
available for inspection by the NRC.
These reviews are independent of the
ongoing performance evaluations
described in § 73.100(b)(6) and (g).
Section 73.100(g) would require that
licensee performance evaluations,
described in § 73.100(b)(6), include
methods appropriate and necessary to
assess, test, and challenge the
integration of the physical protection
program’s functions to protect against
the DBT. The performance evaluations
must also address the licensee’s
measures to protect against cyberattacks,
in accordance with the required
cybersecurity plan, and engineered
systems designed to protect against the
DBT standalone ground vehicle bomb
attack.
Section 73.100(h) would establish
performance requirements for
maintaining security SSCs relied on to
perform security functions to protect
against the DBT. It would require that
corrective actions and compensatory
measures be taken by a licensee in
response to a degradation of security
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equipment or failure of the equipment
to perform its intended functions. The
licensee would be required to maintain
the SSCs described in its design and
licensing basis to ensure that they are
reliable and available.
Section 73.100(i) would establish
requirements for the suspension of
security measures in response to
emergency and extraordinary
conditions. The requirements of this
paragraph, which were developed from
§ 73.55(p), would be intended to
provide flexibility to a licensee for
taking reasonable actions that depart
from a security plan in an emergency
when such actions are immediately
needed to protect the public health and
safety and no action consistent with
license conditions and TS that can
provide adequate or equivalent
protection is immediately apparent in
accordance with proposed § 53.740(h).
Section 73.100(j) would establish
requirements regarding the inspection,
retention and maintenance of records
required to be kept by the NRC
regulations, orders, or license
conditions. These proposed
requirements are developed from
§ 73.55(q).
B. Section 73.110: Technology-Inclusive
Requirements for Protection of Digital
Computer and Communication Systems
and Networks
Section 53.860 would require that a
licensee establish, implement, and
maintain a cybersecurity program in
accordance with § 73.54 or § 73.110.
Section 73.110 would establish
requirements for the development and
maintenance of a cybersecurity program
for commercial nuclear plants licensed
under part 53. This proposed section
would implement a graded approach to
determine the level of cybersecurity
protection required for digital
computers, communication systems,
and networks. The proposed new
section is informed by: (1) the operating
experience from power reactors and fuel
cycle facilities; and (2) the existing
§ 73.54 framework, which addresses
some of the basic issues for
cybersecurity regardless of the type of
reactor. Differences between the § 73.54
requirements and those proposed in
§ 73.110 are primarily based on the
implementation of a consequence-based
approach to cybersecurity that provides
flexibility to accommodate the wide
range of reactor technologies to be
assessed by the NRC. A graded approach
based on consequences is intended to
account for the differing risk levels
among reactor technologies.
Specifically, the proposed new section
would require licensees to demonstrate
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protection against cyberattacks in a
manner that is commensurate with the
potential consequences from those
attacks.
Under proposed § 73.110(a), licensees
would need to ensure that digital
computer and communications systems
are adequately protected against a
potential cyberattack that would result
in: (1) a scenario where the cyberattack
leads to offsite radiation doses that
would endanger public health and
safety (i.e., the resulting consequence
exceeds the reference dose values in
§ 53.210); or (2) a scenario where the
cyberattack adversely impacts the
physical security digital assets used by
the licensee to prevent unauthorized
removal of material or radiological
sabotage. Security digital assets would
include those used for nuclear MC&A.
The proposed § 73.110(b) would
require licensees to protect the
communication system and networks
associated with the functions described
in § 73.110(a)(1) and (a)(2) from
cyberattacks. To accomplish this, the
licensee would establish, implement,
and maintain a cybersecurity program
for protecting digital assets within the
scope of § 73.110 that would make use
of risk insights, including threat
information, and would consider the
resulting level of consequences of the
threats. If the outcome of the assessment
by the licensee under § 73.110(b)(1)
revealed that a potential cyberattack
would not compromise any digital
assets that support safety and security
functions, and thus would not result in
the consequences listed in § 73.110(a)
(e.g., would not exceed the reference
dose values), then only a narrow set of
the cybersecurity program requirements
in § 73.110(d) and (e) would apply. For
example, the licensee would only need
to develop a cybersecurity program that
implements the requirements dealing
with:
• Analyzing modifications of any
asset before implementation to see if
they demonstrate compliance with the
potential consequences in § 73.110(a);
• Ensuring employees and contractors
are aware of cybersecurity requirements
and have some level of cybersecurity
training;
• Evaluating and managing
cybersecurity risks to the plant;
• Reviewing the cybersecurity plan
for any required changes; and,
• Retaining records of the
cybersecurity plan along with any plan
changes.
Section 73.110(c) through (e) were
developed from § 73.54(a)(2), and (c)
through (h), respectively.
The proposed requirements would
address the need for the licensee to
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develop a cybersecurity program that
implements a defense-in-depth
protective strategy as required by
proposed section § 73.110(d)(2). A
defense-in-depth protective strategy for
cybersecurity is represented by
collections of complementary and
redundant security controls that
establish multiple layers of protection to
safeguard critical digital assets. Under a
defense-in-depth protective strategy, the
failure of a single protective strategy or
security control should not result in the
compromise of safety and security
functions.
C. Section 73.120: Access Authorization
Program for Commercial Nuclear Plants
Section 73.120 would address AA for
certain commercial nuclear plants
licensed under part 53. The proposed
language in § 73.120 would provide an
alternate approach to the existing
framework for AA under §§ 73.55,
73.56, and 73.57, commensurate with
risk and consequences to public health
and safety. It would be available to part
53 applicants and licensees who
demonstrate in an analysis that the
offsite consequences of a DBE satisfy the
criterion defined in § 53.860(a)(2)(i) (i.e.,
would not exceed the offsite dose values
in § 53.210(b)). The proposed
requirements in § 73.120 would be
similar to the existing AA program
elements for those NRC licensed
facilities issued additional security
measures (ASMs) orders and for
materials licensees under § 37.21.
Applicants not satisfying the criterion
would need to establish, implement,
and maintain a full AA program,
including an IMP, in accordance with
§ 73.56.
Proposed § 73.120(a) would be based
on an applicant satisfying the eligibility
criterion in § 53.860(a)(2)(i). Section
73.120(b) would identify the categories
of individuals who would be subject to
an AA program in accordance with this
section. The applicability statement in
§ 73.120(b)(1)(i) would encompass
individuals whom the licensee intends
to grant unescorted access to the
facilities’ most sensitive areas,
consistent with § 73.56(b)(1)(i) for
power reactors and the ASM orders and
license conditions issued to any NRC
licensed facility or material licensee.
Sections 73.120(b)(1)(ii) through (iv)
would be consistent with
§ 73.56(b)(1)(ii) through (iv),
respectively. The program would
include individuals who may be onsite
or offsite (e.g., remote operators or
information technology staff) and have
virtual access to important plant
operational and communication systems
based upon assigned duties and
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responsibilities. An individual who has
remote access to plant equipment and
communication systems may have
trusted privileges greater than the
personnel at the plant site. Section
73.120(b)(1)(iii) would state that offsite
law enforcement personnel on official
duty would not be subject to the
licensee AA program.
Section 73.120(c) would provide
general performance objectives and
requirements largely consistent with the
AA program requirements for nuclear
power reactors under § 73.56 and would
provide licensees and applicants the
flexibility in establishing their AA
program to demonstrate compliance
with various performance objectives.
Section 73.120(c)(1) would include
background investigation requirements
consistent with § 37.25, as well as ASMs
and license conditions that are applied
to non-power reactor licensees.
Background investigations include
important elements to establish the
trustworthiness and reliability of an
individual, such that they do not
constitute an unreasonable risk to
public health and safety or the common
defense and security. These include the
following: (1) personal history
disclosure, (2) verification of true
identity, (3) employment history
evaluation, (4) unemployment/military
service/education, (5) credit history
evaluation, (6) character and reputation
evaluation, and (7) Federal Bureau of
Investigation criminal history record
check.
Section § 73.120(c)(2) would establish
behavioral observation requirements,
which are an awareness initiative for
recognizing behaviors adverse to the
safe operation and security of the
facility through observing the behavior
of others in the workplace and reporting
aberrant behavior or changes in
behavior that might reflect negatively on
an individual’s trustworthiness or
reliability. Maintaining behavioral
observation would assist and/or
improve worker safety and reduce the
risk of an insider threat. This proposed
requirement in § 73.120(c)(2) would be
a scaled version of the full BOP required
under § 73.56(f).
Section § 73.120(c)(2) would provide
licensees greater flexibility to
implement behavioral observation
options for individuals granted
unescorted access to the commercial
nuclear plant’s protected area. Such
options on reporting questionable
behavior may include a program similar
to the Department of Homeland
Security’s program, ‘‘If you see
something, say something,’’ or to a
corporate behavioral awareness
program. Commensurate with the
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potential lower safety and security risks
of a commercial nuclear plant that
meets the criterion in § 53.860(a)(2)(i),
§ 73.120(c)(2) would not require the
establishment of a comprehensive
training program for behavioral
observation (i.e., initial and refresher
training including knowledge checks) as
required for power reactors under
§ 73.56 and part 26. Under
§ 73.120(c)(2)(ii), behavioral observation
would be able to be performed in-person
or remotely by video, and identified
behavior of concern would need to be
reported to plant supervision. The
remote access alternative to face-to-face
interactions provides substantial
flexibility for licensees and applicants.
Any video conferencing or other
acceptable electronic means promoting
face-to-face interaction for those
individuals working remotely would
demonstrate compliance with this
regulation.
Section 73.120(c)(3) captures and
maintains the self-reporting of legal
actions as an essential performance
element to enhance the licensee’s
behavioral observation initiative similar
to the current requirements under
§ 73.56(g), assuring that personnel who
are granted and who maintain
unescorted access are trustworthy and
reliable.
Section 73.120(c)(4) would provide a
scalable approach for granting and
maintaining unescorted access. One
component not included from § 73.56 is
the need for a psychological assessment
and reassessment under § 73.56(e) for
granting unescorted access and
§ 73.56(i)(v)(B) for individuals who
perform one or more of the job functions
described in § 73.120(b)(1)(ii) for
maintaining unescorted access.
Moreover, the requirement would
permit criminal history updates to be
completed within 10 years of the last
review, compared to the three- or fiveyear reinvestigation periodicity for
personnel at an operating commercial
nuclear plant. In addition, no credit
check re-evaluation would be required
for these individuals.
The continued need to maintain
unescorted access would be evaluated
on an annual basis by the reviewing
official. Guidance in DG–5074, ‘‘Access
Authorization Program for Commercial
Nuclear Plants,’’ would specify that this
evaluation should be based on a
compilation of personnel interactions as
described in the licensee’s or applicant’s
policy and procedures for behavioral
observation and the maintenance of an
approved AA list.
Section 73.120(c)(5) would require
licensees and applicants to determine
when a person no longer requires the
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need for unescorted access or no longer
satisfies the AA requirement found
within this section. Guidance in DG–
5074 would further explain that
licensees have the flexibility to
terminate unescorted access to specific
areas of the site if individuals lack the
continued need for that access to
perform their duties and
responsibilities.
Section 73.120(c)(6) would be
consistent with the purpose of § 37.23(e)
and would include the individual’s
right to correct and complete
information as required under
§ 37.23(g). The section would include a
requirement for designating a reviewing
official. The language would provide
clarity regarding the roles and
responsibility of a reviewing official,
who would be the only individual
authorized to make unescorted access
determinations.
Section 73.120(c)(7) would align with
the corresponding requirements under
§ 37.23(f), and § 73.120(c)(8) would
align with the corresponding
requirements under § 37.31. These
requirements would encompass the
roles and responsibilities for licensees,
applicants, and if applicable, the
contractor/vendors to establish,
implement, and maintain a system of
files and records to ensure personal
information is not disclosed to
unauthorized persons.
Section 73.120(c)(9) would align with
the requirements of § 37.33. Section
73.120(c)(10) would require licensees,
applicants, and contractors or vendors
to maintain the records that are required
by the regulations in this section and
retain them for a period of 3 years after
the record is superseded or no longer
needed. The record retention period of
three years would be consistent with
§ 37.23(h), contrasting with the five-year
retention period under § 73.56(o).
Records maintained in any database(s)
would need to be available for NRC
review, consistent with the
requirements found under
§ 73.56(o)(6)(ii).
VI. Specific Requests for Comments
The NRC is seeking advice and
recommendations from the public on
this proposed rule. We are particularly
interested in comments and supporting
rationale from the public on the
following:
Part 26—Fitness for Duty Program
1. The proposed rule under
§ 26.603(c) would enable a licensee or
other entity to implement an FFD
program under proposed § 26.604, ‘‘FFD
program requirements for facilities that
satisfy the § 26.603(c) criterion,’’ if the
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licensee or other entity performs a sitespecific analysis to demonstrate that the
facility and its operation satisfy the
criterion in § 53.860(a)(2).
Should the NRC consider replacing its
proposed § 26.603(c) criterion
referencing § 53.860(a)(2) with an
alternative requirement that if the
commercial nuclear plant is of the class
described in § 53.800, ‘‘Facility
licensees for self-reliant-mitigation
facilities,’’ and either § 53.800(a)(1) or
(2) is satisfied, then drug and alcohol
testing would not be required? This
proposal would align the § 26.603(c)
criterion with that proposed in the NRClicensed operator regulatory framework
of part 53. Please provide your
considerations and rationale for your
recommendation.
Should the NRC also consider making
a conforming change to the proposed
§ 73.120 criterion used for the AA
program? Please provide your
considerations and rationale for your
recommendation.
Part 26—Technology-Inclusive
Approaches to Fatigue Management
Requirements Applicable to Unit
Outages
In establishing the outage minimum
days off requirement of § 26.205(d)(4),
the NRC’s objective was to ensure that
individuals performing the duties
described in § 26.4(a)(1) through (a)(4)
have sufficient periodic long-duration
breaks to prevent cumulative fatigue
from degrading their ability to safely
and competently perform their duties.
In addition to the science of fatigue
management, the NRC considered
several factors in establishing the
existing requirements. These additional
factors were practical and safety
considerations associated with the
management of refueling outages for
large LWRs, including the following: (1)
the typical duration and frequency of
outages; (2) the availability of contract
personnel to perform the work; (3) the
risk presented by the outage work while
the reactor is shut down; and (4) the
controls applied to the work that may
limit the potential for latent errors to
challenge reactor safety when the
reactor is returned to power. The details
of such considerations may differ for
new reactor technologies or designs.
Such considerations may not be relevant
for some reactor designs (e.g., reactors
capable of on-line refueling) and there
may be additional, more pertinent
factors to consider for other designs.
The NRC is seeking stakeholder input
on whether alternative fatigue
management requirements applicable to
outages should be adopted to support
technology-inclusive approaches that
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would be appropriate to support the
licensing and regulation of future
commercial nuclear plants. Please
provide your considerations and
rationale for your recommendation.
Part 26—Draft Regulatory Guidance
Approach for Fatigue Management
In support of this proposed rule, the
NRC has issued DG–5078, ‘‘Fatigue
Management for Nuclear Power Plant
Personnel at Commercial Nuclear Plants
Licensed Under 10 CFR part 53.’’ This
DG describes methods the NRC staff
considers acceptable for addressing
certain aspects of FFD programs at
commercial nuclear facilities licensed
under part53.
The NRC staff also intends to
eventually transition this draft guide
into an update to RG 5.73, ‘‘Fatigue
Management for Nuclear Power Plant
Personnel,’’ or the development of a
new RG. At this point, NRC staff is
considering four options for future RG
development:
• Option 1: Amend the existing RG.
The NRC may develop an updated
version of RG 5.73 that continues to
endorse (with clarifications, additions,
and exceptions) the guidance contained
in NEI 06–11, ‘‘Managing Personnel
Fatigue at Nuclear Power Reactor Sites,’’
Revision 1, and incorporates the topics
discussed within DG–5078 as new NRC
staff positions in section C of RG 5.73.
• Option 2: Issue a new RG specific
to part 53 licensees. The NRC may
develop an entirely new RG applicable
specifically to facilities licensed under
part 53. This new RG would capture the
guidance contained in DG–5078 and
incorporate existing guidance (e.g.,
selected guidance in RG 5.73 and NEI
06–11) that is considered to be
technology inclusive in nature. The
existing guidance (i.e., RG 5.73) would
remain in place as the guidance for
facilities licensed under parts 50 and 52.
• Option 3: Review and potentially
endorse new or revised industrydeveloped guidance. The NRC may
engage with the industry regarding a
potential update to industry guidance
document NEI 06–11 or the
development of new, separate industrydeveloped guidance specific to facilities
licensed under part 53. The NRC would
then review the new or revised
industry-developed guidance within the
NRC’s RG process, which includes
opportunities for public participation.
New or revised industry-developed
guidance could incorporate DG–5078 or
propose alternatives for the NRC to
consider.
• Option 4: Develop a comprehensive
revision of the existing RG. The NRC
may develop a more comprehensive
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revision of RG 5.73 that would
explicitly detail all NRC positions
reflected in the existing RG (including
those endorsed positions currently
contained in NEI 06–11, Revision 1),
along with the guidance of DG–5078.
Such a revision would thereby be a
‘‘stand-alone’’ document, without
reference to or explicit endorsement of
separate, industry-developed guidance.
The NRC is seeking stakeholder input
regarding which of the four options
listed above would be optimal (or
whether there are other options that the
NRC should consider). Please provide
your considerations and rationale for
your recommendation.
Part 53—Overall Organization
Part 53 is structured as one framework
with subparts providing technical,
licensing, and administrative
requirements for the various stages of
the life cycle of a commercial nuclear
plant. The organization of part 53 in this
manner puts a complete set of
requirements for each stage of the life
cycle in a separate subpart with
additional subparts for licensing and
administrative requirements.
The NRC is seeking comment on the
proposed organization of the
requirements in part 53 and possible
improvements to how specific
requirements (e.g., examples of which
specific sections) could be consolidated
or otherwise reorganized to make the
rule clearer or more concise.
There are numerous references in
proposed part 53 to other NRC
regulations. Examples of such references
include those in proposed § 53.610 to
NRC regulations related to radiation
protection (part 20), FFD (part 26),
physical security (part 73), and MC&A
(10 CFR part 74, ‘‘Material Control and
Accounting of Special Nuclear
Material’’) for facilities receiving
byproduct or SNMs.
The NRC is seeking comment on
whether such references to other
regulations in various sections in the
proposed part 53 provide benefits to
applicants and licensees, or to other
stakeholders seeking to understand the
regulatory framework under part 53, or
whether such references could be
removed to reduce the length of part 53.
Part 53, Subpart B—Comprehensive
Risk Metrics
The NRC is proposing to require the
use of comprehensive risk metrics and
associated risk performance objectives
as one of several performance standards
in part 53. Comprehensive risk metrics
could include a risk metric or set of risk
metrics that approximate the total
overall risk from the facility to the
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extent practicable. Associated risk
performance objectives are
preestablished values indicative of the
comprehensive risk metrics that are
used during risk-informed decisionmaking to gauge plant safety.
Specifically, comprehensive risk metrics
and associated risk performance
objectives would provide one element of
the safety criteria for LBEs other than
DBAs in the proposed § 53.220.
Comprehensive risk metrics, in the form
of the IEFR and the ILCFR, and
associated risk performance objectives,
in the form of the QHOs of 5×10¥7 per
year and 2×10¥6 per year, respectively,
were similarly used in the LMP
methodology to ensure that other
evaluation criteria were conservatively
defined and as a tool for focusing
attention on matters important to
managing the risks posed by nuclear
power plants. The use of such
comprehensive risk metrics and
associated risk performance objectives
in an integrated risk-informed decisionmaking process is similar to that used in
RG 1.174, ‘‘An Approach for Using
Probabilistic Risk Assessment in RiskInformed Decisions on Plant-Specific
Changes to the Licensing Basis,’’
Revision 3.
The NRC is seeking comment on the
use of comprehensive risk metrics and
associated risk performance objectives
in part 53 as one of several performance
standards. The IEFR and ILCFR and the
QHOs represent comprehensive risk
metrics and associated risk performance
objectives that the NRC has used for
decades in a variety of capacities. What
other performance standards could be
used to address the comprehensive risks
posed by proposed commercial nuclear
plants? Please provide your
considerations and rationale for your
recommendation.
If an applicant proposes a novel
approach to comprehensive plant risk
and the NRC approves the approach,
should the resulting NRC-approved
comprehensive plant risk metrics and
associated risk performance objectives
be codified or otherwise memorialized
over time and, if so, how?
Part 53, Subpart B—Defense in Depth
Proposed § 53.250 would establish
requirements based on the longstanding
NRC philosophy of providing defense in
depth to address uncertainties
concerning the design, operation, and
performance of commercial nuclear
plants during LBEs.
The NRC is seeking comment on the
inclusion of the proposed requirements
to assess and provide defense in depth.
The NRC is also seeking comment on
whether to include specific provisions
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in § 53.250 and subpart B to more
explicitly address the possible role of
inherent characteristics of some SSCs in
preventing or mitigating unplanned
events. The proposed § 53.250 is
worded to preclude relying on a single
engineered design feature to address the
range of LBEs other than DBAs, which
could possibly allow crediting inherent
characteristics without further lines of
defense. How could possible inherent
characteristics of SSCs be considered in
the proposed requirements in § 53.250
or in any alternative requirements for
defense in depth provided in response
to this item? Please provide your
considerations and rationale for your
recommendation.
Part 53, Subpart C—Probabilistic Risk
Assessment
Current consensus PRA standards
provide processes for appropriately
defining the scope of a PRA and
determining applicability of supporting
requirements to suit the specific needs
of a given applicant under proposed
part 53. In addition to assessing other
aspects of PRA acceptability such as
PRA peer reviews, NRC determinations
of the acceptability of such PRAs would
assess the appropriateness of the
applicant-defined scope as part of
determining the applicability of a
consensus PRA standard supporting an
application. This approach is consistent
with the current state of practice and
offers appropriate flexibility for PRAs to
be developed and assessed based on the
application they are used to support,
which includes consideration of how
PRA results and insights are relied
upon, together with factors such as
safety margin, simplicity of design, and
treatment of uncertainty.
The NRC is seeking comment on what
additional guidance, if any, is needed
regarding PRA acceptability for Part 53
applicants and licensees.
Part 53, Subparts C and D—Earthquake
Engineering
Proposed § 53.480 would establish
requirements related to seismic design
considerations. This proposed section is
intended to provide a clear connection
between siting activities and seismic
design activities and to support various
approaches to presenting seismic
hazards and addressing those hazards in
designs. The proposed requirements are
intended to provide sufficient flexibility
to allow approaches like those currently
in parts 50 and 100 or approaches that
might be endorsed by the NRC in the
future that could incorporate more risk
insights from PRAs.
The NRC is seeking comment on
whether the proposed requirements for
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earthquake engineering provide
appropriate flexibility in addressing
seismic risks while also ensuring that
the regulations continue to adequately
address seismic hazards. Please provide
your considerations and rationale for
your recommendation.
Part 53, Subpart E—Construction and
Manufacturing
1. Proposed § 53.610(b)(1)(iii) would
require procedures that describe how
construction will be controlled so as not
to impact other features important to the
design (e.g., dewatering, slope stability,
backfill, compaction, and seepage).
The NRC is seeking comment on
whether such specific requirements are
useful or whether these requirements
could be met through other
requirements proposed in part 53 or
already present in other relevant
regulations (e.g., quality assurance
requirements in appendix B to part 50).
Part 53, Subparts E and H—
Manufacturing Licenses
1. The proposed requirements
governing manufacturing are set forth in
subpart E, and the proposed
requirements governing the licensing
processes are contained in subpart H.
Some of the proposed requirements,
including provisions related to the
loading of unirradiated fuel into a
manufactured reactor, are intended to
cover a factory-fabrication model that
has been suggested for some microreactor designs. However, as written, the
proposed provisions are not limited to
any size or type of reactor.
The NRC is seeking comment on
whether the proposed regulations are
sufficient to govern various scenarios for
the possible manufacturing and
deployment of manufactured reactors.
If a comment indicates that the
proposed regulations are not sufficient,
please describe the reasons why,
including, if applicable, any plausible
scenario for which the commenter
believes the proposed regulations are
not sufficient.
2. The proposed regulations in
subpart H allow holders of or applicants
for a COL to reference an ML but do not
include such a provision for the holder
of or applicant for a CP or OL. This
proposed change from the current
relationship between subparts in part 52
and the part 50 licensing process was
made to simplify the provisions in the
proposed part 53 for licensing and
deploying manufactured reactors.
The NRC seeks comment on whether
part 53 should include provisions for an
applicant for or a holder of a CP or an
OL to reference an ML and, if so, how
this should be done.
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3. Proposed § 53.1295 states that the
holder of an ML could not begin
manufacture of a manufactured reactor
less than 6 months before the expiration
of the license. This limitation is similar
to the current restriction in § 52.177,
which states that the manufacture of a
reactor cannot begin less than 3 years
before the expiration of the license. The
restriction was revised from 3 years in
part 52 to 6 months in the proposed part
53 in recognition of the likely use of
MLs for a factory-fabrication model for
micro-reactors.
The NRC seeks comment on whether
it is necessary or appropriate to revise
the 3-year restriction in part 52 on when
manufacturing activities could begin in
relation to license expiration and, if so,
what that restriction should be.
4. Proposed § 53.1288 provides the
finality provisions for MLs and
includes, as does existing § 52.171,
limitations on the NRC’s imposition of
new requirements on either the design
or the requirements for the manufacture
of a manufactured reactor. No MLs have
been issued under part 52 and there is
no practical experience with the
proposed finality sections. While the
implications of the finality provisions
related to the design of a manufactured
reactor can reasonably be inferred from
experience with DCs and COLs, there is
no experience or available guidance
regarding finality for ‘‘requirements for
the manufacture of the manufactured
reactor.’’
The NRC is seeking comment on the
proposed finality provisions for MLs
and specifically if and how finality for
manufacturing processes might be
requested and used.
5. The NRC is seeking comment on
the proposed regulations for the loading
of fresh (unirradiated) fuel into a
manufactured reactor for subsequent
transport to a site for which the
Commission has issued a COL that
authorizes construction and operation of
a commercial nuclear plant using the
manufactured reactor. The proposed
regulation includes provisions for
loading of fuel into manufactured
reactors at a manufacturing facility prior
to transporting the fueled reactor to its
deployment site, as suggested by some
stakeholders. The NRC has historically
viewed reactor operation as including
fuel load, and existing NRC regulations
reflect this view. While the Act
authorizes the NRC to issue licenses to
manufacture production or utilization
facilities, it does not contain specific
provisions on fueling or operating
facilities licensed under an ML, and
existing ML regulations under part 52
do not include provisions for fuel load.
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The proposed rule addresses this
matter by allowing an applicant to
combine an ML with a part 70 license,
which would authorize possession of a
manufactured reactor in which the
licensee has loaded unirradiated fuel
provided at least two independent
criticality prevention mechanisms are in
place, each of which is sufficient to
prevent criticality assuming optimum
neutron moderation and neutron
reflection conditions. This requirement
would limit the possibility of creating
fission products and allow the control of
SNM, so that the loading of the fuel into
a manufactured reactor could be
governed primarily via a part 70 license
and associated regulations (including
those in subpart H of part 70).
A specific topic on which the NRC is
seeking comment is on the potential
benefits of and issues with including the
requirements of subpart H of part 70
within the proposed regulations for
loading fuel into manufactured reactors
at the manufacturing facility. For
example, should the NRC include a
threshold for including the
requirements of subpart H of part 70
and, if so, what factors and decision
criteria should be considered in such a
threshold? If a comment indicates that
the proposed regulations are not
sufficient, please describe the reasons
why, including the plausible scenarios
for which the proposed regulations
would not work or could be made to
work better.
6. Section 170, ‘‘Indemnification and
Limitation of Liability,’’ of the Act states
that each license under section 103 shall
have as a condition of the license a
requirement that the licensee have and
maintain financial protection of such
type and in such amounts as the NRC
shall require.
The NRC is seeking comment on
whether the proposed regulations
should include amounts of required
financial protections for MLs for fueled
manufactured reactors, and, if so, what
would be appropriate amounts of
required financial protection.
7. Some stakeholders have suggested
that a fueled manufactured reactor with
appropriate protections against
criticality should not be categorized as
a utilization facility under NRC
regulations or Section 11cc. of the Act.
The NRC is seeking comment on
possible approaches where the NRC
could find that a fueled manufactured
reactor would not be a utilization
facility, the basis for such a finding, and
the potential benefits of and potential
issues with such a finding.
8. Proposed requirement
§ 53.620(d)(2)(i) would require a
security program, including a physical
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security plan, for any ML authorizing
possession of a manufactured reactor
into which fuel has been loaded at the
manufacturing facility. Currently,
requirements in § 73.67(c)(1) only
require that a physical security plan be
submitted for those licensees who
possess, use, transport, or deliver to a
carrier for transport SNM of moderate
strategic significance, or 10 kg or more
of SNM of low strategic significance.
The NRC is seeking comment on
whether the proposed requirement: (1)
should be specific to the facility type
(i.e., manufacturing facility) or be
specific to the category of material being
used at the facility; (2) should apply to
all manufacturing plants, including
those at which licensees may only
possess SNM of low strategic
significance (i.e., category III), or only
those facilities for which an applicant
must submit a physical security plan
per § 73.67(c)(1); or (3) should include
more specific requirements on the
supplemental security measures that
may be needed for licensees possessing
SNM of moderate strategic significance
(i.e., category II)?
9. Proposed requirement
§ 53.620(d)(2)(i) would require a
cybersecurity program. The proposed
general cybersecurity performance
requirements would be to provide
reasonable assurance that a cyberattack
could not adversely impact the
functions performed by digital assets
used by the licensee for implementing
the physical security, radiation
monitoring, and criticality
requirements.
The NRC is seeking comment on the
following: (1) to what extent
stakeholders envision physical security
controls, radiation monitoring, and
criticality controls at a manufacturing
facility being digital; (2) to what extent
should the ML holder be required to
protect digital computer and
communications systems that impact
safety and security functions from a
cyberattack at a manufacturing facility
authorized to load fuel; and (3) whether
the rule provides sufficient clarity on
the cybersecurity measures needed for
license issuance or if additional detail
should be included either in the rule or
in guidance?
10. Proposed requirement
§ 53.620(d)(2)(i)(B) would require that
the physical security program be
designed to prevent unintended and
uncontrolled criticality events. This
would include criticality events that are
initiated maliciously.
The NRC is seeking comment on
whether the ML holder should be
required to design its security program
to protect against radiological sabotage
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(i.e., an unintended criticality event
leading to unacceptable radiological
consequences), in addition to theft and
diversion. For example, should the NRC
establish security requirements to
prevent an adversary, including an
insider, from tampering with the reactor
at a manufacturing facility or during
transport in such a way as to cause an
inadvertent criticality event? If so,
should the NRC consider factors such as
the category of fuel and the number of
reactors at a factory that can
simultaneously be loaded with fuel in
establishing the security requirements?
11. Proposed requirement
§ 53.620(d)(2)(i) would require an ML
holder to meet the performance
objectives in § 73.67. Requirements
§ 73.67(e) and § 73.67(g) include
provisions for security of category II and
category III quantities of SNM,
respectively, during transportation.
The NRC is seeking comment on the
extent to which the ML should require
ASMs (i.e., security measures above
those required by § 73.67(e) and
§ 73.67(g)) for transportation of a fueled
reactor to its place of operation. What
should those measures be?
12. Proposed requirement
§ 53.620(d)(2)(i) would require an ML
holder to meet the performance
objectives of § 73.67. For licensees
utilizing a category II quantity of SNM,
the requirement in § 73.67(d)(4) would
have the ML holder conduct a screening
to confirm the identity of an individual
prior to granting unescorted access to
the controlled access area where the
material is used or stored. The purpose
of this requirement is to both confirm
the identity of the individual and
support a determination that the
individual is trustworthy and reliable.
The NRC is seeking comment on
whether the ML requirements should
include ASMs (i.e., measures beyond
those required by § 73.67(d)(4)) in order
to provide reasonable assurance of
identity confirmation and
trustworthiness and reliability.
13. The NRC is seeking comment on
whether provisions regulating the
testing of fueled manufactured reactors
in the manufacturing facility should be
included in part 53 and, if so, what
would be practical for the holder of an
ML while also providing adequate
protection of public health and safety.
One possibility could be COLs that
would be issued to the holders of an ML
to cover low power (e.g., <5% rated
thermal power) nuclear physics testing
of fueled manufactured reactors within
the manufacturing facility prior to the
manufactured reactors being transported
to and incorporated into a commercial
nuclear plant for the purpose of energy
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production. The NRC recognizes
configuration changes are needed to
perform nuclear physics testing and is
seeking comment on what requirements
should apply to the manufactured
reactors and the manufacturing facility
during such testing (e.g., limiting power
levels). If a comment indicates that the
regulations should address limited
operations at manufacturing facilities,
please describe the likely scenarios that
would need to be addressed and suggest
what would be appropriate
requirements for such scenarios.
While an ML holder could
accomplish nuclear physics testing by
applying for a COL under the proposed
subpart H of part 53, stakeholders have
indicated that many of the requirements
would likely be unnecessary, given the
reduced risk profile posed by such
activities. Therefore, the NRC is seeking
comment on what requirements in
subpart H of part 53 should apply to
applicants for a COL who would
perform testing of fueled manufactured
reactors at the manufacturing plant.
Examples of proposed requirements that
might be relaxed or modified for
applications for low power testing at
manufacturing plants include those
related to selection of LBEs to reflect
limited inventory of radionuclides and
decay heat, aircraft impact assessments,
and earthquake engineering.
Additionally, the NRC is seeking
comment on whether several other
requirements in part 53 could be
modified for applications for a low
power testing COL at a manufacturing
facility. For example, the NRC is seeking
comment on how portions of the ML
facility used to support testing should
fall within the requirements for
construction activities under § 53.610;
whether §§ 53.710 and 53.715 (SSC
configuration control) must be
implemented to ensure portions of the
ML facility relied on to limit potential
radiological consequences from LBEs
are available to perform their safety
functions; and whether the
requirements of § 53.730 could be
modified to reflect the conditions of low
power physics testing. If a comment
indicates that some design and analysis
requirements and related application
requirements in subpart H of the
proposed part 53 are not needed for the
testing of fueled manufactured reactors,
please provide a rationale supporting
your comment and, if applicable, what
alternate requirements would be
appropriate.
Moreover, the licensing mechanism
for the facility could present unique
challenges. One option could be to issue
a low power testing COL for each fueled
manufactured reactor to be tested. This
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would comport with the agency’s
practice of issuing one license per
reactor but could prove prohibitive from
a cost standpoint and may provide very
little safety benefit if all manufactured
reactors are the same. Alternatively, one
low power testing COL could be issued
for the portions of the ML facility used
to test the fueled manufactured reactors
and allow multiple fueled manufactured
reactors to be completed and tested over
the course of the ML. Under this
approach, any ITAAC related to testing
of the fueled manufactured reactors
would need to be closed after they were
manufactured but prior to testing, and
the NRC would issue a notice of
intended operation and provide the
public an opportunity to request a
hearing on whether each fueled
manufactured reactor as constructed
complies, or on completion will
comply, with the acceptance criteria of
the license. The NRC is seeking
comment on the potential benefits and
issues with having a COL for each
fueled manufactured reactor to be tested
versus having a COL cover the testing of
multiple fueled manufactured reactors.
If a comment indicates a preference for
a particular approach, please provide a
rationale supporting the comment and
describe the specific scenarios that the
regulations need to address.
Part 53, Subpart F—Staffing and
Generally Licensed Reactor Operators
Under the Act Sections 106 and 107,
the NRC is proposing to group
commercial reactors into classes upon
the basis of the similarity of operating
and technical characteristics of the
facilities, and then to prescribe uniform
conditions for licensing individuals as
operators of any of the various classes;
determine the qualifications of such
individuals; and, for certain classes of
commercial reactors, issue general
licenses (i.e., licenses for which no
application is needed) to such
individuals allowing the individuals to
operate the commercial reactor.
1. Categories of Individuals Who May
Manipulate Facility Controls: The NRC
is proposing requirements that would
allow the manipulation of the controls
of certain facilities by GLROs in lieu of
specifically licensed reactor operators
and senior reactor operators. Reactor
operators and senior reactor operators
are the only categories of individuals
currently allowed to be licensed to
manipulate the controls of utilization
facilities under part 55.
The NRC is interested in public
perspectives on this proposed addition
of the GLRO category, particularly in
light of new reactor technologies and
concepts of operations.
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2. Criteria for GLRO Staffing: The
NRC is proposing criteria under which
facilities would be staffed by GLROs in
lieu of specifically licensed reactor
operators and senior reactor operators.
These criteria establish a new class of
self-reliant-mitigation facilities, as
defined in part 53, for which distinct
GLRO licensing and staffing
requirements would apply.
The NRC is soliciting public feedback
regarding whether these proposed
criteria are appropriate and what, if any,
alternative criteria should be
considered. Please provide your
considerations and rationale for your
answer.
3. Medical Requirements for GLROs:
Based on the proposed criteria that a
self-reliant-mitigation facility, as
defined in part 53, must meet, the NRC
is proposing not to subject GLROs to
requirements for medical fitness and
medical examination. This is in contrast
with the proposed requirements
associated with specifically licensed
reactor operators and senior reactor
operators, as well as the existing
requirements for reactor operators and
senior reactor operators under part 55.
The NRC is soliciting public feedback
regarding whether GLROs should be
subject to medical fitness and/or
medical examination requirements like
reactor operators and senior reactor
operators. Please provide your
considerations and rationale for your
answer.
4. Onshift Engineering Expertise: The
NRC is proposing to require that
engineering expertise be accounted for
within facility staffing plans. This
proposed requirement would be in lieu
of the traditional position of the Shift
Technical Advisor. The NRC is further
proposing that individuals providing
such engineering expertise would need,
among other things, to possess either a
qualifying 4-year degree or licensure as
a Professional Engineer.
The NRC is interested in feedback
from the public regarding the
appropriateness of this requirement,
including any alternatives that should
be considered. Please provide your
considerations and rationale for your
answer.
5. Use of Simulation Facilities as HFE
Testbeds: The NRC is proposing to
establish regulations pertaining to the
use of simulation facilities within the
context of the licensing programs both
for specifically licensed reactor
operators and senior reactor operators as
well as for GLROs. However, these
regulations, as currently proposed, do
not address the use of simulation
facilities within the context of serving as
testbeds for HFE-related analyses and
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assessments. Rather, the NRC currently
envisions that the use of simulation
facilities as HFE testbeds is more
appropriately addressed via guidance
documents.
The NRC is soliciting public feedback
regarding whether simulation facility
requirements should also address the
use of simulation facilities as HFE
testbeds. Please provide your
considerations and rationale for your
answer.
Part 53, Subpart F—Emergency
Preparedness and Security Programs
1. The proposed framework for part
53 would incorporate the changes to
NRC regulations from the final
rulemaking on ‘‘Emergency
Preparedness for Small Modular
Reactors and Other New Technologies’’
(the EP for SMR/ONT rule) by including
references to § 50.160, ‘‘Emergency
preparedness for small modular
reactors, non-light-water reactors, and
non-power production or utilization
facilities,’’ and by making conforming
changes within § 50.160. The proposed
framework for part 53 would also
introduce a graded approach to physical
protection requirements that includes
the criterion in § 53.860(a)(2)(i) to
establish a class of licensees that would
not be required to protect against the
design-basis threat (DBT) of radiological
sabotage. The NRC is soliciting public
comment relating to these topics, which
could include ways that graded
approaches for both emergency
preparedness and security programs
might be assessed and considered
during the licensing process.
The NRC is seeking comment on the
sufficiency and clarity of requirements
in proposed part 53 related to the
assessments needed to support graded
emergency planning and security. If a
comment indicates that there is an issue
with the sufficiency or clarity of the
proposed regulations, please describe
the reasons why, including, if
applicable, any scenario for which the
proposed regulations are not sufficient
and possible ways to clarify the
requirements. The NRC is specifically
seeking comment on possible challenges
arising from the interactions between
the proposed regulations and related
assessments for grading the
requirements for emergency planning
and security.
2. The NRC is preparing various
guidance documents to support this
rulemaking and other ongoing or
recently completed rulemakings related
to emergency preparedness and
security. DG–5076, ‘‘Guidance for
Technology-Inclusive Requirements for
Physical Protection of Licensed
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Activities at Commercial Nuclear
Plants,’’ has been issued along with this
proposed rulemaking and public
comments are requested via this notice
on that draft guidance. The NRC is also
planning to issue a draft revision of RG
1.242, ‘‘Performance-Based Emergency
Preparedness for Small Modular
Reactors, Non-Light-Water Reactors, and
Non-Power Production or Utilization
Facilities,’’ for public comment. The
planned revision to RG 1.242 would add
guidance for part 53 applicants and
licensees.
In the staff requirements
memorandum to SECY–23–0021, the
Commission directed the NRC staff to
address the consideration of securityrelated events for an advanced reactor
that addresses security through design
and engineered safety features when it
harmonizes this rulemaking with the EP
for SMR/ONT rule. In the EP for SMR/
ONT rule, the NRC established an
alternative performance-based and riskinformed approach for emergency
planning, including determining the
need for and size of an emergency
planning zone (EPZ) to support
predetermined, prompt protective
actions. The NRC has incorporated the
relevant rule language from the EP for
SMR/ONT rule into this proposed rule
and is seeking stakeholder feedback as
to whether additional rule language
changes or additional guidance would
be beneficial.
In light of the Commission direction
and the above considerations, the NRC
is assessing how best to address the
treatment of security-related events in
emergency planning, including in the
determination of EPZ size, for reactors
licensed under part 53. Part 53 is
introducing an alternative approach to
meeting security regulations that should
be taken into consideration under
§ 50.160. Stakeholders are encouraged to
take a holistic view of the various
activities and opportunities to provide
comments on this rulemaking and
related guidance supporting this
rulemaking (e.g., DG–5076 on physical
protection requirements, future
revisions to RG 1.242). In developing
comments, the NRC urges stakeholders
to consider various scenarios that might
arise when implementing graded
approaches for security and emergency
planning for various reactor designs.
Scenarios could include the following:
• the potential consequences from
security events up to and including the
DBT of radiological sabotage are
bounded by unlikely and very unlikely
event sequences such that security
events do not need separate analyses in
the EPZ size determination;
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• the potential consequences from
security events up to and including the
DBT are not bounded by unlikely and
very unlikely event sequences but could
otherwise support a reduced EPZ size
consistent with considerations
discussed in RG 1.242 and NUREG–
0396, ‘‘Planning Basis for the
Development of State and Local
Government Radiological Emergency
Response Plans in Support of Light
Water Nuclear Power Plants’’; or
• the potential consequences from
security events up to and including the
DBT are not bounded by unlikely and
very unlikely event sequences and
warrant consideration of increasing the
size of the EPZ.
The NRC is interested in comments
on the need for additional rule language
or guidance to address graded
approaches for emergency planning and
security programs under the scenarios
described above for part 53 applicants
and licensees. Please address within the
comments any technical, policy, or legal
issues that are associated with your
suggestions.
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Part 53, Subpart F—Integrity
Assessment Program Requirements
Decades of operating experience with
LWRs suggests that phenomena such as
environmentally assisted fatigue and
chemical interactions could impact
certain SSCs during the life of a
commercial nuclear plant. Under the
existing regulatory framework,
historically, some of these phenomena
were not addressed during early
licensing reviews but were identified
and addressed later when significant
safety issues arose (e.g., see numerous
generic letters, bulletins, orders, and
development and implementation of
vessel integrity and materials reliability
programs) or a licensee voluntarily
pursued renewal of an OL under part
54. The NRC is proposing to include a
new set of programmatic requirements
for an Integrity Assessment Program that
would ensure these phenomena are
addressed early in the life of a
commercial nuclear plant licensed
under part 53. The requirements would
be provided in § 53.870.
The NRC is seeking comment on
whether the proposed requirements
under the Integrity Assessment Program
appropriately complement design
requirements to address concerns
regarding aging, cyclic or transient load
limits, and degradation mechanisms
related to chemical interactions,
operating temperatures, effects of
irradiation, and other environmental
factors. In addition, the NRC is
interested in views on whether, and if
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so how, degradation mechanisms are or
could be addressed in other programs.
Part 53, Subpart G—Decommissioning
1. On March 3, 2022, the NRC
published the proposed rule entitled
‘‘Regulatory Improvements for
Production and Utilization Facilities
Transitioning to Decommissioning’’ (87
FR 12254). This rulemaking would
amend the NRC’s current regulations to
provide an appropriate regulatory
framework for nuclear power reactors
transitioning from operations to
decommissioning. The rulemaking
would address lessons learned from
licensees that have completed or are
currently in the decommissioning
process. The NRC staff sent a draft final
rule to the Commission for its
consideration on January 31, 2024, in
SECY–24–0011, ‘‘Final Rule: Regulatory
Improvements for Production and
Utilization Facilities Transitioning to
Decommissioning (3150–AJ59; NRC–
2015–0070).’’
What aspects of this draft final rule,
if any, should be incorporated in a part
53 final rule and why?
2. Proposed § 53.1060(b) in subpart G
would require that, ‘‘No later than 30
days after the Commission publishes
notice in the Federal Register under
§ 53.1452(a), the licensee must submit a
report containing a certification that
financial assurance for
decommissioning is being provided in
an amount specified in the licensee’s
most recent updated certification,
including a copy of the financial
instrument obtained to satisfy
§ 53.1040.’’ This is similar to the current
requirement in § 50.75(e)(3) for part 52
COL holders. The NRC is seeking
comment on whether commercial
nuclear plant COL holders under part 53
should have the same requirement as
COL holders under part 52 to
demonstrate that they have financial
assurance in place no later than 30 days
after the Commission issues the notice
of intended operation under § 53.1452.
Please provide your considerations and
rationale for your answer.
Part 53, Subpart H—Licenses To
Construct and Operate Commercial
Nuclear Plants of Identical Design at
Multiple Sites
In addition to including provisions in
part 53, subpart H, for referencing ESPs,
standard design approvals, and design
certifications in applications for
commercial nuclear plants, the
proposed § 53.1470 provides optional
requirements related to the submittal
and NRC review of CP, OL, and COL
applications to construct and operate
commercial nuclear plants of identical
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design at multiple sites, similar to
requirements found in appendix N in
both 10 CFR parts 50 and 52. This
section would set out the particular
requirements and provisions applicable
to situations in which applications for
CPs and subsequent OLs, or COLs,
under this part, are filed by one or more
applicants for licenses to construct and
operate nuclear power reactors of
identical design (‘‘common design’’) to
be located at multiple sites. Hearings for
applications filed under appendix N in
both parts 50 and 52 are governed by
subpart D of part 2, as would be the case
for future part 53 applications under
proposed § 53.1470.
Under the proposed requirements in
this section, each application is to be
treated as a separate application, with
the exception of the common design,
and so would require separate
applications, separate determinations of
sufficiency for docketing, separate
notices of docketing, and so forth.
Proposed § 53.1470 would also require
that each application list all the
applications that are to be treated
together to ensure that the NRC is
clearly informed of the intentions of all
applicants. Ordinarily, the NRC would
publish in the Federal Register a
separate notification of docketing for
each application, so that delays in the
docketing of one application would not
delay the docketing and subsequent
technical review of other applications.
However, if circumstances allow (e.g.,
sufficiency review for multiple
applications are completed
simultaneously), the NRC could publish
a single notice of docketing for multiple
applications.
With regard to how the NRC would
fulfill its obligations under the National
Environmental Policy Act of 1969, as
amended, the NRC staff would prepare
a separate environmental document for
each application, but the NRC could
conduct joint scoping on environmental
issues related to the common design. If
the applications reference a standard
design certification or the use of a
manufactured reactor, then the
environmental document would need to
incorporate by reference the
environmental assessment (EA)
prepared for either the design
certification or the ML, as applicable. In
addition, § 53.1470 would require the
ACRS to report on each of the
applications, as would be required by
provisions in subpart H of part 53. Each
ACRS report would be limited to the
safety matters which are not relevant to
the common design. In addition, the
ACRS would need to issue a report on
the safety of the common design—
except for those matters relevant to the
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safety of a referenced design
certification or manufactured reactor.
Given this synopsis of how the
requirements in proposed § 53.1470
would be implemented as currently
written, the NRC is seeking comment on
whether there are opportunities to allow
added flexibility for applicants under
these provisions. This could include
consideration of whether applications
for which the ‘‘common design’’ is not
completely identical could be evaluated
under this provision and, if so, what the
process would be for determining the
appropriateness of a common review. In
addition, the NRC is interested in
feedback about the pros and cons of
requiring that applications under these
proposed provisions be submitted at the
same time versus allowing them to be
submitted on a staggered basis.
Part 53, Subparts H and I—Probabilistic
Risk Assessment Information
Proposed § 53.1239(a)(18) in subpart
H and the related references to this
proposed requirement for the holders of
OLs and COLs would require a
description of the PRA required by
§ 53.450(a), and its results to be
included in FSARs. However, guidance
documents may further clarify the
division of PRA-related information
needed to be in the FSAR, in other
possible licensing basis documents, and
controlled as plant records subject to
inspections and audits. For example, a
possible approach could be to include a
summary of the PRA results in the
FSAR and control that information
under § 53.1545 and create a separate
document related to the broader PRA
analyses and related processes as a
program document under § 53.1560. The
program document would provide more
detail than the summaries in the FSAR
but still be a much-condensed source of
information in comparison to the
documentation of the PRA. This
possible approach would reflect the role
of the PRA in the licensing process
under part 53 and in maintaining
margins to the safety and evaluation
criteria in subparts B and C but may
allow a more appropriate evaluation
process to address the particulars and
complexities of the PRA-related
documents.
The NRC is seeking comment on the
appropriate placement of PRA-related
information among various licensing
basis documents and plant records. In
addition to the placement of PRArelated information, the NRC is seeking
comment on the appropriate control of
that information and on the routine
submittal of updates to the NRC. Please
provide your considerations and
rationale for your answer.
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Part 53, Subparts H and I—Changes to
Manufacturing Licenses
Proposed § 53.1530 would not allow
the holder of an ML or the holder of a
COL using a manufactured reactor to
make changes to the design of the
manufactured reactor without
requesting a license amendment from
the NRC. The proposed requirements do
not include a specific mention of the
manufacturing processes for which the
NRC could possibly provide finality
under proposed § 53.1288.
The NRC is seeking comment on the
appropriate change control provisions
for MLs, including whether criteria
could be developed to determine when
a license amendment request would not
be required and whether those criteria
should address changes in
manufacturing processes as well as
changes in the design. Please provide
your considerations and rationale for
your recommendation.
Financial Qualifications
Utility new reactor applicants are
exempt under § 50.33(f) from financial
qualification reviews because they are
generically presumed to be financially
qualified for operations. In contrast,
merchant power plant new reactor
applicants are required under
§ 50.33(f)(2) to submit information that
demonstrates they possess or have
reasonable assurance of obtaining the
funds necessary to cover estimated
construction and operating costs for the
period of the license. A ‘‘merchant
power plant new reactor applicant’’ is a
non-rate-regulated entity (e.g., a
nonutility) that engages in the business
of production, manufacturing,
generating, buying, aggregating,
marketing, or brokering electricity for
sale at wholesale or for retail sale to the
public. Over the past decade, the agency
has heard some concerns about the
challenges that merchant power plant
applicants face in demonstrating
compliance with the current financial
qualification requirements.
Does this standard continue to pose
challenges for merchant power plant
applicants? If so, please provide a
detailed explanation of these challenges.
Should part 53 have the same
financial qualification requirements as
parts 50 and 52? Why or why not?
Are there categories of merchant new
reactor applicants for which a part 70
‘‘appears to be financially qualified’’
standard would be more appropriate? 15
If so, please explain what types of
applicants should be able to use the part
70 financial qualification standard and
15 Section
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what distinguishes these applicants
from ones that should not be able to use
this standard.
If a part 70 financial qualification
standard were to apply to a category of
merchant new reactor applicants,
should it also apply to pre-construction
license transfer applications for these
reactors? Why or why not?
Is there another standard the agency
should consider for financial
qualification of merchant new reactor
applicants? Commenters are encouraged
to provide specific suggestions and the
basis for those suggestions.
Part 73, Section 73.100—Physical
Security
The proposed § 73.100 would identify
the proposed performance-based
physical security requirements with
which future commercial power reactor
applicants or licensees’ physical
protection programs would need to
demonstrate compliance, without
prescribing the specific methods that
must be used to satisfy them. Applicants
and licensees would have increased
flexibility regarding the modern
technologies and methods that they
could use. Implementing guidance in
DG–5076 (proposed RG 5.97),
‘‘Guidance for Technology Inclusive
Requirements for Physical Protection of
Licensed Activities at Commercial
Nuclear Plants,’’ would be available to
assist applicants and licensees. For
example, DG–5076 provides detailed
guidance, including performance
standard recommendations, on the
probability of detection and alternative
sources of power for exterior intrusion
detection systems (subsection 4.1.1.1.A),
interior intrusion detection (subsection
4.1.1.1.B), intrusion assessment
(subsection 4.1.1.2.A), security
response/neutralization subsection
(4.1.1.4.A), security communication
(subsection 4.1.1.3.A), and security
delay (subsection 4.1.1.4.C).
Does the NRC’s proposed approach in
§ 73.100 provide a sufficient level of
detail to be readily understood and
easily applied to the licensing and
oversight of new and advanced power
reactors, or should the NRC consider
moving some objective and measurable
security performance standard
recommendations from the draft
implementing guidance in DG–5076
into proposed § 73.100? If so, which
objective and measurable security
performance standard recommendations
should be moved from DG–5076 to
§ 73.100? Please provide the basis for
your response.
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Part 73, Section 73.110—Cybersecurity
The proposed § 73.110 would require
licensees to demonstrate protection
against cyberattacks in a manner that is
commensurate with the potential
consequences from those attacks,
without prescribing the specific
methods that must be used to
demonstrate protection. Under proposed
§ 73.110(a), licensees would need to
ensure that digital computer and
communications systems are adequately
protected against a potential cyberattack
that would, for example, result in
adverse impacts to the physical security
digital assets used by the licensee to
prevent unauthorized removal of
material per § 53.860(a). Protecting
against such a potential cyberattack
would involve requiring cybersecurity
for SNM at a commercial nuclear reactor
licensed under part 53. Applicants and
licensees would have increased
flexibility regarding the modern
technologies and methods that they
could use for protecting against such a
potential cyberattack. Detailed
implementing guidance in DG–5075
(proposed RG 5.96), ‘‘Establishing
Cybersecurity Programs for Commercial
Nuclear Plants licensed under 10 CFR
part 53,’’ would be available to assist
applicants and licensees. For example,
DG–5075 provides guidance on the
implementation of security by design
features (e.g., facility design) for
negating the potential consequences
from such a potential cyberattack.
If a cyberattack were to compromise
the availability, integrity, or
confidentiality of data or systems
associated with security systems/
measures for the protection of SNM at
a commercial nuclear reactor licensed
under part 53, do the potential
consequences warrant requiring
cybersecurity for such material? Please
provide the basis for your response
including a detailed explanation of
challenges, if any, posed by requiring
cybersecurity for SNM at a commercial
nuclear reactor licensed under part 53.
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Recent Legislation
On July 9, 2024, the President signed
into law the Accelerating Deployment of
Versatile, Advanced Nuclear for Clean
Energy Act of 2024, also referred to as
the ADVANCE Act. Section 203,
‘‘Licensing Considerations Relating to
Use of Nuclear Energy for Nonelectric
Applications,’’ and Section 208,
‘‘Regulatory Requirements for MicroReactors,’’ of the ADVANCE Act
specifically mention the technologyinclusive regulatory framework to be
established under section 103(a)(4) of
NEIMA as a potential vehicle to be
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considered for the report to Congress
required under section 203 and a
potential vehicle to implement
strategies and guidance for the licensing
and regulation of micro-reactors
required under section 208. This
proposed rulemaking is, in part, how
the NRC is implementing section
103(a)(4) of NEIMA.
The NRC is seeking comment on how
part 53 could be revised to better enable
its potential use to implement the
ADVANCE Act. Specifically, Section
208 of the ADVANCE Act requires the
NRC to develop and implement ‘‘riskinformed and performance-based
strategies and guidance’’ in several areas
for the licensing and regulation of
micro-reactors, including with respect
to ‘‘licensing mobile deployment.’’ The
ADVANCE Act requires the NRC to
consider ‘‘the unique characteristics of
micro-reactors,’’ including physical size,
design simplicity, and source term;
opportunities to incorporate specific
improvements related to streamlining
the review process; and other policy and
licensing issues. With regard to
implementation, the ADVANCE Act
provides the NRC with three options.
The NRC may implement the developed
strategies and guidance, as appropriate,
via (1) the existing regulatory
framework, (2) the Part 53 rulemaking,
or (3) a pending or new rulemaking.
Given the language included in Section
208, the NRC is seeking comment on
how part 53 could be revised to better
address the ADVANCE Act’s
requirements related to strategies and
guidance for micro-reactors.
VII. Section-by-Section Analysis
The following paragraphs describe the
specific changes proposed by this
rulemaking.
§ 1.43 Office of Nuclear Reactor
Regulation
This proposed rule would revise
§ 1.43(a)(2) to extend the authority of
the Office of Nuclear Reactor Regulation
to regulate source, byproduct, and SNM
at facilities licensed under part 53.
§ 2.1 Scope
This proposed rule would revise
§ 2.1(e) to apply to standard design
approvals under part 53.
§ 2.4 Definitions
This proposed rule would revise § 2.4
to update the definition of ‘‘Contested
proceeding’’ to include NRC
enforcement actions against applicants
for a standard DC under part 53. It
would also update the definition of
‘‘Facility’’ to encompass utilization
facilities as defined in § 53.020 (there
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are no production facilities under part
53).
§ 2.100
Scope of Subpart
This proposed rule would revise
§ 2.100 to extend the scope of subpart A
to licenses and standard design
approvals issued under §§ 53.1200
through 53.1221.
§ 2.101
Filing of Application
This proposed rule would revise
§ 2.101 to be applicable to part 53
applicants in addition to part 50 and 52
applicants by adding references to part
53 in paragraphs (a)(3)(i), (a)(5), and
(a)(9).
§ 2.104
Notice of Hearing
This proposed rule would extend the
hearing notice requirement in § 2.104(a)
to applications concerning facilities
covered under part 53. Footnote 1 to
§ 2.104 would be revised in a
corresponding manner.
§ 2.105
Notice of Proposed Action
This proposed rule would revise
§ 2.105 to extend the requirement in
§ 2.104 to publish a notice of intended
operation or a notice of proposed action,
as applicable, to part 53 applicants in
addition to part 50 and 52 applicants by
adding corresponding references to part
53 in paragraphs (a), (a)(4), (a)(10),
(a)(12), (a)(13), and (b)(3).
§ 2.106
Notice of Issuance
This proposed rule would revise
§ 2.106 to extend the issuance notice
requirement to applications concerning
facilities covered under part 53 through
updated references in paragraphs (a)(2)
and (3), and (b)(2).
§ 2.109 Effect of Timely Renewal
Application
This proposed rule would revise
§ 2.109 to add references to part 53 in
paragraphs (b), (c), and (d) regarding the
timing of license renewal applications.
§ 2.110 Filing and Administrative
Action on Submittals for Standard
Design Approval or Early Review of Site
Suitability Issues
This proposed rule would revise
§ 2.110 to include references to part 53
in paragraphs (a)(1) and (b).
§ 2.202
Orders
This proposed rule would revise
§ 2.202(e) to add references to part 53
regarding the requirements to be
followed for orders involving the
modification of a license, COL, ESP,
standard DC rule, standard design
approval, or ML.
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§ 2.309 Hearing Requests, Petitions To
Intervene, Requirements for Standing,
and Contentions
This proposed rule would revise
§ 2.309 to include references to part 53
in paragraphs (a), (f)(1)(i), (f)(1)(vi) and
(vii), (g), (h)(2), (i)(2), and (j) regarding
a request for hearing under § 53.1452.
§ 2.310 Selection of Hearing
Procedures
This proposed rule would revise
§ 2.310 by revising paragraph (a), the
introductory text for paragraph (h), and
paragraphs (i) and (j) to incorporate
references to part 53 regarding hearing
procedures.
§ 2.329 Prehearing Conference
This proposed rule would revise
§ 2.329(a) to extend the timing
requirements for prehearing conferences
involving CPs and licenses under part
53.
§ 2.402 Separate Hearings on Separate
Issues; Consolidation of Proceedings
This proposed rule would revise
§ 2.402(a) to apply provisions regarding
separate hearings and the consolidation
of proceedings to part 53 applicants.
§ 2.403 Notice of Proposed Action on
Applications for Operating Licenses
Pursuant To Appendix N of 10 CFR Part
50
This proposed rule would revise
§ 2.403 to require the Commission to
publish a notification of proposed
action in the Federal Register after
applications under part 53 are docketed.
§ 2.404 Hearings on Applications for
Operating Licenses Pursuant to
Appendix N of 10 CFR Part 50
§ 2.339 Expedited Decision-Making
Procedure
This proposed rule would revise
§ 2.339(d) to include references to part
53 regarding expedited decision-making
procedures.
This proposed rule would revise
§ 2.404 to apply to applications for an
OL under part 53.
§ 2.340 Initial Decision in Certain
Contested Proceedings; Immediate
Effectiveness of Initial Decisions;
Issuance of Authorizations, Permits and
Licenses
This proposed rule would revise
§ 2.340 regarding initial decisions of a
presiding officer in certain contested
proceedings, the effective date of those
decisions, and the issuance of
authorizations, permits, and licenses, by
incorporating references to part 53 in
paragraphs (b), (c), (d), (f), (i), and (j).
This proposed rule would revise
§ 2.405 to be applicable to CPs, fullpower OLs, and COLs under part 53.
§ 2.341 Review of Decisions and
Actions of a Presiding Officer
This proposed rule would revise
§ 2.341(a)(1) to include an updated
reference to part 53 regarding the
allowance of a period of interim
operation.
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the hearing notice requirement to
applications concerning facilities
covered under part 53.
§ 2.400 Scope of Subpart
This proposed rule would revise
§ 2.400 to extend the scope of subpart D
of part 2 to include part 53 applicants
for licenses to construct or operate
nuclear power reactors of identical
design at multiple sites.
§ 2.401 Notice of Hearing on
Construction Permit or Combined
License Applications Pursuant to
Appendix N of 10 CFR Parts 50, 52, or
53
This proposed rule would revise the
section heading and § 2.401 to extend
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§ 2.405 Initial Decisions in
Consolidated Hearings
§ 2.406 Finality of Decisions on
Separate Issues
This proposed rule would revise
§ 2.406 to be applicable to proceedings
conducted pursuant to part 53.
§ 2.500
Scope of Subpart
This proposed rule would revise
§ 2.500 to extend the provisions of
subpart E of part 2 to include
applications for a license to
manufacture nuclear power reactors
under part 53.
§ 2.501 Notice of Hearing on
Application Under Subpart F of 10 CFR
Part 52 or 53 for a License To
Manufacture Nuclear Power Reactors
This proposed rule would revise the
section heading and § 2.501(a) by
extending its provisions to applications
for a license to manufacture nuclear
power reactors under part 53.
§ 2.643 Acceptance and Docketing of
Application for Limited Work
Authorization
This proposed rule would revise
§ 2.643(b) regarding the acceptance and
docketing of an application for a CP for
a utilization facility of the type specified
in part 53.
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§ 2.645 Notice of Hearing
This proposed rule would revise
§ 2.645(a) to incorporate a reference to
part 53.
§ 2.649 Partial Decisions on Limited
Work Authorization
This proposed rule would revise
§ 2.649 to extend its provisions to LWAs
issued under part 53.
§ 2.800 Scope and Applicability
This proposed rule would revise
§ 2.800 by revising paragraphs (c) and
(d) to incorporate references to part 53
regarding the scope and applicability of
the rulemaking procedures contained in
this subpart.
§ 2.801 Initiation of Rulemaking
This proposed rule would revise
§ 2.801 to include a reference to part 53.
§ 2.813 Written Communications
This proposed rule would revise
§ 2.813(a) to apply general requirements
for correspondence with the
Commission to communications
concerning part 53, in addition to parts
50, 52, and 100.
§ 2.1103 Scope of Subpart K
This proposed rule would revise the
first sentence of § 2.1103 to extend the
provisions of subpart K of part 2 to
licenses under part 53 to expand the
spent fuel capacity at the site of a
civilian nuclear power plant.
§ 2.1202 Authority and Role of NRC
Staff
This proposed rule would amend
§ 2.1202 by revising paragraphs (a)(1)
through (3), and (a)(6) to include
references to part 53.
§ 2.1301 Public Notice of Receipt of a
License Transfer Application
This proposed rule would revise
§ 2.1301(b) to include a corresponding
reference to license transfers under part
53 in addition to parts 50 and 52.
§ 2.1403 Authority and Role of the
NRC Staff
This proposed role would update
§ 2.1403 to specify that ‘‘significant
hazards considerations’’ has the same
meaning as defined in part 53.
§ 2.1500 Purpose and Scope
This proposed rule would revise
§ 2.1500 to extend the scope of subpart
O of part 2 to DC rulemaking hearings
under part 53.
§ 2.1502 Commission Decision To
Hold Legislative Hearing
This proposed rule would revise
§ 2.1502, paragraphs (a) and (b)(1) to
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incorporate references to part 53
regarding the Commission’s decision to
hold a DC rulemaking.
§ 10.1 Purpose
This proposed rule would revise
§ 10.1(a)(3) to include a reference to part
53.
§ 10.2 Scope
This proposed rule would revise
§ 10.2(b) to extend the scope of subpart
A to applicants and holders of licenses,
certificates, and standard design
approvals under part 53 in addition to
part 52.
§ 11.7 Definitions
This proposed rule would revise
§ 11.7 such that terms defined in part 53
have the same meaning when used in
part 11.
§ 19.2 Scope
This proposed rule would revise
§ 19.2(a) to include references to part 53.
§ 19.3 Definitions
This proposed rule would revise the
definitions of ‘‘License’’ and ‘‘Regulated
entities’’ in § 19.3 to incorporate
references to part 53.
§ 19.11 Posting of Notices to Workers
This proposed rule would amend
§ 19.11 by revising paragraphs (a), (b),
and (e)(1) to apply to applicants and
holders of licenses, permits, standard
design approvals, and standard DCs
under part 53 in addition to part 52.
§ 19.14 Presence of Representatives of
Licensees and Regulated Entities, and
Workers During Inspections
This proposed rule would revise
§ 19.14(a) to apply to applicants and
holders of a license, standard design
approval, ESP, or standard DC under
part 53 in addition to part 52.
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§ 19.20 Employee Protection
This proposed rule would revise
§ 19.20 to include a reference to
protected activities under part 53.
§ 20.1002 Scope
This proposed rule would revise the
first sentence of 10 CFR part 20,
‘‘Standards for Protection Against
Radiation,’’ § 20.1002 to extend the
scope of part 20 to apply to persons
licensed by the Commission to receive,
use, transfer, or dispose of byproduct,
source, or SNM or to operate a
production or utilization facility under
part 53.
§ 20.1003 Definitions
This proposed rule would revise
§ 20.1003 to update the definition of
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‘‘License’’ to include those issued under
part 53.
§ 20.2201 Reports of Theft or Loss of
Licensed Material
§ 20.1101
Programs
This proposed rule would revise
§ 20.2201 to include references to part
53 in paragraphs (a)(2)(i), (b)(2)(i) and
(c) regarding requirements for reports of
theft or loss of licensed material.
Radiation Protection
This proposed rule would revise
§ 20.1101(d) to exclude licensees subject
to § 53.260 from its requirements.
§ 20.1401
Scope
General Provisions and
This proposed rule would revise
§ 20.1401, paragraphs (a) and (c) to
extend the scope of subpart E of part 20
to apply to the decommissioning of
facilities licensed under part 53 and the
release of part of a facility or site for
unrestricted use in accordance with
§ 53.1080.
§ 20.1403 Criteria for License
Termination Under Restricted
Conditions
This proposed rule would revise
§ 20.1403(d) to include
decommissioning plans under part 53.
§ 20.1404 Alternate Criteria for License
Termination
This proposed rule would revise
§ 20.1404(a)(4) to include a reference to
part 53 regarding alternate criteria for
license termination.
§ 20.1406 Minimization of
Contamination
This proposed rule would revise
§ 20.1406(a) to include references to
applicants for licenses other than ESPs
or MLs under part 53. It would also
revise § 20.1406(b) to include references
to standard DCs and standard design
approvals under part 53 in addition to
part 52.
§ 20.1501
General
This proposed rule would revise
§ 20.1501(b) regarding the requirement
for retention of records from surveys
describing the location and amount of
subsurface residual radioactivity at a
site to include a reference to the
retention requirements under part 53.
§ 20.1905 Exemptions to Labeling
Requirements
This proposed rule would revise
§ 20.1905(g) to apply to facilities
licensed under part 53 in addition to
parts 50 and 52 regarding exemptions to
labeling requirements.
§ 20.2004 Treatment or Disposal by
Incineration
This proposed rule would revise
§ 20.2004(b)(1) to include references to
part 53 regarding the treatment or
disposal of waste oil by incineration.
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§ 20.2202
Notification of Incidents
This proposed rule would revise
§ 20.2202(d)(1) to add references to part
53 regarding reports to the NRC
Operations Center.
§ 20.2203 Reports of Exposures,
Radiation Levels, and Concentrations of
Radioactive Material Exceeding the
Constraints or Limits
This proposed rule would revise
§ 20.2203(c) to refer to procedures under
part 53 for reporting occurrences of
exposures, radiation levels, and
concentrations of radioactive material
exceeding the constraints or limits.
§ 20.2206 Reports of Individual
Monitoring
This proposed rule would revise
§ 20.2206(a)(1) to include a reference to
part 53.
§ 21.2
Scope
This proposed rule would revise
§ 21.2, paragraphs (a), (b), and (c) to
include references to part 53 regarding
the scope and applicability of part 21
requirements.
§ 21.3
Definitions
This proposed rule, in § 21.3 would
revise the definitions of ‘‘Basic
component,’’ ‘‘Commercial grade item,’’
‘‘Critical characteristics,’’ ‘‘Dedicating
entity,’’ ‘‘Dedication,’’ ‘‘Defect,’’ and
‘‘Substantial safety hazard’’ with
references to part 53.
§ 21.21 Notification of Failure To
Comply or Existence of a Defect and Its
Evaluation
This proposed rule would revise
§ 21.21, by incorporating references to
part 53, to update the requirements for
notifying the Commission of a failure to
comply or defect in paragraphs (a)(3)
and (d)(1).
§ 21.51 Maintenance and Inspection of
Records
This proposed rule would revise
§ 21.51(a)(4) and (5) to apply to
applicants for standard DC and
applicants or holders of a standard
design approval under part 53, in
addition to part 52, regarding the
retention of records.
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§ 21.61 Failure To Notify
This proposed rule would revise
§ 21.61(b) to include references to part
53 licensees and applicants regarding
failure to provide the notice required in
§ 21.21.
§ 25.5 Definitions
This proposed rule would update the
definition of ‘‘License’’ to include those
issued under part 53.
§ 25.17 Approval for Processing
Applicants for Access Authorization
This proposed rule would revise
§ 25.17(a) to add a reference to part 53
regarding AAs for individuals who need
access to classified information in
connection with activities under part
53.
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§ 25.35 Classified Visits
This proposed rule would update
§ 25.35(a) to apply the requirements for
classified visits to licensees, certificate
holders, and applicants under part 53 in
addition to part 52.
26.605, 26.606, 26.607, 26.608, 26.609,
26.611, 26.613, 26.617, and 26.619.
§ 26.205 to incorporate references to
§§ 26.606 and 26.202(a) and (b).
§ 26.21
§ 26.207
Fitness-for-Duty Program
This proposed rule would revise
§ 26.21 to include a reference to
§ 26.3(f).
§ 26.51
Applicability
This proposed rule would revise
§ 26.51 to extend the requirements of
subpart C of part 26 to licensees and
other entities identified in § 26.3(f) that
do not implement the requirements of
subpart M of part 26, as well as
licensees and other entities that
implement the requirements of § 26.605.
§ 26.53
General Provisions
This proposed rule would revise
§ 26.53 paragraphs (e), (g), (h), and (i) to
include references to § 26.3(f).
§ 26.63
Suitable Inquiry
This proposed rule would revise
§ 26.63(d) with a reference to § 26.3(f).
§ 26.73
Applicability
§ 26.3 Scope
This proposed rule would amend
§ 26.3 by revising paragraph (d) and
adding new paragraph (f) which would
establish the phase of construction or
operation by which applicants and
licensees under part 53 would be
required to comply with subpart M of
part 26, or all of the requirements of part
26 except subparts K and M.
This proposed rule would revise
§ 26.73 to extend the requirements of
subpart D of part 26 to licensees and
other entities identified in § 26.3(f) that
do not implement the requirements of
subpart M of part 26, as well as
licensees and other entities that
implement the requirements of
§ 26.605(b).
§ 26.4 FFD Program Applicability to
Categories of Individuals
This proposed rule would revise
paragraphs (a), (b), (c), (e), (f), (g), and
(h) of § 26.4 to include references to part
53 and provisions for implementing an
FFD program under subpart M.
This proposed rule would revise
§ 26.81 to extend the requirements of
subpart E of part 26 to licensees and
other entities identified in § 26.3(f) that
do not implement the requirements of
subpart M of part 26, as well as
licensees and other entities that
implement the requirements of § 26.605.
§ 26.5 Definitions
This proposed rule would amend
§ 26.5 by adding definitions for
‘‘Biological marker,’’ ‘‘Change,’’ ‘‘Illicit
substance,’’ ‘‘Reduction in FFD program
effectiveness,’’ and ‘‘Special Nuclear
Material.’’ It would also revise
definitions of ‘‘Constructing or
construction activities,’’ ‘‘Contractor/
vendor (C/V),’’ ‘‘Other entity,’’
‘‘Questionable validity,’’ ‘‘Reviewing
official,’’ ‘‘Safety-related structures,
systems, and components (SSCs),’’
‘‘Security-related SSCs,’’ and ‘‘Unit
outage’’ within this section.
§ 26.8 Information Collection
Requirements: OMB Approval
This proposed rule would revise
§ 26.8(b) with the new information
collection requirements contained in
proposed §§ 26.202, 26.603, 26.604,
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§ 26.81
§ 26.201
Purpose and Applicability
Applicability
This proposed rule would revise
§ 26.201 to include references to the
proposed provisions in §§ 26.3(f) and
26.202, as well as revise the
applicability of requirements in subpart
I of part 26.
§ 26.202 General Provisions for
Facilities Licensed Under Part 53
This proposed rule would add new
§ 26.202, which would require
applicable licensees under part 53 to
incorporate a policy for fatigue
management into their FFD program in
accordance with the provisions of this
section.
§ 26.205
Work Hours
This proposed rule would revise
paragraphs (d)(7)(iii) and (d)(8) of
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Waivers and Exceptions
This proposed rule would revise
§ 26.207(a)(1)(ii) to include references to
§§ 26.608 and 26.202(c) and to include
provisions for implementing certain
face-to-face supervisor assessments
using electronic communications.
§ 26.211
Fatigue Assessments
This proposed rule would revise
§ 26.211, paragraphs (a)(1), (a)(3), and
(b) to incorporate references to
§§ 26.202(c), 26.607(b), 26.608, and
26.619 and to include provisions for
implementing certain face-to-face
assessments using electronic
communications.
Subpart M—Fitness for Duty Programs
for Facilities Licensed Under Part 53
This proposed rule would add new
Subpart M of part 26 containing
§§ 26.601, 26.603, 26.604 through
26.611, 26.613, 26.615, 26.617, and
26.619, which adds an optional
technology-inclusive, risk-informed,
and performance-based approach for the
application of drug and alcohol testing
and fatigue management requirements
for facilities licensed under part 53.
§ 26.601
Applicability
This proposed rule would add
§ 26.601, which would allow a licensee
or other entity in § 26.3(f) to establish an
FFD program in accordance with the
requirements of subpart M of part 26.
§ 26.603
General Provisions
This proposed rule would add
§ 26.603, which would establish the
general requirements for implementing
an FFD program under subpart M of part
26.
§ 26.604 FFD Program Requirements
for Facilities That Satisfy the § 26.603(c)
Criterion
This proposed rule would add
§ 26.604, which would establish the
FFD program elements for a licensee or
other entity whose facilities and
operations demonstrate compliance
with the criterion in § 26.603(c).
§ 26.605 FFD Program Requirements
for Facilities That Do Not Implement
§ 26.604
This proposed rule would add
§ 26.605, which would establish the
FFD program elements for a licensee or
other entity that does not demonstrate
compliance with the criterion in
§ 26.603(c), or otherwise chooses to
maintain an FFD program under this
section.
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§ 26.606 Written Policies and
Procedures
This proposed rule would add
§ 26.606, which would require licensees
and other entities that implement an
FFD program under subpart M of part 26
to develop a written FFD policy
statement and provide it to all
individuals subject to the FFD program,
and to establish, implement, and
maintain written procedures addressing
the topics outlined in this section.
audits to monitor the effectiveness of
FFD program elements.
§ 26.607 Drug and Alcohol Testing
This proposed rule would add
§ 26.607, which would establish
requirements for licensees and other
entities performing drug and alcohol
testing as part of an FFD program under
subpart M of part 26.
§ 26.619 Suitability and Fitness
Determinations
§ 26.608 FFD Program Training
This proposed rule would add
§ 26.608, which would require
individuals who are subject to the FFD
program under subpart M of part 26 to
receive periodic training on FFD
policies and procedures, including their
duties and responsibilities under the
BOP.
§ 26.609 Behavioral Observation
This proposed rule would add
§ 26.609, which would establish the
requirements for a BOP under subpart M
of part 26.
§ 26.610 Sanctions
This proposed rule would add
§ 26.610, which would require licensees
and other entities implementing an FFD
program under subpart M of part 26 to
establish sanctions for FFD policy
violations.
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§ 26.611 Protection of Information
This proposed rule would add
§ 26.611, which would require licensees
and other entities implementing an FFD
program under subpart M of part 26 to
establish a system to protect personal
information against unauthorized
disclosure.
§ 26.613 Appeals Process
This proposed rule would add
§ 26.613, which would require licensees
and other entities that implement an
FFD program under subpart M of part 26
to establish procedures for an individual
to appeal a policy violation
determination.
§ 26.615 Audits
This proposed rule would add
§ 26.615, which would establish
provisions for licensees and other
entities that implement an FFD program
under subpart M of part 26 to conduct
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§ 26.617
Recordkeeping and Reporting
This proposed rule would add
§ 26.617, which would require licensees
or other entities implementing an FFD
program under subpart M of part 26 to
retain records pertaining to the
administration of the program and to
make reports in accordance with the
requirements of this section.
This proposed rule would add
§ 26.619, which would require licensees
and other entities that implement FFD
programs to develop, implement, and
maintain procedures to assess whether
individuals are fit to perform the duties
that make them subject to the FFD
program.
§ 26.709
Applicability
This proposed rule would designate
the current paragraph as new paragraph
(a), and it would be revised to reference
paragraphs (a) through (d) of § 26.3. It
would also add paragraph (b) to
§ 26.709, which would extend the
requirements of subpart N of part 26 to
licensees and other entities identified in
§ 26.3(f) that do not implement the
requirements of subpart M of part 26, as
well as licensees and other entities that
implement the requirements of
§ 26.605(b).
§ 26.711
General Provisions
This proposed rule would revise
§ 26.711(c) and (d) to incorporate a
reference to § 26.3(f).
§ 26.825
Criminal Penalties
This proposed rule would revise
§ 26.825(b) to include a reference to the
proposed § 26.601.
§ 30.4
Definitions
This proposed rule would revise the
definition for ‘‘Utilization facility’’ in
§ 30.4 to include utilization facilities
defined in the regulations under part 53
in addition to part 50.
§ 30.50
Reporting Requirements
This proposed rule would revise
§ 30.50(c)(3) to include references to
part 53 in addition to part 50.
§ 40.60
Reporting Requirements
This proposed rule would revise
§ 40.60(c)(3) to include references to
part 53 in addition to part 50 regarding
reporting requirements.
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§ 50.47
Emergency Plans
This proposed rule would revise
§ 50.47(a)(1) and (e) with appropriate
references to part 53.
§ 50.54
Conditions of Licenses
This proposed rule would revise
§ 50.54(q)(2), (q)(4), and (gg)(1) with
appropriate references to part 53.
§ 50.160 Emergency Preparedness for
Small Modular Reactors, Non-LightWater Reactors, and Non-Power
Production or Utilization Facilities
This proposed rule would revise
§ 50.160(b)(3) and (c)(2) with the
appropriate references to part 53.
Appendix B to 10 CFR Part 50—Quality
Assurance Criteria for Nuclear Power
Plants and Fuel Reprocessing Plants
This proposed rule would revise
appendix B to part 50 by revising the
introduction and specific criteria to
incorporate the appropriate references
and terminology for part 53.
§ 51.20 Criteria for and Identification
of Licensing and Regulatory Actions
Requiring Environmental Impact
Statements
This proposed rule would revise
§ 51.20(b)(1) and (2) to require an EIS
prior to the issuance of a CP, LWA, or
ESP under part 53, or the issuance to
renewal of a full power or design
capacity license to operate a nuclear
power reactor, testing facility, or fuel
reprocessing plant under part 53.
§ 51.22 Criterion for Categorical
Exclusion; Identification of Licensing
and Regulatory Actions Eligible for
Categorical Exclusion or Otherwise Not
Requiring Environmental Review
This proposed rule would revise
§ 51.22 to include corresponding
references to part 53 in paragraphs
(c)(3), (c)(9), (c)(12), (c)(17), (c)(22) and
(23).
§ 51.26 Requirement To Publish Notice
of Intent and Conduct Scoping Process
This proposed rule would revise
§ 51.26(d) to add a reference to part 53.
§ 51.30
Environmental Assessment
This proposed rule would revise the
introductory text to paragraph (a) and
revise paragraphs (d) and (e) of § 51.30
to incorporate the appropriate
references to part 53 regarding EAs.
§ 51.31 Determinations Based on
Environmental Assessment
This proposed rule would revise
§ 51.31(a) to include a reference to part
53.
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§ 51.32
Impact
§ 51.95 Postconstruction
Environmental Impact Statements
Finding of No Significant
This proposed rule would revise
§ 51.32(b)(1) and (3), finding there is no
significant environmental impact
associated with the issuance of standard
DCs and MLs under part 53.
§ 51.49 Environmental Report-Limited
Work Authorization
This proposed rule would revise the
introductory text of § 51.49(c) to require
applicants for an ESP under part 53
requesting a LWA to include the
environmental report required by
§ 51.50(b).
§ 51.50 Environmental Report—
Construction Permit, Early Site Permit,
or Combined License Stage
This proposed rule would revise
§ 51.50, paragraphs (a), (b)(4), and the
introductory text for paragraph (c) to
incorporate the appropriate references
to part 53.
§ 51.53 Postconstruction
Environmental Reports
This proposed rule would revise
§ 51.53(d) to include the appropriate
references to part 53 regarding a license
termination plan or decommissioning
plan and related requirements for
postconstruction environmental reports.
§ 51.54 Environmental Report—
Manufacturing License
This proposed rule would update
§ 51.54(a) to require applicants for MLs
under part 53 to submit an
environmental report with the
application.
§ 51.55 Environmental Report—
Standard Design Certification
This proposed rule would update
§ 51.55(a) to require applicants for a
standard DC under part 53 to submit an
environmental report with the
application.
§ 51.58 Environmental Report—
Number of Copies; Distribution
This proposed rule would revise
§ 51.58(b) to incorporate the appropriate
references to part 53.
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§ 51.77 Distribution of Draft
Environmental Impact Statement
§ 51.92 Supplement to the Final
Environmental Impact Statement
This proposed rule would revise
§ 51.92(b) to apply to COL applications
referencing an ESP under part 53.
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§ 51.101
Limitations on Actions
This proposed rule would revise
§ 51.101(a)(2) to include the
corresponding references to part 53
where appropriate.
§ 51.103
Record of Decision—General
This proposed rule would update
§ 51.103(a)(6) to apply to the issuance of
a LWA in connection with a CP or COL
under part 53.
§ 51.105 Public Hearings in
Proceedings for Issuance of
Construction Permits or Early Site
Permits; Limited Work Authorizations
This proposed rule would revise
§ 51.105(c)(1) to include the appropriate
reference to LWAs under part 53 for CPs
or ESPs.
§ 51.107 Public Hearings in
Proceedings for Issuance of Combined
Licenses; Limited Work Authorizations
This proposed rule would amend
§ 51.107 by revising the introductory
text for paragraphs (a) and (b) and
updating paragraph (d)(1) to include the
appropriate corresponding references to
part 53.
§ 51.108 Public Hearings on
Commission Findings That Inspections,
Tests, Analyses, and Acceptance
Criteria of Combined Licenses Are Met
This proposed rule would revise
§ 51.108 to incorporate the appropriate
references to part 53.
10 CFR part 53—Risk-Informed,
Technology-Inclusive Regulatory
Framework for Commercial Nuclear
Plants
This proposed rule would add a new
part to 10 CFR Chapter I, designated as
Part 53 including §§ 53.000 through
53.9010.
§ 53.000
This proposed rule would revise the
introductory text for § 51.77(a) to add a
reference to part 53.
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This proposed rule would revise the
introductory text for § 51.95(c) to
include a reference to part 53 regarding
the Commission’s obligations to prepare
an EIS following the renewal of an
operating or COL for a nuclear plant
under part 53.
Purpose
This proposed rule would add
§ 53.000 which provides an optional
technology-inclusive, performancebased framework for the issuance,
amendment, renewal, and termination
of licenses, permits, certifications, and
approvals for commercial nuclear plants
licensed under section 103 of the
Atomic Energy Act of 1954, as amended.
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Subpart A—General Provisions
This proposed rule would add subpart
A, to establish a set of general
provisions, which apply to all
applicants and licensees under part 53.
§ 53.015
Scope
This proposed rule would add
§ 53.015, which would extend the
provisions of subpart A to all applicants
and licensees under part 53.
§ 53.020
Definitions
This proposed rule would add
§ 53.020, which would define key terms
in part 53.
§ 53.040
Written Communications
This proposed rule would add
§ 53.040, which would govern how
applicants and licensees submit written
communications to the NRC, including
applications, submissions related to the
security plans, emergency plan, and
quality assurance, certifications of
permanent cessation of operations and
permanent fuel removal, and other
submittals required under part 53.
§ 53.050
Deliberate Misconduct
This proposed rule would add
§ 53.050, which would prohibit
licensees or applicants, contractors and
subcontractors, or employees of those
entities from deliberately violating NRC
rules, regulations, or orders, or the
terms, conditions, and limitations of a
part 53 license. This proposed rule
would also prohibit deliberate
submissions of incomplete or inaccurate
information. Violations would be
subject to enforcement actions under
subpart B of part 2.
§ 53.060
Employee Protection
This proposed rule would add
§ 53.060, which would prohibit
applicants and licensees from
discriminating against employees for
engaging in the protected activities
listed in this section and provide
remedial procedures for employees who
believe they are the subjects of
discrimination.
§ 53.070 Completeness and Accuracy
of Information
This proposed rule would add
§ 53.070, which would require licensees
and applicants under part 53 to provide
complete and accurate information in
accordance with all applicable laws,
Commission regulations, and the terms
and conditions of their license. This
proposed rule would also require
licensees to notify the Commission
within two days of identifying
information with material implications
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for public health and safety or common
defense and security.
§ 53.080
Specific Exemptions
This proposed rule would add
§ 53.080, which would establish the
special circumstances under which the
Commission could grant exemptions to
part 53 licensees and the Commission’s
criteria for making such a
determination.
§ 53.090
Standards for Review
This proposed rule would add
§ 53.090 to establish the standards that
the Commission would consider when
determining whether to issue a permit
or license under part 53.
§ 53.100
Jurisdictional Limits
This proposed rule would add
§ 53.100, which would provide that
permits, licenses, standard design
approvals, and standard DCs are solely
issued for activities within the
jurisdiction of the United States.
§ 53.110
Attacks and Destructive Acts
This proposed rule would add
§ 53.110, which would exempt licensees
or applicants under part 53 from
providing design features to protect
against attacks or destructive acts
directed at the facility by United States
adversaries.
§ 53.115 Rights Related to Special
Nuclear Material
This proposed rule would add
§ 53.115, which would establish
provisions regarding the rights to SNM
under a part 53 license.
§ 53.117 License Suspension and
Rights of Recapture
This proposed rule would add
§ 53.117, which would provide that the
Commission may suspend licenses and
recapture material or control of a facility
in a state of war or national emergency
declared by Congress.
§ 53.120 Information Collection
Requirements: OMB Approval
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This proposed rule would add
§ 53.120, which would establish
requirements for information collection
requirements and Office of Budget and
Management approval.
§ 53.210 Safety Criteria for DesignBasis Accidents
This proposed rule would add
§ 53.210 to set dose values to ensure that
plants are designed to limit the public’s
radiation exposure in the event of a
DBA.
§ 53.220 Safety Criteria for LicensingBasis Events Other Than Design-Basis
Accidents
This proposed rule would add
§ 53.220 to require plants to implement
a combination of design features and
programmatic controls to control risks
to the public in the event of a LBE other
than a DBA.
§ 53.230 Safety Functions
This proposed rule would add
§ 53.230, which specifies that limiting
the release of radioactive materials from
the facility is the primary safety
function of a commercial nuclear plant,
and that additional safety functions
must be defined to support the retention
of radioactive materials during LBEs.
§ 53.240 Licensing-Basis Events
This proposed rule would add
§ 53.240 to require commercial nuclear
plants to conduct an analysis of LBEs to
confirm that design features and
programmatic controls satisfy the safety
criteria under §§ 53.210 and 53.220, or
alternatively, under § 53.470.
§ 53.250 Defense in Depth
This proposed rule would add
§ 53.250 to establish a performancebased, defense-in-depth approach to
address uncertainties about the
effectiveness and reliability of plant
SSCs, personnel, and programmatic
controls.
§ 53.260 Normal Operations
This proposed rule would add
§ 53.260, requiring holders of licenses to
operate commercial nuclear plants to
control public doses and dose rates in
unrestricted areas to meet the
requirements in part 20, during normal
plant operation.
Subpart B—Technology-Inclusive Safety
Requirements
§ 53.270 Protection of Plant Workers
This proposed rule would add
§ 53.270, requiring holders of licenses to
operate commercial nuclear plants to
control occupational doses to meet the
requirements in part 20.
This proposed rule would add subpart
B, to establish a set of technologyinclusive performance standards that
would be used throughout part 53 to
determine appropriate regulatory
controls for SSCs, human actions, and
programs.
Subpart C—Design and Analysis
Requirements
This proposed rule would add subpart
C, which requires the implementation of
certain design features and the
performance of risk assessments and
analyses to demonstrate compliance
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with the safety criteria and safety
functions in subpart B.
§ 53.400 Design Features for LicensingBasis Events
This proposed rule would add
§ 53.400, which would require design
features that satisfy the safety criteria
defined in § 53.210 and § 53.220 or
§ 53.470 and fulfill the safety functions
identified in § 53.230 during LBEs.
§ 53.410 Functional Design Criteria for
Design-Basis Accidents
This proposed rule would add
§ 53.410, which would stipulate that
functional design criteria must be
defined for each design feature required
by § 53.400 to demonstrate compliance
with the safety criteria defined in
§ 53.210 for DBAs.
§ 53.415 Protection Against External
Hazards
This proposed rule would add
§ 53.415, which would require SR SSCs
to be designed to withstand the effects
of natural phenomena and constructed
hazards while performing the intended
safety functions.
§ 53.420 Functional Design Criteria for
Licensing-Basis Events Other Than
Design-Basis Accidents
This proposed rule would add
§ 53.420, which would require
functional design criteria to be defined
for each design feature required by
§ 53.400 to demonstrate compliance
with the safety criteria defined in
§ 53.220 for LBEs other than DBAs.
§ 53.425 Design Features and
Functional Design Criteria for Normal
Operations
This proposed rule would add
§ 53.425, which would require
commercial nuclear plants to implement
design features and define functional
design criteria sufficient to demonstrate
compliance with § 53.850 and show
through functional design criteria that
design features and corresponding
programmatic controls control wastes,
as required under part 20.
§ 53.430 Design Features and
Functional Design Criteria for Protection
of Plant Workers
This proposed rule would add
§ 53.430, which would require
commercial nuclear plants to implement
design features and define functional
design criteria sufficient to demonstrate
compliance with § 53.270.
§ 53.440
Design Requirements
This proposed rule would add
§ 53.440, which would establish various
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design feature requirements, including
protection against fires and explosions,
criticality accidents, and the impact of
a large commercial aircraft.
§ 53.450
Analysis Requirements
This proposed rule would add
§ 53.450, which would require
commercial nuclear plants to perform
PRAs in combination with other
analytical methods to identify and
assess risks and determine compliance
with the safety criteria in subpart B. In
addition, § 53.450 would require
analysis of DBAs and other analyses to
assess the adequacy of protections
against fire, aircraft impact, and the
release of effluents.
§ 53.460 Safety Categorization and
Special Treatments
This proposed rule would add
§ 53.460 to address the safety
classification of SSCs and determine
appropriate special treatments.
This proposed rule would add
§ 53.470 to permit applicants and
licensees to implement more restrictive
criteria than that defined in §§ 53.220
and 53.450(e) to support operational
flexibilities.
Earthquake Engineering
This proposed rule would add
§ 53.480 to provide overall seismic
design considerations based on the
safety criteria in subpart B and siting
requirements in subpart D to ensure that
SSCs are able to withstand the effects of
earthquakes without loss of capability to
fulfill safety functions.
Subpart D—Siting Requirements
This proposed rule would add subpart
D, which would address requirements
associated with the siting of commercial
nuclear facilities under part 53,
including considerations of external
hazards and potential adverse impacts
on the surrounding population.
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§ 53.500 General Siting and Siting
Assessment
This proposed rule would add
§ 53.500, which would require a siting
assessment for each commercial nuclear
plant to ensure that design features and
programmatic controls are sufficient to
address LBEs and mitigate potential
adverse impacts of the plant on the
surrounding environs.
§ 53.510
External Hazards
This proposed rule would add
§ 53.510, which would require site-
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§ 53.520
Site Characteristics
This proposed rule would add
§ 53.520, which would require the
design and analyses conducted under
subpart C to consider how site
characteristics may contribute to LBEs.
§ 53.530 Population-Related
Considerations
This proposed rule would add
§ 53.530, which would establish
requirements related to the facility’s
exclusion area, low-population zone,
and population center distance.
§ 53.540
§ 53.470 Maintaining Analytical Safety
Margins Used To Justify Operational
Flexibilities
§ 53.480
specific assessments, including an
evaluation of geological and seismic
siting factors, to identify and
characterize the external hazard level
for a range of natural and constructed
hazards.
Siting Interfaces
This proposed rule would add
§ 53.540, which would require that
external hazards and site characteristics
must be accounted for in the design
features, programmatic controls, and
supporting analyses used to
demonstrate compliance with the safety
criteria in §§ 53.210 and 53.220.
Subpart E—Construction and
Manufacturing Requirements
This proposed rule would add subpart
E, which would establish requirements
for the construction and manufacture of
commercial nuclear plants.
§ 53.600 Construction and
Manufacturing—Scope and Purpose
This proposed rule would add
§ 53.600, which would indicate that this
subpart applies to construction and
manufacturing activities authorized by a
CP, COL, ML, or LWA issued under this
part.
§ 53.605 Reporting of Defects and
Noncompliance
This proposed rule would add
§ 53.605, which would describe the
procedures, notification requirements,
and records retention requirements that
each CP, ML, and COL is subject to with
respect to reporting of defects and
noncompliance.
§ 53.610
Construction
This proposed rule adds § 53.610 to
address the management and control of
the construction of a commercial
nuclear plant, including specific
requirements for procedures and quality
assurance, control of radioactive
materials, and post construction
inspections.
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§ 53.620 Manufacturing
This proposed rule would add
§ 53.620, which would ensure that the
holders of an ML under part 53 develop
plans, programs, and organizational
units to manage and control
manufacturing activities, and would
establish requirements for the loading of
fuel into a manufactured reactor for
subsequent transport to a commercial
nuclear plant and operation pursuant to
a COL.
Subpart F—Requirements for Operation
This proposed rule would add subpart
F, which would establish regulatory
requirements to ensure that the safety
criteria in subpart B are satisfied
whenever a commercial nuclear plant
licensed under part 53 is operational.
This includes periods of normal
operation and unplanned events.
§ 53.700 Operational Objectives
This proposed rule would add
§ 53.700, which would establish general
operational objectives to ensure that
licensees under part 53 have
implemented and maintained the SSCs
necessary to demonstrate compliance
with the safety functions identified in
subpart B for addressing normal
operations and responding to LBEs.
§ 53.710 Maintaining Capabilities and
Availability of Structures, Systems, and
Components
This proposed rule would add
§ 53.710, which would require licensees
under part 53 to demonstrate
compliance with the safety criteria in
subpart B by establishing TS for all SR
SSCs and developing documents and
procedures for all NSRSS SSCs.
§ 53.715 Maintenance, Repair, and
Inspection Programs
This proposed rule would add
§ 53.715, which would require licensees
to develop, implement, and maintain
programs to assess and manage any risks
posed by maintenance activities and to
evaluate the efficacy of performance,
condition monitoring, and maintenance
activities.
§ 53.720 Response to Seismic Events
This proposed rule would add
§ 53.720, which would establish
requirements for licensees to respond to
a seismic event during the operating
phase of the life cycle of a commercial
nuclear plant.
§ 53.725 General Staffing, Training,
Personnel Qualifications, and Human
Factors Requirements
This proposed rule would add
§ 53.725, which would provide an
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overview of the staffing, training,
personnel qualifications, and human
factors requirements established in
§§ 53.725 through 53.830 and would
provide definitions of ‘‘Automation,’’
‘‘Auxiliary operator,’’ ‘‘Controls,’’
‘‘Generally licensed reactor operator,’’
‘‘Load following,’’ ‘‘Operator,’’
‘‘Performance testing,’’ ‘‘Reference
plant,’’ ‘‘Self-reliant mitigation facility,’’
‘‘Senior operator,’’ ‘‘Simulation
facility,’’ and ‘‘Systems approach to
training.’’ Proposed §§ 53.725 through
53.830 would apply to applicants for or
holders of OLs or COLs under part 53.
§ 53.726
Communications
This proposed rule would add
§ 53.726, which would contain
communications requirements
applicable to sections §§ 53.725 through
53.830. It also contains requirements to
notify the Commission within 30 days
should a specifically licensed operator
or senior operator be reassigned,
terminated, or suffer permanent
disability or illness.
§ 53.728 Completeness and Accuracy
of Information
This proposed rule would add
§ 53.728, which would require
submitted information to be complete
and accurate in all material respects.
This proposed rule would add
§ 53.730, which would establish
technical requirements for applicants or
holders of OLs or COLs within the areas
of HFE, human-system interface design,
concept of operations, functional
requirements analysis, function
allocation, operating experience,
procedures, staffing, operator training,
operator examinations, and operator
proficiency.
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Operator Licensing
This proposed rule would add
§ 53.760, which would address the
applicability of the requirements of
§§ 53.760 through 53.795 for specifically
licensed operators and senior operators.
§ 53.765
Medical Requirements
This proposed rule would add
§ 53.765, which would establish
medical requirements for specifically
licensed operators and senior operators.
§ 53.770 Incapacitation Because of
Disability or Illness
This proposed rule would add
§ 53.770, which would establish
requirements to address permanent
medical conditions for specifically
licensed operators and senior operators.
§ 53.775 Applications for Operators
and Senior Operators
This proposed rule would add
§ 53.775, which would establish the
application process and requirements
for individuals applying for specific
operator and senior operator licenses.
§ 53.780 Training, Examination, and
Proficiency Program
§ 53.730 Defining, Fulfilling, and
Maintaining the Role of Personnel in
Ensuring Safe Operations
§ 53.735
§ 53.760
General Exemptions
This proposed rule would add
§ 53.780, which would contain the
requirements associated with
specifically licensed operator and senior
operator initial training, initial
examinations, requalification training,
requalification examinations,
examination integrity, simulation
facilities, waivers, and proficiency.
§ 53.785 Conditions of Operator and
Senior Operator Licenses
This proposed rule would add
§ 53.785, which would establish
conditions for specific operator and
senior operator licenses.
This proposed rule would add
§ 53.735, which would establish general
exemptions for licensed operators.
§ 53.790 Issuance, Modification, and
Revocation of Operator and Senior
Operator Licenses
§ 53.740 Facility Licensee
Requirements—General
This proposed rule would add
§ 53.790, which would contain
requirements associated with the
issuance, modification, or revocation of
specific operator and senior operator
licenses.
This proposed rule would add
§ 53.740, which would establish staffing
requirements for interaction-dependentmitigation facilities and self-reliant
mitigation facilities.
§ 53.745 Operator License
Requirements
This proposed rule would add
§ 53.745, which would require
individuals to be licensed to perform
certain functions.
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§ 53.795 Expiration and Renewal of
Operator and Senior Operator Licenses
This proposed rule would add
§ 53.795, which would contain
requirements associated with the
expiration and renewal of specific
operator and senior operator licenses.
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§ 53.800 Facility Licensees for SelfReliant-Mitigation Facilities
This proposed rule would add
§ 53.800, which would establish the
technical criteria by which commercial
nuclear plants under part 53 are
determined to be of the self-reliant
mitigation class of facilities that would
be staffed by GLROs in lieu of
specifically licensed operators and
senior operators.
§ 53.805 Facility Licensee
Requirements Related to Generally
Licensed Reactor Operators
This proposed rule would add
§ 53.805, which would establish
requirements that apply to the facility
licensee at those facilities staffed by
GLROs.
§ 53.810 Generally Licensed Reactor
Operators
This proposed rule would add
§ 53.810, which would issue and
describe the general license for GLROs
that manipulate the controls of a selfreliant mitigation facility.
§ 53.815 Generally Licensed Reactor
Operator Training, Examination, and
Proficiency Programs
This proposed rule would add
§ 53.815, which would contain the
requirements for GLRO initial training,
initial examinations, continuing
training, requalification examinations,
examination integrity, simulation
facilities, examination waivers, and
proficiency.
§ 53.820 Cessation of Individual
Applicability
This proposed rule would add
§ 53.820, which would address the
requirements by which the general
license for GLROs would cease to be
applicable on an individual basis.
§ 53.830 Training and Qualification of
Commercial Nuclear Plant Personnel
This proposed rule would add
§ 53.830, which would address training
and qualification requirements for
supervisors, technicians, and other
appropriate operating personnel at
commercial nuclear plants.
§ 53.845
Programs
This proposed rule would add
§ 53.845, which would require licensees
under part 53 to establish programs that
include, but are not limited to, radiation
protection, emergency preparedness,
security, quality assurance, integrity
assessment, fire protection, ISI and IST,
and facility safety, to ensure that the
safety criteria and functions in subpart
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B are maintained during normal
operations and LBEs.
Subpart G—Decommissioning
Requirements
for using decommissioning trust funds
and required terms.
§ 53.850
This proposed rule would add subpart
G, to establish decommissioning
requirements for applicants for or
holders of an OL or COL under part 53.
§ 53.1050
Radiation Protection
This proposed rule would add
§ 53.850, which would require licensees
under part 53 to implement and
maintain programs and processes to
limit and monitor radioactive plant
effluents and limit the exposure of plant
personnel and the public.
§ 53.855
Emergency Preparedness
This proposed rule would add
§ 53.855, which would require licensees
under this part to have an emergency
response plan for radiological
emergencies.
§ 53.860
Security Programs
This proposed rule would add
§ 53.860, which would require licensees
under part 53 to develop, implement,
and maintain programs for physical
security, FFD, AA, cybersecurity, and
information security.
§ 53.865
Quality Assurance
This proposed rule would add
§ 53.865, which would require licensees
under part 53 to establish a quality
assurance program that includes a
written manual to ensure activities are
conducted in accordance with codes
and standards found acceptable by the
NRC.
§ 53.870 Integrity Assessment
Programs
This proposed rule would add
§ 53.870, which would require licensees
under part 53 to establish an integrity
assessment program to ensure that the
plant continues to fulfill safety criteria
and functional design criteria as it ages.
§ 53.875
Fire Protection
This proposed rule would add
§ 53.875, which would require licensees
under part 53 to establish a fire
protection plan and describe the
necessary elements that the plan must
incorporate.
§ 53.880 Inservice Inspection and
Inservice Testing
This proposed rule would add
§ 53.880, which would require licensees
under part 53 to develop and implement
a program for ISI and IST in accordance
with the requirements of this section.
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§ 53.910
Procedures and Guidelines
This proposed rule would add
§ 53.910, which would require licensees
under part 53 to develop, maintain, and
implement procedures and guidelines
that address normal plant operations
and responses to unplanned events.
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§ 53.1000
Scope and Purpose
This proposed rule would add
§ 53.1000, which would establish the
scope of the decommissioning
requirements for applicants and
licensees under part 53 and describe the
contents of subpart G of part 53.
§ 53.1010 Financial Assurance for
Decommissioning
This proposed rule would add
§ 53.1010, which would establish the
requirement that applicants for an OL or
COL under part 53 provide reasonable
assurance that funds will be available
for the decommissioning process. This
section would describe the requirements
associated with the required plan and
an associated decommissioning report
that ensures and documents that
adequate funding for decommissioning
will be available.
§ 53.1020 Cost Estimates for
Decommissioning
This proposed rule would add
§ 53.1020, which would require sitespecific cost estimates for
decommissioning and establish the
aspects that must be included in the
estimate.
§ 53.1030 Annual Adjustments to Cost
Estimates for Decommissioning
This proposed rule would add
§ 53.1030, which would require that
holders of an OL or COL under part 53
annually adjust their cost estimate for
decommissioning to account for
escalation in labor, energy, and waste
burial costs. This section would allow
licensees to elect either a site-specific
adjustment factor or a generic
adjustment factor.
§ 53.1040 Methods for Providing
Financial Assurance for
Decommissioning
This proposed rule would add
§ 53.1040, which would establish
suitable methods that holders of an OL
or COL under part 53 may use to
provide financial assurance for
decommissioning to the NRC.
§ 53.1045 Limitations on the Use of
Decommissioning Trust Funds
This proposed rule would add
§ 53.1045, which would establish
requirements for decommissioning trust
funds under part 53, including criteria
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NRC Oversight
This proposed rule would add
§ 53.1050, which would outline the
steps the NRC may take to ensure
adequate accumulation of
decommissioning funds.
§ 53.1060 Reporting and
Recordkeeping Requirements
This proposed rule would add
§ 53.1060, which would contain
reporting and recordkeeping
requirements related to
decommissioning for each holder of an
OL or COL under part 53. This section
would outline requirements for
documents such as: certification of
decommissioning funding,
decommissioning cost estimates and
copies of financial instruments, licensee
records of information important to safe
and effective decommissioning, postshutdown decommissioning activities
report, financial assurance reports, and
reports on the status of funding for
managing irradiated fuel.
§ 53.1070
Termination of License
This proposed rule would add
§ 53.1070, which would establish
procedures for decommissioning and
license termination applicable to
licensees under part 53 that have
determined to permanently cease
operations.
§ 53.1075 Program Requirements
During Decommissioning
This proposed rule would add
§ 53.1075, which would require
licensees under part 53 to establish and
maintain a decommissioning fire
protection program to prevent, detect,
and control fires, and ensure that the
risk of fire induced radiological hazards
are minimized through the various
stages of facility decommissioning.
§ 53.1080 Release of Part of a
Commercial Nuclear Plant or Site for
Unrestricted Use
This proposed rule would add
§ 53.1080, which would establish
licensee procedures for requesting and
NRC procedures for approving partial
release of a commercial nuclear plant or
site for unrestricted use prior to
receiving approval of a license
termination plan from the Commission
under part 53.
Subpart H—Licenses, Certifications, and
Approvals
This proposed rule would add subpart
H, which would govern the process of
applying for, amending, renewing, or
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terminating a LWA, ESP, standard
design approval, standard DC, ML, CP,
OL, or COL under part 53.
§ 53.1100 Filling of Application for
Licenses, Certifications, or Approvals;
Oath or Affirmation
This proposed rule would add
§ 53.1100, which would establish
requirements for applicants seeking a
standard design approval, standard DC,
license, or permit under part 53 to
submit an application.
§ 53.1101
Requirement for License
This proposed rule would add
§ 53.1101, which would prohibit any
use of a utilization facility except as
authorized by a license issued by the
NRC or by an exception as described in
§ 53.1120.
§ 53.1103
Licenses
Combining Applications and
filed an application for a CP or COL for
that site.
This proposed rule would add
§ 53.1115, which would require
applicants to agree in writing, prior to
receiving a license or standard design
approval under part 53, to restrict any
facilities, or any individuals with access
to plant facilities, from possessing
Restricted Data or classified National
Security Information until they have
received the appropriate authorization.
§ 53.1144 Contents of Applications for
Early Site Permits; General Information
§ 53.1118 Ineligibility of Certain
Applicants
This proposed rule would add
§ 53.1118, which would prevent
citizens, nationals, or agents of a foreign
country or corporations owned,
controlled, or dominated by a foreign
entity from applying for or obtaining a
license under part 53.
§ 53.1120 Exceptions and Exemptions
From Licensing Requirements
This proposed rule would add
§ 53.1103, which would permit
applicants under part 53 seeking
multiple licenses to submit a single
application, and the Commission to
issue a single license for activities that
would otherwise be licensed separately.
This proposed rule would add
§ 53.1120, which would establish the
activities that are exempt from licensing
requirements.
§ 53.1106
This proposed rule would add
§ 53.1121, which would allow applicant
submissions to be made publicly
available under the provisions of part 2.
Elimination of Repetition
This proposed rule would add
§ 53.1106, which would allow
applicants under part 53 to reference
information contained in previous
documents filed with the Commission
so long as those references are clear and
specific.
§ 53.1109 Contents of Applications;
General Information
This proposed rule would add
§ 53.1109, which would establish the
general content to be included in
applications made under part 53,
including but not limited to the
identifying information of the applicant
and the radiological emergency
response plans of government entities
within the plume exposure pathway
EPZ.
§ 53.1112
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§ 53.1115 Agreement Limiting Access
to Classified Information
Environmental Conditions
This proposed rule would add
§ 53.1112, which would allow the
Commission to attach conditions to CPs,
ESPs, and licenses issued under part 53
to address environmental issues during
construction, operation, or
decommissioning. These conditions will
be derived from the information
contained in the environmental report
submitted as part of the application for
a permit or license.
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§ 53.1121 Public Inspection of
Applications
§ 53.1124
Sections
Relationship Between
This proposed rule would add
§ 53.1124, which would outline the
relationship between LWAs, ESPs,
standard design approvals, standard
DCs, MLs, CPs, OLs, and COLs under
part 53.
§ 53.1130 Limited Work
Authorizations
This proposed rule would add
§ 53.1130, which would establish
requirements for requesting an LWA
and grounds for the Commission to
issue an LWA. It would also contain
details about the effect of an LWA and
the implementation of a redress plan.
§ 53.1140
Early Site Permits
This proposed rule would add
§ 53.1140, which would provide an
overview of the requirements regarding
applications for and the issuance of
ESPs under part 53.
§ 53.1143
Filing of Applications
This proposed rule would add
§ 53.1143, which would enable an
applicant under part 53 to apply for an
ESP, regardless of whether they have
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This proposed rule would add
§ 53.1144, which would require
applications for ESPs to include the
information required by § 53.1109(a)
through (d) and (j).
§ 53.1146 Contents of Applications for
Early Site Permits; Technical
Information
This proposed rule would add
§ 53.1146, which would require
applicants for ESPs to submit technical
information, including but not limited
to a Site Safety Analysis Report and
emergency plans.
§ 53.1149
Review of Applications
This proposed rule would add
§ 53.1149, which would establish
standards for review of applications for
ESPs under part 53, including
requirements for the Commission to
prepare an EIS and assess the adequacy
of protective actions in the event of a
radiological emergency. It would also
require the administrative review of
applications and hearings to follow the
procedural requirements of part 2.
§ 53.1155 Referral to the Advisory
Committee on Reactor Safeguards
This proposed rule would add
§ 53.1115, which would require the
ACRS to review SR content in the
application for an ESP under part 53.
§ 53.1158
Issuance of Early Site Permit
This proposed rule would add
§ 53.1158, which would establish the
conditions under which the
Commission may issue an ESP under
part 53, as well as the information,
terms, and conditions to be included in
the permit.
§ 53.1161 Extent of Activities
Permitted
This proposed rule would add
§ 53.1161, which would require that a
valid ESP only be used for the purpose
of site redress, unless the site is
referenced in an application for a CP or
COL under part 53.
§ 53.1164
Duration of Permit
This proposed rule would add
§ 53.1164, which would govern the
conditions under which an ESP remains
valid following the date of issuance.
§ 53.1167 Limited Work Authorization
After Issuance of Early Site Permit
This proposed rule would add
§ 53.1167, which would permit the
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holder of an ESP to request a LWA
under § 53.1130.
§ 53.1170
Transfer of Early Site Permit
This proposed rule would add
§ 53.1170, which would govern the
transfer of an ESP in accordance with
§ 53.1570.
§ 53.1173
Application for Renewal
This proposed rule would add
§ 53.1173, which would establish the
conditions and procedures for renewing
an ESP under part 53.
§ 53.1176
Criteria for Renewal
This proposed rule would add
§ 53.1176, which would establish the
criteria that the Commission may use to
grant a renewal of an ESP under part 53.
§ 53.1179
Duration of Renewal
This proposed rule would add
§ 53.1179, which would govern the
duration of a renewed ESP under part
53.
§ 53.1182
Purposes
Use of Site for Other
This proposed rule would add
§ 53.1182, which would govern
acceptable uses of the site for purposes
other than those described in the
permit.
§ 53.1188 Finality of Early Site Permit
Determinations
This proposed rule would add
§ 53.1188, which would address the
finality of ESP determinations under
part 53.
§ 53.1200
Standard Design Approvals
This proposed rule would add
§ 53.1200, which would address the
procedures for filing an application for
a standard design approval under part
53, the process of review by NRC staff,
and referral to the ACRS of standard
designs.
§ 53.1203
Filing of Applications
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This proposed rule would add
§ 53.1203, which would enable
applicants to submit a final design for
the entire facility, or major portions, to
the NRC staff for review.
§ 53.1206 Contents of Applications for
Standard Design Approvals; General
Information
This proposed rule would add
§ 53.1206, which would require
applications for a standard design
approval under part 53 to contain the
information required by § 53.1109(a)
through (c) and (j).
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§ 53.1209 Contents of Applications for
Standard Design Approvals; Technical
Information
This proposed rule would add
§ 53.1209, which would require the
inclusion of certain technical
information, including a FSAR, site
parameters, and design information,
when an applicant seeks review of
major portions of a standard design.
§ 53.1210 Contents of Applications for
Standard Design Approvals; Other
Application Content
This proposed rule would add
§ 53.1210, which would require
applications for standard design
approvals under part 53 to include a
description of the availability controls
used to satisfy the safety criteria of
§ 53.220, the program to protect
Safeguards Information against
unauthorized disclosure, evidence that
safety questions associated with SSCs
have been resolved, and a description of
how design features fulfill design
criteria.
§ 53.1212 Standards for Review of
Applications
This proposed rule would add
§ 53.1212, which would require
applications for standard design
approval to be reviewed under the
standards in parts 20, 53, and 73.
§ 53.1215 Referral to the Advisory
Committee on Reactor Safeguards
This proposed rule would add
§ 53.1215, which would require the
ACRS to report on any portions of the
application for a standard design
approval under part 53 concerning
safety.
§ 53.1218 Staff Approval of Design
This proposed rule would add
§ 53.1218, which would require the NRC
staff to make a determination on the
acceptability of the design, publish its
decision in the Federal Register, and
issue a report analyzing the design that
is available at https://nrc.gov.
Additionally, the rule would establish
the conditions under which a design
approval under part 53 remains valid.
§ 53.1221 Finality of Standard Design
Approvals; Information Requests
This proposed rule would add
§ 53.1221, which would require NRC
staff and the ACRS to rely upon an
approved design in their review of any
standard DC, ML, or individual facility
license application under part 53 that
references the standard design approval.
The proposed rule would also govern
requirements for issuing information
requests.
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§ 53.1230 Standard Design
Certifications
This proposed rule would add
§ 53.1230, which would provide an
overview of the requirements and
procedures that govern the issuance of
standard DCs under part 53.
§ 53.1233
Filing of Applications
This proposed rule would add
§ 53.1233, which would enable an
application for DC to be filed, regardless
of whether an application for a CP, COL,
or ML has been filed, provided it
complies with the filing requirements in
§ 53.040 and §§ 2.811 through 2.819.
§ 53.1236 Contents of Applications for
Standard Design Certifications; General
Information
This proposed rule would add
§ 53.1236, which would require an
application for a standard DC under part
53 to contain all of the information
required by § 53.1109(a) through (c) and
(j).
§ 53.1239 Contents of Applications for
Standard Design Certifications;
Technical Information
This proposed rule would add
§ 53.1239, which would require
applicants for a standard DC under part
53 to submit a FSAR that includes
technical design information at a level
of detail sufficient to enable the
Commission to make a safety
determination.
§ 53.1241 Contents of Applications for
Standard Design Certifications; Other
Application Content
This proposed rule would add
§ 53.1241, which would require
applications for standard DCs under
part 53 to include an environmental
report, as well as a description of the
availability controls used to satisfy the
safety criteria of § 53.220, proposed
ITAAC, the program to protect
Safeguards Information against
unauthorized disclosure, evidence that
safety questions associated with SSCs
have been resolved, and a description of
how design features fulfill design
criteria.
§ 53.1242
Review of Applications
This proposed rule would add
§ 53.1242, which would require
applications for standard DCs to be
reviewed for compliance with the
standards in parts 20, 51, 53, and 73. It
would also establish procedural
requirements for reviewing applications
and holding hearings in accordance
with subpart H of part 2.
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§ 53.1273
§ 53.1245 Referral to the Advisory
Committee on Reactor Safeguards
This proposed rule would add
§ 53.1245, which would require the
ACRS to report on any portions of the
application for a standard DC under part
53 concerning safety.
§ 53.1248 Issuance of Standard Design
Certification
This proposed rule would add
§ 53.1248, which would establish the
conditions under which the
Commission may issue a DC rule that
specifies the site parameters, design
characteristics, and any additional terms
and conditions of the DC rule.
§ 53.1251
Duration of Certification
This proposed rule would add
§ 53.1251, which would set the
conditions under which a standard DC
remains valid.
§ 53.1254
Application for Renewal
This proposed rule would add
§ 53.1254, which would establish the
conditions and procedures for renewing
a standard DC under part 53.
§ 53.1257
Criteria for Renewal
This proposed rule would add
§ 53.1257, which would enable the
Commission to issue a rule granting the
renewal of a standard DC under part 53,
impose additional requirements, and
grant amendment requests.
§ 53.1260
Duration of Renewal
This proposed rule would add
§ 53.1260, which would provide that a
renewal of a standard DC under part 53
is valid for not less than 10 years, nor
more than 15 years.
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§ 53.1263 Finality of Standard Design
Certifications
This proposed rule would add
§ 53.1263, which would establish
limited conditions under which the
Commission may initiate a rulemaking
to modify, rescind, or impose new
requirements on a standard DC rule
under part 53. It would also address
requests for an exemption from
elements of the certification
information, and require that applicants
for a CP, COL, or ML that references a
DC rule make information normally
contained in engineering documents
available for audit.
§ 53.1270
Manufacturing Licenses
This proposed rule would add
§ 53.1270, which would provide an
overview of the requirements and
procedures for applying for and issuing
an ML under part 53.
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Filing of Applications
This proposed rule would add
§ 53.1273, which would establish the
requirements to apply for an ML under
part 53.
§ 53.1276 Contents of Applications for
Manufacturing Licenses; General
Information
This proposed rule would add
§ 53.1276, which would require
applicants for an ML under part 53 to
include the information contained in
§ 53.1109(a) through (e) and (j).
§ 53.1279 Contents of Applications for
Manufacturing Licenses; Technical
Information
This proposed rule would add
§ 53.1279, which would require an
applicant for an ML under part 53 to
include certain technical information in
a FSAR, including but not limited to
information about site parameters,
design information, manufacturing
information, and information related to
the potential fueling and ultimate
deployment of a completed
manufactured reactor.
§ 53.1282 Contents of Applications for
Manufacturing Licenses; Other
Application Content
This proposed rule would add
§ 53.1282, which would require
applicants for an ML under part 53 to
include in their application the
proposed ITAAC, an environmental
report, a description of the program to
protect Safeguards Information against
unauthorized disclosure, and a
description of how design features
fulfill design criteria. It would also
include content requirements for the
ITAAC and environmental reports in
applications that reference a standard
DC.
§ 53.1285
Review of Applications
This proposed rule would add
§ 53.1285, which would require
applications for MLs under part 53 to be
reviewed for compliance with
applicable standards and establish
procedural requirements for reviewing
applicants and holding hearings in
accordance with part 2.
§ 53.1286 Referral to the Advisory
Committee on Reactor Safeguards
This proposed rule would add
§ 53.1286, which would require the
ACRS to report on any portions of the
application for an ML under part 53
concerning safety.
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§ 53.1287 Issuance of Manufacturing
Licenses
This proposed rule would add
§ 53.1287, which would establish the
conditions under which the
Commission may issue an ML under
part 53.
§ 53.1288 Finality of Manufacturing
Licenses
This proposed rule would add
§ 53.1288, which would address the
limited circumstances in which the
Commission may modify, rescind, or
impose new requirements following the
issuance of an ML under part 53. It
would also address requests for a
departure from the specifications of the
license.
§ 53.1291 Duration of Manufacturing
Licenses
This proposed rule would add
§ 53.1291, which would govern the
expiration of an ML, which is valid for
no less than 5, nor more than 15 years
from the date of issuance.
§ 53.1293 Transfer of Manufacturing
Licenses
This proposed rule would add
§ 53.1293, which would provide that an
ML under part 53 may be transferred in
accordance with § 53.1570.
§ 53.1295 Renewal of Manufacturing
Licenses
This proposed rule would add
§ 53.1295, which would establish the
procedures for applicants to apply for
and the Commission to grant a renewal
of an ML under part 53.
§ 53.1300 Construction Permits
This proposed rule would add
§ 53.1300, which would provide an
overview of the requirements and
procedures for applicants to apply for
and the Commission to grant a CP under
part 53.
§ 53.1306 Contents of Applications for
Construction Permits; General
Information
This proposed rule would add
§ 53.1306, which would require
applicants for a CP under part 53 to
submit the general information required
by § 53.1109, as well as financial
information.
§ 53.1309 Contents of Applications for
Construction Permits; Technical
Information
This proposed rule would add
§ 53.1309, which would require
applicants for a CP under part 53 to
submit a PSAR and a description of the
program to protect Safeguards
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Information from unauthorized
disclosure.
§ 53.1312 Contents of Applications for
Construction Permits; Other Application
Content
This proposed rule would add
§ 53.1312, which would require
applicants for a CP under part 53 to
submit an environmental report and to
provide additional details in the PSAR
if the application references an ESP,
standard design approval, or standard
DC.
§ 53.1315
Review of Applications
This proposed rule would add
§ 53.1315, which would require
applications for CPs under part 53 to be
reviewed for compliance with
applicable standards and establish
procedural requirements for reviewing
applications and holding hearings in
accordance with part 2.
§ 53.1318 Finality of Referenced NRC
Approvals, Permits, and Certifications
This proposed rule would add
§ 53.1318, which would address the
finality of ESPs, standard design
approvals, and standard DCs referenced
in the CP application.
§ 53.1324 Referral to the Advisory
Committee on Reactor Safeguards
§ 53.1327 Authorization To Conduct
Limited Work Authorization Activities
§ 53.1330 Exemptions, Departures,
and Variances
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This proposed rule would add
§ 53.1330, which would govern requests
for and issuance of exemptions from the
Commission’s regulations and
exemptions, departures, and variances
from NRC approvals, permits, and
certifications.
This proposed rule would add
§ 53.1333, which would establish the
conditions under which the
Commission may issue CPs and
accompanying terms and conditions
under part 53.
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§ 53.1345 Transfer of Construction
Permits
This proposed rule would add
§ 53.1345, which would govern the
transfer of CPs under part 53.
§ 53.1348 Termination of Construction
Permits
This proposed rule would add
§ 53.1348, which would require the
holder of a permit under part 53 to
provide written certification to the
Commission within 30 days of
determining to permanently cease
construction.
§ 53.1360
Operating Licenses
This proposed rule would add
§ 53.1360, which would provide an
overview of the requirements and
procedures for applicants to apply for
and the Commission to issue an OL
under part 53.
This proposed rule would add
§ 53.1366, which would require an
application for an OL under part 53 to
include the information required by
§ 53.1109 as well as financial
information.
§ 53.1369 Contents of Applications for
Operating Licenses; Technical
Information
This proposed rule would add
§ 53.1327, which would govern
authorization to conduct LWA
activities.
Issuance of Construction
§ 53.1342 Duration of Construction
Permits
This proposed rule would add
§ 53.1342, which would establish
requirements for the expiration of a CP.
§ 53.1366 Contents of Applications for
Operating Licenses; General Information
This proposed rule would add
§ 53.1324, which would require the
ACRS to report on any portions of the
application for a CP under part 53
concerning safety.
§ 53.1333
Permits
§ 53.1336 Finality of Construction
Permits
This proposed rule would add
§ 53.1336, which would address the
finality of CPs.
This proposed rule would add
§ 53.1369, which would require an
application for an OL under part 53 to
include certain technical information in
an FSAR at a level of detail sufficient for
the Commission to reach a final
conclusion on all safety matters.
§ 53.1372 Contents of Applications for
Operating Licenses; Other Application
Content
This proposed rule would add
§ 53.1372, which would require an
application for an OL under part 53 to
include an environmental report and a
description of availability controls.
§ 53.1375
Review of Applications
This proposed rule would add
§ 53.1375, which would establish the
standards and procedures for reviewing
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applications and holding hearings on
OLs under part 53.
§ 53.1381 Referral to the Advisory
Committee on Reactor Safeguards
This proposed rule would add
§ 53.1381, which would require the
ACRS to report on any portions of the
application for a CP under part 53
concerning safety.
§ 53.1384 Exemptions, Departures,
and Variances
This proposed rule would add
§ 53.1384, which would govern requests
for and the issuance of exemptions from
the Commission’s regulations and
exemptions, departures, and variances
from NRC approvals, permits, and
certifications.
§ 53.1387
Licenses
Issuance of Operating
This proposed rule would add
§ 53.1387, which would establish the
conditions under which the
Commission may issue OLs and
accompanying conditions and
limitations, including TS, under part 53.
§ 53.1390
Licenses
Backfitting of Operating
This proposed rule would add
§ 53.1390, which would prevent the
Commission from modifying, adding, or
deleting any terms or conditions of the
OL, except in accordance with
§ 53.1590.
§ 53.1396
Licenses
Duration of Operating
This proposed rule would add
§ 53.1396, which would provide that an
OL under part 53 may be valid for up
to 40 years.
§ 53.1399
License
Transfer of an Operating
This proposed rule would add
§ 53.1399, which would provide that an
OL under part 53 may be transferred
under § 53.1570.
§ 53.1402
Application for Renewal
This proposed rule would add
§ 53.1402, which would provide that an
application for a renewed OL under part
53 must be filed in accordance with
§ 53.1595.
§ 53.1405 Continuation of an
Operating License
This proposed rule would add
§ 53.1405, which would govern the
continuing obligations of the holder of
an OL under part 53 following the
permanent cessation of operations.
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§ 53.1434 Authorization To Conduct
Limited Work Authorization Activities
Combined Licenses
This proposed rule would add
§ 53.1410, which would provide an
overview of the requirements and
procedures for applicants to apply for
and the Commission to issue a COL
under part 53.
This proposed rule would add
§ 53.1434, which would address
authorization to conduct LWA
activities.
§ 53.1413 Contents of Applications for
Combined Licenses; General
Information
This proposed rule would add
§ 53.1413, which would require an
application for a COL under part 53 to
include the information required by
§ 53.1109 as well as financial
information.
§ 53.1416 Contents of Applications for
Combined Licenses; Technical
Information
This proposed rule would add
§ 53.1416, which would require
applicants for a COL under part 53 to
submit an FSAR with a level of
technical information sufficient to reach
a final conclusion on all safety matters.
§ 53.1419 Contents of Applications for
Combined Licenses; Other Application
Content
This proposed rule would add
§ 53.1419, which would require
applicants for a COL under part 53 to
submit an environmental report, a
description of availability controls, the
ITAAC that the licensee must perform.
It would also include ITAAC
requirements for applications that
reference an ESP, standard DC, ML, or
combination thereof.
§ 53.1422
Review of Applications
This proposed rule would add
§ 53.1422, which would require
applications for COLs under part 53 to
be reviewed for compliance with
applicable standards and establish
procedural requirements for reviewing
applications and holding hearings in
accordance with part 2.
§ 53.1425 Finality of Referenced NRC
Approvals
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This proposed rule would add
§ 53.1425 which would address the
finality of ESPs, standard DC rules,
standard design approvals, or MLs
referenced in the application for a COL
under part 53.
§ 53.1431 Referral to the Advisory
Committee on Reactor Safeguards
This proposed rule would add
§ 53.1431, which would require the
ACRS to report on any portions of the
application for a COL under part 53
concerning safety.
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§ 53.1437 Exemptions, Departures,
and Variances
This proposed rule would add
§ 53.1437, which would govern the
conditions in which the Commission
may grant an exemption for one or more
of its regulations, or an exemption,
variance, or departure from a permit,
design approval, or license.
§ 53.1440
Licenses
Issuance of Combined
This proposed rule would add
§ 53.1440, which would establish the
conditions under which the
Commission may issue COLs and
accompanying conditions and
limitations, including TS, under part 53.
§ 53.1443
Licenses
Finality of Combined
This proposed rule would add
§ 53.1443, which would govern
permissible modifications or
amendments that the Commission may
make to a COL, as well as permissible
changes that a licensee may make to
facilities and procedures as described in
the FSAR.
§ 53.1449 Inspection During
Construction
This proposed rule would add
§ 53.1449, which would establish
requirements related to inspections,
tests, or analyses for the holder of a COL
under part 53.
§ 53.1452 Operation Under a
Combined License
This proposed rule would add
§ 53.1452, which would establish
requirements describing the
notifications, hearings, and findings to
be made prior to commencing facility
operations.
§ 53.1455
License
Duration of a Combined
This proposed rule would add
§ 53.1455, which would govern the
duration of a COL under part 53.
§ 53.1456
License
Transfer of a Combined
This proposed rule would add
§ 53.1456, which would permit the
transfer of a COL under part 53 in
accordance with § 53.1570.
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§ 53.1458 Application for Renewal
This proposed rule would add
§ 53.1458, which would provide that an
application for renewal of a COL must
be filed in accordance with § 53.1595.
§ 53.1461 Continuation of Combined
License
This proposed rule would add
§ 53.1461, which would govern the
continuing obligations of the holder of
a COL under part 53 following the
permanent cessation of operations.
§ 53.1470 Standardization of
Commercial Nuclear Plant Designs:
Licenses To Construct and Operate
Nuclear Power Reactors of Identical
Design at Multiple Sites
This proposed rule would add
§ 53.1470, which would govern the
requirements and procedures for filing
and issuing applications for a CP, OL, or
COL under part 53 in which the
applicant seeks approval of the same
design for multiple sites.
Subpart I—Maintaining and Revising
Licensing-Basis Information
This proposed rule would add subpart
I, which would address the maintenance
of licensing-basis information for part
53.
§ 53.1500 Licensing-Basis Information
This proposed rule would add
§ 53.1500, describing the purpose of
subpart I, which would be to provide
the requirements for the maintenance of
licensing-basis information for
commercial nuclear plants licensed
under part 53.
§ 53.1502 Specific Terms and
Conditions of Licenses
This proposed rule would add
§ 53.1502, which would outline the
specific terms and conditions for
obtaining a license under part 53.
§ 53.1505 Changes to Licensing-Basis
Information Requiring Prior NRC
Approval
This proposed rule would add
§ 53.1505, which would provide an
overview of the process for licensees to
request, and the Commission to issue,
amendments to licensing-basis
information under part 53.
§ 53.1510 Application for Amendment
of License
This proposed rule would add
§ 53.1510, which would require
licensees under part 53 to file an
application to request an amendment to
the license. Applicants must assess how
their requested changes would impact
the safety criteria and analysis
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§ 53.1545 Updating Final Safety
Analysis Reports
requirements in subpart B and C, as
applicable, whether the amendment
involves no significant hazards
consideration using the standards in
§ 53.1520 and consider potential
impacts on environmental factors.
This proposed rule would add
§ 53.1545, which would require
licensees under part 53 to regularly
update FSARs in accordance with the
requirements of this section to reflect
changes to licensing-basis information.
§ 53.1515 Public Notices; State
Consultation
This proposed rule would add
§ 53.1515, which would outline the
Commission’s procedures for issuing a
notification in the Federal Register and
consulting with the State in which the
commercial nuclear facility is located in
connection with its consideration of
applications for an amendment to an OL
or COL under part 53.
§ 53.1520
Issuance of Amendment
This proposed rule would add
§ 53.1520, which would outline criteria
for the Commission to consider in
issuing license amendments under part
53.
§ 53.1525 Revising Certification
Information Within a Design
Certification Rule
This proposed rule would add
§ 53.1525, which would address the
requirements for applicants to request,
and the Commission to grant, an
exemption to a DC rule under part 53.
§ 53.1530 Revising Design Information
Within a Manufacturing License
This proposed rule would add
§ 53.1530, which would require the
holder of an ML to request an
amendment under § 53.1510 and, as
applicable, § 53.1520 to make changes to
the design of a manufactured reactor. It
would also outline the requirements for
holders of a COL under part 53 to
request amendments for changes to the
design information of a manufactured
reactor.
§ 53.1535 Amendments During
Construction
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This proposed rule would add
§ 53.1535, which would outline the
process for licensees under part 53 to
request amendments to CPs or LWAs
during construction.
This proposed rule would add
§ 53.1540, which would provide an
overview of the regulations in subpart I
for holders of an OL or COL under part
53 to modify licensing-basis information
and definitions relevant to §§ 53.1545
through 53.1565.
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This proposed rule would add
§ 53.1550, which would require
licensees under part 53 to follow the
guidelines outlined in this section in
determining whether changes to
licensing-basis information described in
the FSAR (as updated) require them to
obtain a license amendment.
§ 53.1560 Updating Program
Documents Included in Licensing-Basis
Information
This proposed rule would add
§ 53.1560, which would require the
holders of an OL or COL under part 53
to regularly update the program
documents that they submitted in their
application for a license.
§ 53.1565 Evaluating Changes to
Programs Included in Licensing-Basis
Information
This proposed rule would add
§ 53.1565, which would enable
licensees under part 53 to make changes
to the facility, procedures, or
organization, or address changes to site
environs as described in program
documents without NRC approval if
these changes satisfy the criteria
outlined in this section.
§ 53.1570
Transfer of Licenses
This proposed rule would add
§ 53.1570, which would outline the
requirements for an application for
transfer of a license issued under part
53.
§ 53.1575
§ 53.1540 Updating Licensing-Basis
Information and Determining the Need
for NRC Approval
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§ 53.1550 Evaluating Changes to
Facility as Described in Final Safety
Analysis Reports
Termination of Licenses
This proposed rule would add
§ 53.1575, which would outline the
process for terminating an OL or COL
issued under part 53.
§ 53.1580
Information Requests
This proposed rule would add
§ 53.1580, which would address the
process and circumstances under which
the NRC may send information requests
to the various types of licensees within
part 53.
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§ 53.1585 Revocation, Suspension,
Modification of Licenses and Approvals
for Cause
This proposed rule would add
§ 53.1585, which would address
grounds for the revocation, suspension,
or modification of a license or standard
design approval issued under part 53.
§ 53.1590 Backfitting
This proposed rule would add
§ 53.1590, which would define
backfitting and establish requirements
to be met by the NRC when it takes
backfitting actions under part 53.
§ 53.1595 Renewal
This proposed rule would add
§ 53.1595, which would provide for the
renewal of a license under part 53 upon
expiration.
Subpart J—Reporting and Other
Administrative Requirements
This proposed rule would add subpart
J, to establish various reporting and
other administrative requirements for
licensees under part 53.
§ 53.1600 General Information
This proposed rule would add
§ 53.1600, which provides an overview
of the sections that would require
applicants and licensees under part 53
to provide NRC inspectors with
unfettered access to sites and facilities,
maintain records and make reports,
demonstrate compliance with financial
qualification and reporting
requirements, and maintain required
financial protection for accidents.
§ 53.1610 Unfettered Access for
Inspections
This proposed rule would add
§ 53.1610, which would require
applicants and licensees under part 53
to provide unfettered access to NRC
inspectors, including access to records,
premises, activities, and licensed
materials, in addition to providing office
space to accommodate temporary or
resident inspectors.
§ 53.1620 Maintenance of Records,
Making of Reports
This proposed rule would add
§ 53.1620, which would require part 53
licensees to maintain all records and
make reports as required by the
conditions of the license or by the
regulations in part 53.
§ 53.1630 Immediate Notification
Requirements for Operating Commercial
Nuclear Plants
This proposed rule would add
§ 53.1630, which would impose
immediate notification requirements on
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part 53 licensees following the
declaration of an Emergency Class or the
discovery of certain non-emergency
events.
financial qualifications information
outlined in this section within seventyfive days of ceasing to be an electric
utility.
(k) to include the appropriate references
to part 53.
§ 53.1640
System
§ 53.1700
This proposed rule would revise
§ 70.24(d) to include the appropriate
references to part 53.
Licensee Event Report
This proposed rule would add
§ 53.1640, which would require any
commercial plant licensee holding an
OL under part 53 to submit a Licensee
Event Report in accordance with the
specifications outlined in this section.
§ 53.1645 Reports of Radiation
Exposure to Members of the Public
The proposed rule would add
§ 53.1645, which would require annual
reports to the Commission, including
radiological reports as required by part
20, an Annual Radioactive Effluent
Release Report, and an Annual
Environmental Operating Report.
§ 53.1650 Facility Information and
Verification
The proposed rule would add
§ 53.1650, which would include a
reporting requirement for applicants
and holders of a CP or license under
part 53 to support safeguards
agreements between the United States
and the IAEA.
§ 53.1660
Financial Requirements
This proposed rule would add
§ 53.1660, which would introduce
requirements and procedures related to
financial qualifications and reporting
requirements for applicants, licensees,
and CP holders under part 53.
§ 53.1670
Financial Qualifications
This proposed rule would add
§ 53.1670, which would require an
applicant for a CP, OL, or COL under
part 53 to must demonstrate possession
or ability to obtain funds necessary for
the activities for which the permit or
license is sought.
§ 53.1680
Annual Financial Reports
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This proposed rule would add
§ 53.1680, which would require
licensees and holders of a CP under part
53 to submit annual financial reports to
the Commission, with exceptions for
those that submit financial forms to the
Securities and Exchange Commission or
the Federal Energy Regulatory
Commission.
§ 53.1690 Licensee’s Change of Status;
Financial Qualifications
This proposed rule would add
§ 53.1690, which would require electric
utility licensees that hold an OL or COL
for a commercial nuclear plant under
part 53 to provide the NRC with the
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Creditor Regulations
This proposed rule would add
§ 53.1700, which would establish
regulations with respect to the creditors
of any facility under part 53.
§ 53.1710
Financial Protection
This proposed rule would add
§ 53.1710, which would establish
requirements for licenses under part 53
to obtain and maintain insurance to
cover the costs of an accident.
§ 53.1720 Insurance Required To
Stabilize and Decontaminate Plant
Following an Accident
§ 70.24 Criticality Accident
Requirements
§ 70.32
Conditions of Licenses
This proposed rule would revise
§ 70.32(c)(1) and (d) to incorporate the
appropriate references to part 53.
§ 70.50
Reporting Requirements
This proposed rule would revise
§ 70.50(d) to clarify the applicability of
the reporting requirements of this
section to part 53 licensees.
§ 72.3
Definitions
This proposed rule would add
§ 53.1720, which would require
commercial nuclear plant licensees
under part 53 to obtain insurance
sufficient to cover the costs of
stabilizing and decontaminating the
plant in the event of an accident.
This proposed rule would revise the
definition of ‘‘Independent spent fuel
storage installation or ISFSI’’ in § 72.3 to
include a reference to facilities licensed
under part 53.
§ 53.1730 Financial Protection
Requirements
This proposed rule would revise
§ 72.30(e)(5) to include the appropriate
references to part 53.
This proposed rule would add
§ 53.1730, which would require
commercial nuclear plant licensees
under part 53 to satisfy the provisions
of part 140.
Subpart M—Enforcement
This proposed rule would add subpart
M, which would address certain
violations and penalties associated with
violations of part 53 regulations.
§ 53.9000
Violations
This proposed rule would add
§ 53.9000, providing notice of the
Commission’s authority to obtain
injunctions or other court orders for the
violations enumerated in this section.
§ 53.9010
Criminal Penalties
This proposed rule would add
§ 53.9010, providing notice to all
persons and entities subject to part 53
that they are subject to criminal
sanctions for willful violations,
attempted violations, or conspiracy to
violate certain regulations under part
53.
§ 70.20a General License to Possess
Special Nuclear Material for Transport
This proposed rule would revise
§ 70.20a(b) to include a reference to part
53.
§ 70.22
Contents of Applications
This proposed rule would revise
§ 70.22, paragraphs (b), (h)(1), (j)(1), and
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§ 72.30 Financial Assurance and
Recordkeeping for Decommissioning
§ 72.32
Emergency Plan
This proposed rule would revise
§ 72.32(c)(2) to include a reference to
the exclusion area as defined in part 53.
§ 72.40
Issuance of License
This proposed rule would revise
§ 72.40(c) regarding the issuance of a
license under part 72 to include a
reference to previous licensing actions,
including the issuance of a CP under
part 53.
§ 72.75 Reporting Requirements for
Specific Events and Conditions
This proposed rule would revise
§ 72.75(i)(1)(ii) regarding reporting
requirements for specific events and
conditions with references to reactors
licensed under part 53.
§ 72.184
Safeguards Contingency Plan
This proposed rule would revise
§ 72.184(a) regarding the requirements
of a licensee’s safeguarding contingency
plan with a reference to nuclear
facilities licensed under part 53.
§ 72.210
General License Issued
This proposed rule would revise
§ 72.210 to issue a general license for
the storage of spent fuel in an
independent spent storage installation
at power to persons authorized to
possess or operate nuclear power
reactors under part 53.
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§ 72.212 Conditions of General License
Issued Under § 72.210
This proposed rule would revise
§ 72.212(b)(8) regarding the conditions
of a general license issued under
§ 72.210 to include a reference to license
amendments for a facility made
pursuant to part 53.
§ 72.218 Termination of Licenses
This proposed rule would revise
§ 72.218(a) to include a reference to the
notification required under part 53
regarding the plan for managing spent
fuel prior to decommissioning. It would
also extend the provisions of § 72.218(b)
to a reactor operating or COL under part
53.
§ 73.1 Purpose and Scope
This proposed rule would revise
§ 73.1(b)(1)(i) to extend the scope of part
73 to production and utilization
facilities licensed under part 53, in
addition to parts 50 and 52.
§ 73.57 Requirements for Criminal
History Records Checks of Individuals
Granted Unescorted Access to a Nuclear
Power Facility, a Non-power Reactor, or
Access to Safeguards Information
This proposed rule would revise
§ 73.57(a)(3) to incorporate the
appropriate references to OLs granted
under part 53 and Commission findings
under § 53.1452(g) regarding the
requirement for license applicants to
submit fingerprints for all personnel
with unescorted access.
§ 73.58 Safety/Security Interface
Requirements for Nuclear Power
Reactors
This proposed rule would revise
§ 73.58(a) to extend the requirements of
this section to part 53 licensees.
§ 73.2 Definitions
This proposed rule would revise
§ 73.2 introductory text and paragraph
(a) such that terms defined in part 53
have the same meaning in part 73.
§ 73.67 Licensee Fixed Site and InTransit Requirements for the Physical
Protection of Special Nuclear Material
of Moderate and Low Strategic
Significance
§ 73.8 Information Collection
Requirements: OMB Approval
This proposed rule would revise
§ 73.8(b) with the new information
collection requirements contained in
proposed §§ 73.77, 73.100, 73.110, and
73.120.
This proposed rule would revise
§ 73.67(d) and (f) to include a reference
to licensees authorized to operate a
nuclear power plant under part 53.
§ 73.50 Requirements for Physical
Protection of Licensed Activities
This proposed rule would revise
§ 73.50 to exempt nuclear reactor
facilities licensed under part 53, in
addition to parts 50 and 52, from the
requirements of this section.
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under part 53 who do not demonstrate
compliance with certain requirements
under part 53.
§ 73.56 Personnel Access
Authorization Requirements for Nuclear
Power Plants
This proposed rule would revise
§ 73.56(a)(3) to apply this section’s
personnel AA requirements to
applicants for an OL or holders of a COL
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This proposed rule would revise
§ 73.77, paragraphs (a), (b), (c)(6) and (7)
regarding the notification process for
cybersecurity events to include
notifications for the declaration of an
emergency class made under part 53.
Subpart J—Security Requirements at
Commercial Nuclear Plants
§ 73.55 Requirements for Physical
Protection of Licensed Activities in
Nuclear Power Reactors Against
Radiological Sabotage
This proposed rule would revise
§ 73.55, paragraphs (a)(4) and (6),
(i)(4)(iii), (l)(1), (l)(7)(ii), (p)(1)(i), (r)(2),
and (r)(4)(iii), to incorporate the
appropriate references to part 53
regarding requirements for physical
protection of licensed activities in
nuclear power reactors against
radiological sabotage.
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§ 73.77 Cybersecurity Event
Notifications
This proposed rule would add new
Subpart J of part 73 containing
§§ 73.100, 73.110, and 73.120, to
establish security requirements for
commercial nuclear plants licensed
under part 53.
§ 73.100 Technology-Inclusive
Requirements for Physical Protection of
Licensed Activities at Commercial
Nuclear Plants Against Radiological
Sabotage
This proposed rule would add
§ 73.100, which would establish a
performance-based regulatory
framework for physical protection as an
alternative to the prescriptive
requirements of § 73.55, which also
governs physical protection programs
for part 50 and 52 licensees.
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§ 73.110 Technology-Inclusive
Requirements for Protection of Digital
Computer and Communication Systems
and Networks
This proposed rule would add
§ 73.110, which would establish a
consequence-based approach to
cybersecurity and would require that
part 53 licensees demonstrate
reasonable assurance that digital
computer and communication systems
and networks are adequately protected
against cyberattacks in a manner that is
commensurate with the potential
consequences of those attacks.
§ 73.120 Access Authorization
Program for Commercial Nuclear Plants
This proposed rule would add
§ 73.120, which would establish
performance objectives as an alternative
to compliance with the AA provisions
of §§ 73.55, 73.56, and 73.57. This
proposed rule would afford part 53
licensees additional flexibility in
establishing an AA program that
demonstrates compliance with the
performance objectives and
requirements of this section.
§ 73.1200 Notification of Physical
Security Events
This proposed rule would revise
§ 73.1200, paragraphs (a), (c)(1), (e)(1),
(e)(3), (e)(4), (g)(1), (o)(5)(i), (o)(6)(i), (r),
and (s) to extend the requirements of
this section to part 53 licensees.
§ 73.1205 Written Follow-Up Reports
of Physical Security Events
This proposed rule would revise
§ 73.1205(b)(2) to extend the
requirements of this section to part 53
licensees.
§ 73.1210 Recordkeeping of Physical
Security Events
This proposed rule would revise
§ 73.1210(a)(1) and (b)(3)(i) to extend
the requirements of this section to part
53 licensees.
§ 73.1215 Suspicious Activity Reports
This proposed rule would revise
§ 73.1215(d)(1) to include a reference to
§ 73.100.
Appendix B to part 73—General Criteria
for Security Personnel
This proposed rule would revise
appendix B to part 73 to state that terms
defined in part 53 have the same
meaning when used in this appendix.
§ 74.31 Nuclear Material Control and
Accounting for Special Nuclear Material
of Low Strategic Significance
This proposed rule would revise
§ 74.31(a) to include a reference to
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§ 140.12 Amount of Financial
Protection Required for Other Reactors
production or utilization facilities
licensed under part 53, in addition to
parts 50 and 70.
§ 74.41 Nuclear Material Control and
Accounting for Special Nuclear Material
of Moderate Strategic Significance
This proposed rule would revise
§ 74.41(a) to include a reference to
nuclear reactors licensed under part 53.
§ 74.51 Nuclear Material Control and
Accounting for Strategic Special
Nuclear Material
This proposed rule would revise
§ 74.51(a) to include a reference to
nuclear reactors licensed under part 53.
§ 75.4
Definitions
This proposed rule would revise
§ 75.4 such that terms defined in
§ 53.020 have the same meaning when
used in this part. The definition of
‘‘Facility’’ would also be revised to
include any plant or location where
more than 1 effective kilogram of
nuclear material is licensed pursuant to
part 53.
§ 95.5
Definitions
This proposed rule would revise the
definition of ‘‘License’’ in § 95.5 to
include those issued under part 53.
§ 95.39 External Transmission of
Documents and Material
This proposed rule would revise
§ 95.39(a) to apply restrictions to the
external transmission of documents and
material containing classified
information in connection with NRC
licenses, certificates, standard design
approvals, or standard DCs issued under
part 53.
§ 140.2
Scope
This proposed rule would revise
§ 140.2(a)(1) and (2) to include part 53
applicants and licensees within the
scope of part 140 regulations.
§ 140.10
Scope
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This proposed rule would revise
§ 140.10 to apply the provisions of
subpart B to applicants or holders of a
license to operate a nuclear reactor
under part 53, as well as applicants and
holders of a COL under part 53.
This proposed rule would revise
§ 140.12(c) to require the licensee’s
primary financial protection to cover all
reactors in any case where a person is
authorized under part 53 to operate two
or more nuclear reactors at the same
location.
§ 140.13 Amount of Financial
Protection Required of Certain Holders
of Construction Permits and Combined
Licenses Under 10 CFR Part 52
This proposed rule would revise
§ 140.13 with the appropriate references
to part 53 regarding the requirement for
holders of a CP or COL under part 53
to obtain financial protection.
§ 140.20
Liens
Indemnity Agreements and
This proposed rule would revise
§ 140.20(a)(1)(i) and (ii) with
appropriate references to part 53.
§ 150.15
Persons Not Exempt
The proposed rule would revise
§ 150.15, paragraphs (a)(7)(iii) and (a)(8)
to add a reference to facilities licensed
under parts 53 and 52.
§ 170.3
Definitions
The proposed rule would revise
§ 170.3 to incorporate references to part
53 into the definitions of
‘‘Manufacturing license,’’ ‘‘Part 55
Reviews,’’ ‘‘Power reactor,’’ and
‘‘Special projects.’’
§ 170.12
Payment of Fees
The proposed rule would revise
§ 170.12(d)(1)(v) regarding special
project fees in connection with FSARs
to include part 53.
§ 170.21 Schedule of Fees for
Production and Utilization Facilities,
Review of Standard Referenced Design
Approvals, Special Projects,
Inspections, And import and Export
Licenses
The proposed rule would revise
§ 170.21, footnote 1 to include fees
charged for approvals issued under the
exemption provision in § 53.080.
§ 140.11 Amounts of Financial
Protection for Certain Reactors
§ 170.41 Failure by Applicant or
Licensee to Pay Prescribed Fees
This proposed rule would revise
§ 140.11(b) to require the licensee’s
primary financial protection to cover all
reactors in any case where a person is
authorized under part 53 to operate two
or more nuclear reactors at the same
location.
The proposed rule would revise
§ 170.41 to include a general reference
to part 53 in connection with remedial
actions that the Commission might take
when an applicant or licensee fails to
pay a prescribed fee required by this
part.
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§ 171.3
Scope
The proposed rule would revise
§ 171.3 to apply the provisions of this
part to any person holding an OL for a
power reactor licensed under part 53 or
a COL issued under part 53.
§ 171.5
Definitions
This proposed rule would revise the
definitions of ‘‘Operating license’’ and
‘‘Power reactor’’ in § 171.5 to
incorporate the appropriate references
to part 53.
§ 171.15 Annual fees: Non-Power
Production or Utilization Licenses,
Reactor Licenses, and Independent
Spent Fuel Storage Licenses
This proposed rule would revise
§ 171.15, paragraphs (a), (b)(2)(iii),
(c)(1), and (d)(1) regarding annual fees
that are applicable to part 53 licensees.
§ 171.17
Proration
This proposed rule would revise
§ 171.17, paragraphs (a), (a)(1)(ii) and
(a)(2) with references to part 53 licenses.
VIII. Regulatory Flexibility
Certification
The Regulatory Flexibility Act of
1980, as amended at 5 U.S.C. 601 et seq,
requires that agencies consider the
impact of their rulemakings on small
entities and, consistent with applicable
statutes, consider alternatives to
minimize these impacts on the
businesses, organizations, and
government jurisdictions to which they
apply.
In accordance with the Small
Business Administration’s (SBA’s)
regulation at 13 CFR 121.903(c), the
NRC has developed its own size
standards for performing an RFA
analysis and has verified with the SBA
Office of Advocacy that its size
standards are appropriate for NRC
analyses. The NRC size standards at
§ 2.810, ‘‘NRC size standards,’’ are used
to determine whether an applicant or
licensee qualifies as a small entity in the
NRC’s regulatory programs. Section
2.810 defines the following types of
small entities:
Small business is a for-profit concern
and is a—(1) Concern that provides a
service or a concern not engaged in
manufacturing with average gross
receipts of $8.0 million or less over its
last 5 completed fiscal years; or (2)
Manufacturing concern with an average
number of 500 or fewer employees
based upon employment during each
pay period for the preceding 12 calendar
months.
Small organization is a not-for-profit
organization which is independently
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owned and operated and has annual
gross receipts of $8.0 million or less.
Small governmental jurisdiction is a
government of a city, county, town,
township, village, school district, or
special district with a population of less
than 50,000.
Small educational institution is one
that is—(1) Supported by a qualifying
small governmental jurisdiction; or (2)
Not State or publicly supported and has
500 or fewer employees.
Number of Small Entities Affected
The NRC is currently not aware of any
known small entities as defined in
§ 2.810 that are planning to apply for a
commercial nuclear plant ESP, CP, OL,
ML, or COL under part 53 that would be
impacted by this proposed rule. Based
on this finding, the NRC has
preliminarily determined that the
proposed rule would not have a
significant economic impact on a
substantial number of small entities.
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Economic Impact on Small Entities
Depending on how the ownership
and/or operating responsibilities for
such an enterprise were structured,
applicants for a commercial nuclear
plant rated 8 Megawatts electric (MWe)
or less could conceivably qualify as
small entities as defined by § 2.810.
Owners that operate power reactors
rated greater than 8 MWe could generate
sufficient electricity revenue that
exceeds the gross annual receipts limit
of $8 million, assuming a 90 percent
capacity factor and the June 2021 DOE’s
Energy Information Administration U.S.
average price of electricity to the
ultimate customer for all sectors of 11.3
cents per kilowatt-hour.
Although the NRC is not aware of any
small entities that would be affected by
the proposed rule, there is a possibility
that future applications for a
commercial nuclear plant permit or
license could be submitted by small
entities who plan to own and operate a
commercial nuclear plant rated 8 MWe
or less. Commercial nuclear plants that
are rated 8 MWe or less would most
likely be used to support electrical
demand for military bases or small
remote towns and would provide
process heat, so they would not directly
compete with a larger commercial
nuclear plant that would typically
produce electricity for the grid. As a
result of these differing purposes, the
NRC would expect that small and large
entities would not be in direct
competition with each other.
Therefore, the NRC preliminarily
concludes that this proposed rule would
not have a significant economic impact
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on a substantial number of small
entities.
Request for Comments
The NRC is seeking comment on both
its initial RFA analysis and on its
preliminary conclusion that this
proposed rule would not have a
significant economic impact on a
substantial number of small entities
because of the likelihood that most
expected applicants would not qualify
as a small entity. Additionally, the NRC
is seeking comment on its preliminary
conclusion that if a small entity were to
submit a commercial nuclear plant
application, the small entity would not
incur a significant economic impact as
it would most likely not be in
competition with a large entity.
Any small entity that could be subject
to this regulation that determines,
because of its size, it is likely to bear a
disproportionate adverse economic
impact should notify the Commission of
this opinion in a comment that
indicates—
1. The applicant’s size and how the
proposed regulation would impose a
significant economic burden on the
applicant as compared to the economic
burden on a larger applicant;
2. How the proposed regulations
could be modified to take into account
the applicant’s differing needs or
capabilities;
3. The benefits that would accrue or
the detriments that would be avoided if
the proposed regulations were modified
as suggested by the applicant;
4. How the proposed regulation, as
modified, would more closely equalize
the impact of NRC regulations or create
more equal access to the benefits of
Federal programs as opposed to
providing special advantages to any
individual or group; and
5. How the proposed regulation, as
modified, would still adequately
demonstrate compliance with the NRC’s
obligations under the Act.
IX. Regulatory Analysis
The NRC has prepared a draft
regulatory analysis for this proposed
rule. The analysis examines the costs
and benefits of the alternatives
considered by the NRC. The conclusion
from the analysis is that this proposed
rule and associated guidance would
result in net averted costs to the
industry and the NRC of $28.1 million
using a 7-percent discount rate and
$34.5 million using a 3-percent discount
rate due to reductions in exemption
requests. The analysis also assumes one
applicant under part 53. As the number
of applicants increases, so do the
estimated averted costs. The NRC
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requests public comment on the draft
regulatory analysis, which is available
as indicated in the ‘‘Availability of
Documents’’ section of this document.
Comments on the draft regulatory
analysis may be submitted to the NRC
as indicated under the ADDRESSES
caption of this document.
X. Backfitting and Issue Finality
This section describes the backfitting
and issue finality implications of this
proposed rule and the draft guidance
documents described in section XVIII,
‘‘Availability of Guidance,’’ in this
document, as applied to pertinent NRC
approvals and certain applicants that
reference NRC approvals in their
applications. The NRC’s current
backfitting provisions associated with
nuclear power plants appear in § 50.109,
‘‘Backfitting,’’ and apply to CPs and OLs
under part 50. Issue finality provisions
(analogous to the backfitting provisions
in § 50.109) for approvals under part 52
are located in various provisions of part
52. The NRC Management Directive 8.4,
‘‘Management of Backfitting, Forward
Fitting, Issue Finality, and Information
Requests,’’ describes the Commission’s
policies on backfitting and issue
finality.
This proposed rule would provide a
regulatory scheme for entities to apply
for approvals under part 53. The part 50
backfitting provisions and part 52 issue
finality provisions apply to actions
taken by the NRC under part 50 or part
52, respectively, or actions taken by the
NRC under other parts of 10 CFR
chapter I that, for holders of certain
approvals under part 50 or part 52,
inextricably affect their activities
regulated under part 50 or part 52.
Issuance and implementation of
proposed part 53 would not constitute
actions taken under part 50 or part 52.
Also, proposed part 53 would not allow
an applicant to reference approvals
issued under part 50 or part 52.
Therefore, the issuance and
implementation of proposed part 53
would not affect part 50 or part 52
entities’ activities regulated under part
50 or part 52. Therefore, the addition of
part 53 through this proposed rule
would not be within the scope of the
part 50 backfitting and part 52 issue
finality provisions.
The NRC also proposes conforming
changes to parts 1, 2, 10, 11, 19, 20, 21,
25, 26, 30, 40, 50, 51, 70, 72, 73, 74, 75,
95, 140, 150, 170, and 171 to reflect the
addition of part 53. These changes
would not meet the definition of
‘‘backfitting’’ in § 50.109 or § 70.76,
‘‘Backfitting,’’ because the proposed
changes would not modify or add to the
systems, structures, components, or
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design of a facility or to the procedures
or organization required to operate a
facility under part 50 or 70. These
changes would not meet the definition
of ‘‘backfitting’’ in § 72.62,
‘‘Backfitting,’’ because the proposed
changes would not add, eliminate, or
modify the SSCs of an independent
spent fuel storage installation (ISFSI) or
the procedures or organization required
to operate an ISFSI. These proposed
changes would not inextricably affect
activities regulated under parts 50, 52,
70, or 72. Therefore, the proposed
changes to parts 1, 2, 10, 11, 19, 20, 21,
25, 26, 30, 40, 50, 51, 70, 72, 73, 74, 75,
95, 140, 150, 170, and 171 would not
constitute backfitting under parts 50, 70,
or 72 or affect the issue finality of an
approval under part 52.
The NRC is issuing 10 draft guidance
documents that, if issued as final
guidance documents, would provide
guidance on the methods acceptable to
the NRC for complying with aspects of
this proposed rule. These documents
would not apply to holders of approvals
issued under part 50 or part 52. Further,
as discussed in the guidance
documents, applicants and licensees
would not be required to comply with
the positions set forth in the guidance.
Therefore, issuance of the guidance
documents as final guidance would not
constitute backfitting under part 50 or
affect the issue finality of any approval
issued under part 52.
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XI. Cumulative Effects of Regulation
The NRC seeks to minimize any
potential negative consequences
resulting from the cumulative effects of
regulation (CER). The CER describes the
challenges that licensees, or other
impacted entities such as State partners,
may face while implementing new
regulatory positions, programs, or
requirements (e.g., rules, generic letters,
backfits, inspections). The CER is an
organizational effectiveness challenge
that may result from a licensee or
impacted entity implementing a number
of complex regulatory actions,
programs, or requirements within
limited available resources. The NRC’s
CER process involved engaging with
external stakeholders throughout this
proposed rule and related regulatory
activities. Public involvement has
included numerous public meetings to
examine the part 53 risk-informed,
technology-inclusive requirements for
commercial nuclear plants and the
publication of numerous versions of
preliminary proposed rule language.
The NRC is considering holding
additional public meetings during the
remainder of the rulemaking process.
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In parallel with this proposed rule,
the NRC is issuing 10 draft
implementing guidance documents for
comment to support informed external
stakeholder feedback. Section XVII,
‘‘Availability of Guidance,’’ of this
document describes how the public can
access the draft implementing guidance.
In addition to the questions in the
‘‘Specific Requests for Comments’’
section of this document, the NRC is
requesting CER feedback on the
following questions:
1. In light of any current or projected
CER challenges, does the proposed
rule’s effective date provide sufficient
time to implement the new proposed
requirements, including changes to
programs, procedures, and the facility?
2. If CER challenges currently exist or
are expected, what should be done to
address them? For example, if more
time is required for implementation of
the new requirements, what period of
time is sufficient?
3. Do other (NRC or other agency)
regulatory actions (e.g., orders, generic
communications, license amendment
requests, inspection findings of a
generic nature) influence the
implementation of the proposed rule’s
requirements?
4. Are there unintended
consequences? Does the proposed rule
create conditions that would be contrary
to the proposed rule’s purpose and
objectives? If so, what are the
unintended consequences, and how
should they be addressed?
5. Please comment on the NRC’s cost
and benefit estimates in the regulatory
analysis that supports this proposed
rule. The draft regulatory analysis is
available as indicated under the
‘‘Availability of Documents’’ section of
this document.
XII. Plain Writing
The Plain Writing Act of 2010 (Pub.
L. 111–274) requires Federal agencies to
write documents in a clear, concise, and
well-organized manner. The NRC has
written this document to be consistent
with the Plain Writing Act as well as the
Presidential Memorandum, ‘‘Plain
Language in Government Writing,’’
published June 10, 1998 (63 FR 31885).
The NRC requests comment on this
document with respect to the clarity and
effectiveness of the language used.
XIII. Environmental Assessment and
Proposed Finding of No Significant
Environmental Impact
The Commission has preliminarily
determined under the National
Environmental Policy Act of 1969, as
amended, and the Commission’s
regulations in subpart A of part 51, that
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this rule, if adopted, would not be a
major Federal action significantly
affecting the quality of the human
environment, and an EIS is not required.
The implementation of the proposed
rule requirements does not have a
significant impact on the environment.
The proposed rulemaking would either
have requirements that are
administrative in application, matters of
procedure, or provide an equivalent
level of safety as existing requirements;
therefore, there would be similar
environmental impacts from the
implementation of the part 53
regulations as there are for existing
requirements.
The preliminary determination of this
EA is that there will be no significant
effect on the quality of the human
environment from this action. Public
stakeholders should note, however, that
comments on any aspect of this EA may
be submitted to the NRC as indicated
under the ADDRESSES section of this
document. The EA is available as
indicated under the ‘‘Availability of
Documents’’ section of this document.
The NRC has sent a copy of the EA,
and this proposed rule to every State
Liaison Officer and has requested
comments.
XIV. Paperwork Reduction Act
This proposed rule contains new
collections of information contained in
parts 26, 50, 53, and 73 and NRC Forms
361S, 366, 366A, 366B, 893, and 894
that are subject to the Paperwork
Reduction Act of 1995 (44 U.S.C. 3501
et seq). The collections of information
have been submitted to the OMB for
review and approval. The proposed
changes to parts 2, 10, 11, 19, 20, 21, 25,
30, 40, 51, 70, 72, 74, 75, 95, 140, 150,
170, and 171 do not contain any new or
amended collections of information
subject to the Paperwork Reduction Act
of 1995. Existing collections of
information were approved by the OMB,
approval numbers 3150–0062 (part 11),
3150–0044 (part 19), 3150–0014 (part
20), 3150–0035 (part 21), 3150–0046
(part 25), 3150–0017 (part 30), 3150–
0020 (part 40), 3150–0021 (part 51),
3150–0024 (NRC Form 396), 3150–0090
(NRC Form 398), 3150–0009 (part 70),
3150–0132 (part 72), 3150–0123 (part
74), 3150–0055 (part 75), 3150–0047
(part 95), 3150–0039 (part 140), and
3150–0032 (part 150).
Type of submission, new or revision:
Revision and new.
The title of the information collection:
Risk-Informed, Technology-Inclusive
Regulatory Framework for Advanced
Reactors.
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The form number if applicable: NRC
Forms 361S, 366, 366A, 366B, 893, and
894.
How often the collection is required or
requested: Once, on occasion, every 30
days, biannually, annually, biennially,
every four years, every five years, every
ten years.
Who will be required or asked to
respond: Part 53 commercial nuclear
plant licensees and license applicants
for commercial nuclear plants to be
licensed under part 53.
An estimate of the number of annual
responses: 15 (2 responses for Part 26,
11 responses for Part 53, 2 responses for
Part 50 and 0 responses for Part 73 and
NRC Forms 361S, 366, 366A, 366B, 893,
and 894)
The estimated number of annual
respondents: 2 (2 respondents for Part
26, 2 respondents for Part 53, 2
respondents for Part 50 and 0
respondents for Part 73 and NRC Forms
361S, 366, 366A, 366B, 893, and 894)
An estimate of the total number of
hours needed annually to comply with
the information collection requirement
or request: 230,244 hours. (656 hours for
Part 26, 220,801 hours for Part 53, 8,767
hours for Part 50 and 0 hours for Part
73 and NRC Forms 361S, 366, 366A,
366B, 893, and 894)
Abstract: The NRC is proposing to
establish an optional technologyinclusive regulatory framework for use
by applicants for new commercial
nuclear plant designs. The regulatory
requirements developed in this
rulemaking would use methods of
evaluation, including risk-informed and
performance-based methods, that are
flexible and practicable for application
to a variety of new reactor technologies.
The NRC’s goals in amending these
regulations are to continue to provide
reasonable assurance of adequate
protection of public health and safety
and the common defense and security at
reactor sites at which new nuclear
reactor designs are deployed to at least
the same degree of protection as
required for current-generation LWRs;
protect health and minimize danger to
life or property to at least the same
degree of protection as required for
current-generation LWRs; provide
greater operational flexibilities where
supported by enhanced margins of
safety that may be provided in new
nuclear designs; and promote regulatory
stability, predictability, and clarity.
The proposed rule covers diverse
topics, which result in recordkeeping
and reporting requirements related to
contents of applications, plant design
and analysis, siting, construction and
manufacturing, licensing-basis
information, facility operations,
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programs, staffing, FFD, physical
security, cyber-security, AA,
decommissioning, and quality
assurance.
In addition to the new information
collections in the proposed regulations,
part 53 would result in new collections
via NRC Forms 361S, 366, 366A, 366B,
893, and 894. NRC Forms 366, 366A,
and 366B would be modified to include
part 53 reportable events covering an
equivalent scope as the requirements in
10 CFR 50.73, but without LWR-specific
terminology to ensure technology
inclusiveness. The proposed rule also
would require part 53 licensees to use
NRC Forms 893 and 894 to report on
positive drug and alcohol test results
(NRC Form 893) and annual fitness-forduty program performance (NRC Form
894). Finally, a new version of NRC
Form 361 (NRC Form 361S) would be
created for use by part 53 licensees,
covering an equivalent scope as the
requirements in 10 CFR 50.72, but
without LWR-specific terminology to
ensure technology inclusiveness.
The NRC is seeking public comment
on the potential impact of the
information collections contained in
this proposed rule and on the following
issues:
1. Is the proposed information
collection necessary for the proper
performance of the functions of the
NRC, including whether the information
will have practical utility? Please
explain your response.
2. Is the estimate of the burden of the
proposed information collection
accurate? Please explain your response.
3. Is there a way to enhance the
quality, utility, and clarity of the
information to be collected? Please
explain your response.
4. How can the burden of the
proposed information collection on
respondents be minimized, including
the use of automated collection
techniques or other forms of information
technology? Please explain your
response.
The OMB clearance documents and
proposed rule is available as indicated
under the ‘‘Availability of Documents’’
section in this document or may be
viewed free of charge by contacting the
NRC’s PDR reference staff at 1–800–
397–4209, at 301–415–4737, or by email
to PDR.resource@nrc.gov. You may
obtain information and comment
submissions related to the OMB
clearance package by searching on
https://www.regulations.gov under
Docket ID NRC–2019–0062.
You may submit comments on any
aspect of these proposed information
collections, including suggestions for
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reducing the burden and on the above
issues, by the following methods:
• Federal Rulemaking Website: Go to
https://www.regulations.gov and search
for Docket ID NRC–2019–0062.
• Mail comments to: FOIA, Library,
and Information Collections Branch,
Office of the Chief Information Officer,
Mail Stop: T6–A10M, U.S. Nuclear
Regulatory Commission, Washington,
DC 20555–0001 or by email to
Infocollects.Resource@nrc.gov or to the
OMB reviewer at: OMB Office of
Information and Regulatory Affairs
(3150–XXXX, 3150–0002, –0104, –0146,
–0238), Attn: Desk Officer for the
Nuclear Regulatory Commission, 725
17th Street NW, Washington, DC 20503.
Submit comments by December 2,
2024. Comments received after this date
will be considered if it is practical to do
so, but the NRC staff is able to ensure
consideration only for comments
received on or before this date.
Public Protection Notification
The NRC may not conduct or sponsor,
and a person is not required to respond
to, a collection of information unless the
document requesting or requiring the
collection displays a currently valid
OMB control number.
XV. Criminal Penalties
For the purposes of Section 223 of the
Act, the NRC is issuing this proposed
rule that would add a new part 53 and
amend parts 26 and 73 under one or
more of Sections 161b, 161i, or 161o of
the Act, except as noted in proposed
§ 53.9010(b) and § 26.825(b). Willful
violations of the part 53 and part 26
regulations not listed in proposed
§ 53.9010(b) and § 26.825(b) would be
subject to criminal enforcement.
Criminal penalties as they apply to
regulations in part 53 would be
discussed in § 53.9010.
XVI. Voluntary Consensus Standards
The NTTAA requires that Federal
agencies use technical standards that are
developed or adopted by voluntary
consensus standards bodies unless the
use of such a standard is inconsistent
with applicable law or otherwise
impractical. In this proposed rule, the
NRC would revise regulations by adding
a risk-informed, technology-inclusive
regulatory framework for commercial
advanced nuclear reactors. This action
does not constitute the establishment of
a standard that contains generally
applicable requirements.
XVII. Availability of Guidance
As discussed in section II,
Background, of this document, the
NRC’s development of proposed part 53
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built upon recent and ongoing activities
such as those described in SECY–19–
0117. Because a number of those
activities are ongoing to support new
reactor applications under the existing
regulatory framework of 10 CFR parts 50
and 52, the NRC staff identified in its
response to SRM–SECY–20–0032 that
the timing of guidance document
development to support the part 53
rulemaking was a key risk and
uncertainty to publishing the final part
53 rule. To mitigate this risk, the NRC
engaged external stakeholders to ensure
a common prioritization of the
development of these guidance
documents and to work diligently on
those that would be needed to support
this rulemaking, forthcoming
applications, or broader efforts such as
the Advanced Reactor Demonstration
Program being sponsored by the DOE.
The NRC also recognizes that guidance
development to support part 53 and
advanced reactors will continue as the
industry and NRC learn lessons from
licensing reviews and operating
experience. Therefore, the NRC
categorized guidance supporting the
part 53 rulemaking into three categories:
(1) guidance issued or under
development to support applications
under the existing regulatory
framework; (2) implementing guidance
for part 53-specific proposed rule
language; and (3) future guidance
activities that would need to be
completed after the part 53 proposed
rule is published for public comment.
(1) Hundreds of guidance documents
exist for the current fleet of operating
reactors. While some of the guidance is
specific to LWR technologies, other
guidance is technology inclusive in
nature and should be considered, as
appropriate, in the development of all
licensing applications and NRC reviews.
In addition, the NRC has undertaken
efforts to incorporate or reference the
most relevant guidance in its efforts to
develop additional guidance for future
advanced reactors. The NRC has issued
the following guidance to support
licensing reviews of advanced reactors
under the existing regulatory framework
that will continue to inform applicant
development and NRC reviews under
parts 50 and 52. Conforming changes to
these guidance documents would be
needed to ensure they are applicable
under part 53. The NRC will issue
revisions or part 53-related companions
to these guidance documents for public
comment after the publication of this
proposed rule and then finalize and
issue the guidance documents with or
after the final part 53 rule.
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• RG 1.233, ‘‘Guidance for a
Technology-Inclusive, Risk-Informed,
and Performance-Based Methodology
to Inform the Licensing Basis and
Content of Applications for Licenses,
Certifications, and Approvals for NonLight-Water Reactors’’
• RG 1.247 for trial use, ‘‘Acceptability
of Probabilistic Risk Assessment
Results for Non-Light-Water Reactor
Risk-Informed Activities’’
• NUREG–2246, ‘‘Fuel Qualification for
Advanced Reactors’’
• RG 1.87, Revision 2, ‘‘Acceptability of
ASME Code, Section III, Division 5,
‘‘High Temperature Reactors’’
• RG 1.246, ‘‘Acceptability of ASME
Code, Section XI, Division 2,
‘Requirements for Reliability And
Integrity Management (RIM) Programs
for Nuclear Power Plants,’ for NonLight Water Reactors’’
Also, the NRC continues to develop
additional guidance to support licensing
reviews of advanced reactors under the
existing regulatory framework. Some of
these guidance documents have been
issued and others will be issued before
the finalization of part 53 to support
near-term applicants and NRC reviews.
For example, the NRC has been and
continues to be engaged with the DOE
and industry to develop content of
application guidance and other
regulatory guidance for advanced
reactors to support applications and
subsequent operations under the
existing regulatory framework. These
guidance documents, such as the
industry-led Technology-Inclusive
Content of Application Project guidance
found in NEI 21–07, Revision 1, and the
NRC-led Advanced Reactor Content of
Application Project (ARCAP) interim
staff guidance (ISG) documents and
NRC regulatory guidance endorsing NEI
21–07, Revision 1, will support
developers in preparing advanced
reactor applications. These guidance
documents provide an overview of the
information that should be included in
an advanced reactor application, a
review roadmap for the NRC with the
principal purpose of ensuring
consistency, quality, and uniformity of
NRC reviews, and a well-defined base
from which the NRC can evaluate
proposed changes in the scope and
requirements of reviews. While specific
sections of the information are primarily
aligned with the LMP methodology, as
endorsed in RG 1.233, as one acceptable
process for applicants to use when
developing portions of an application,
the concepts and general information
may be used to inform the review of an
application submitted using other
traditional licensing approach
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methodologies (as applicable). Other
sections of the information are generally
applicable and independent of the
methodology used to develop an
advanced reactor application. The
ARCAP ISGs provide references to
numerous regulatory guidance
documents that should be considered by
both applicants and the NRC in
developing and reviewing, respectively,
advanced reactor applications. The NRC
has issued the following documents
separately from this proposed rule. The
NRC may issue other, related guidance
documents with or after the final part 53
rule.
• RG 1.253, ‘‘Guidance for a Technology
Inclusive Content of Application
Methodology to Inform the Licensing
Basis and Content of Applications for
Licenses, Certifications, and
Approvals for Non-Light-Water
Reactors’’
• DANU–ISG–2022–01, ‘‘Advanced
Reactor Content of Application
Project, ‘Review of Risk-Informed,
Technology-Inclusive Advanced
Reactor Applications—Roadmap’ ’’
• DANU–ISG–2022–02, ‘‘Advanced
Reactor Content of Application
Project Chapter 2, ‘Site Information’ ’’
• DANU–ISG–2022–03, ‘‘Advanced
Reactor Content of Application
Project Chapter 9, ‘Control of Routine
Plant Radioactive Effluents, Plant
Contamination and Solid Waste’ ’’
• DANU–ISG–2022–04, ‘‘Advanced
Reactor Content of Application
Project Chapter 10, ‘Control of
Occupational Dose’ ’’
• DANU–ISG–2022–05, ‘‘Advanced
Reactor Content of Application
Project Chapter 11, ‘Organization and
Human-System Considerations’ ’’
• DANU–ISG–2022–06, ‘‘Advanced
Reactor Content of Application
Project Chapter 12, ‘Post-Construction
Inspection, Testing, and Analysis
Program’ ’’
• DANU–ISG–2022–07, ‘‘Advanced
Reactor Content of Application
Project, ‘Risk-Informed Inservice
Inspection/Inservice Testing’ ’’
• DANU–ISG–2022–08, ‘‘Advanced
Reactor Content of Application
Project, ‘Risk-Informed Technical
Specifications’ ’’
• DANU–ISG–2022–09, ‘‘Advanced
Reactor Content of Application
Project, ‘Risk-Informed, PerformanceBased Fire Protection Program (for
Operations)’ ’’
• RG 1.242, ‘‘Performance-Based
Emergency Preparedness for Small
Modular Reactors, Non-Light-Water
Reactors, and Non-Power Production
or Utilization Facilities’’
• RG 4.7, ‘‘General Site Suitability
Criteria for Nuclear Power Stations’’
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(2) The NRC is issuing for comment
nine draft guidance documents for the
implementation of the proposed
requirements in this rulemaking. The
guidance is available in ADAMS under
the Accession Numbers as indicated
under the ‘‘Availability of Documents’’
section in this document. Comments on
this draft regulatory guidance may be
submitted by the methods outlined in
the ADDRESSES section of this document.
Interested persons may obtain
information and comment submissions
related to the draft guidance by
searching on https://www.regulations.gov
under Docket ID NRC–2019–0062.
• DG–1413, ‘‘Technology-Inclusive
Identification of Licensing Events for
Commercial Nuclear Plants’’
This DG describes an acceptable
approach for identifying licensing
events that can be used to inform the
design basis, licensing basis, and
content of applications for commercial
nuclear plants, including large LWRs
and non-LWRs. It applies to nuclear
power reactor designers, applicants, and
licensees of commercial nuclear plants
applying for permits, licenses,
certifications, and approvals under parts
50, 52, and 53. In this DG, the term
‘‘licensing events’’ is used in a generic
sense to refer to collections of
designated event categories such as, but
not limited to AOOs, DBAs, DBEs, and
postulated accidents. Specifically, this
DG provides an acceptable approach for:
(1) conducting a comprehensive and
systematic search for initiating events;
(2) using a systematic process to
delineate a comprehensive set of event
sequences; (3) grouping initiating events
and event sequences into designated
licensing event categories; and (4)
providing assurance that the set of
licensing events is complete.
• DG–5073, ‘‘Fitness For Duty Programs
for Commercial Nuclear Plants And
Manufacturing Facilities Licensed
Under 10 CFR part 53’’
This DG describes guidance for
applicants under part 53 and licensees
and other entities described in § 26.3(f)
who would elect to or be required to
implement FFD programs for facilities
licensed under part 53. The FFD
program requirements would be
detailed in subpart M of part 26 and
involve, in part, policies, procedures,
drug and alcohol testing, laboratory
requirements, behavioral observation,
MRO responsibilities, fitness
determinations, reporting, and
recordkeeping. The FFD program for
facilities licensed under part 53 subject
to part 26 would also include
requirements for a PMRP and FFD
program change control that licensees or
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other entities must implement to
maintain an effective FFD program.
• DG–5074, ‘‘Access Authorization
Program for Commercial Nuclear
Plants’’
This DG describes a method that the
staff considers acceptable to comply
with requirements in proposed § 73.120,
‘‘Access authorization program for
commercial nuclear plants,’’ related to
an AA program. This document
provides guidance and would be one
NRC-approved method (not the only
method) for meeting regulatory
requirements for part 53. The proposed
language in § 73.120 would provide
flexibility through availability of the use
of an alternate approach, commensurate
with risk and consequence to public
health and safety, for part 53 applicants
who demonstrate in an analysis that the
offsite consequences satisfy the criterion
defined in proposed § 53.860(a)(2)(i).
• DG–5075, ‘‘Establishing Cybersecurity
Programs for Commercial Nuclear
Plants Licensed Under 10 CFR part
53’’
This DG describes an approach the
NRC staff deems acceptable for
complying with the Commission’s
proposed regulations for establishing,
implementing, and maintaining a
cybersecurity program at commercial
nuclear plants that would be licensed
under part 53. This guidance provides
an approach for meeting the
requirements of proposed § 73.110,
‘‘Technology-inclusive requirements for
protection of digital computer and
communication systems and networks.’’
• DG–5076, ‘‘Guidance for Technology
Inclusive Requirements for Physical
Protection of Licensed Activities at
Commercial Nuclear Plants’’
This DG describes methods and
approaches that the NRC staff considers
acceptable for meeting the proposed
physical security requirements of part
53 and § 73.100. The guidance is
intended to provide methods and
considerations for complying with
§ 53.440(f) safety and security design
process considerations, determining
eligibility for meeting the performance
criterion in § 53.860 to relieve the
applicant from the applicable
requirements to defend against
radiological sabotage outlined in § 73.55
or § 73.100, and (if the required analysis
for eligibility is not satisfied) applying
the physical security requirements of
§ 73.100 as an alternative pathway from
§ 73.55 for protection against
radiological sabotage.
• DG–5078, ‘‘Fatigue Management for
Nuclear Power Plant Personnel at
PO 00000
Frm 00095
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87011
Commercial Nuclear Plants Licensed
Under 10 CFR part 53’’
This DG describes proposed methods
that the NRC staff considers acceptable
for addressing certain aspects of FFD
programs that would be established at
commercial nuclear facilities licensed
under part 53. This guidance, in
conjunction with the existing RG 5.73,
‘‘Fatigue Management for Nuclear Plant
Personnel,’’ would provide
comprehensive guidance regarding
acceptable methods for the development
and implementation of licensee fatiguemanagement programs.
The NRC is issuing for public
comment the following draft ISG
documents for the implementation of
NRC staff review of applications under
the proposed requirements in this
rulemaking:
• DRO–ISG–2023–01, ‘‘Operator
Licensing Programs’’
This draft ISG provides guidance for
the review of tailored operator licensing
programs that are submitted for review
consistent with the technical
requirements of proposed § 53.730(g).
This guidance primarily addresses the
review of operator licensing
examination processes to facilitate the
ability of reviewers to assess whether a
proposed approach to the testing of
licensed operators and trainees reflects
sound assessment testing practices that
are suitable for the screening of
competent licensed operators.
Additionally, this ISG provides further
review guidance in other areas such as
licensed operator continuing training
and proficiency programs.
• DRO–ISG–2023–02, ‘‘Interim Staff
Guidance Augmenting NUREG–1791,
‘Guidance for Assessing Exemption
Requests from the Nuclear Power
Plant Licensed Operator Staffing
Requirements Specified in 10 CFR
50.54(m),’ for Licensing Commercial
Nuclear Plants under 10 CFR part 53’’
This draft ISG provides guidance for
the review of customized facility
operator staffing plans that are
submitted for review consistent with the
technical requirements of proposed
§ 53.730(f). This ISG is structured as a
companion document to the existing
NUREG–1791 and adapts the existing
HFE-based methodologies of that
document for use in the evaluation of
staffing plans that would be submitted
within the context of part 53 facilities.
Additionally, this ISG provides further
guidance to address other staffingrelated considerations, such as
provisions for engineering expertise.
• DRO–ISG–2023–03, ‘‘Development of
Scalable Human Factors Engineering
Review Plans’’
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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
This draft ISG applies to the HFE
review of applications for OLs, COLs,
DCs, and standard design approvals for
commercial nuclear plants submitted
under proposed part 53. The purpose of
this ISG is to facilitate NRC
understanding of an acceptable method
for developing a scalable (i.e.,
application-specific) plan for the review
of these applications for compliance
with applicable HFE requirements. The
ISG describes a process and provides
implementation guidance for the NRC to
tailor HFE review plans to each
application to achieve an effective and
efficient review.
(3) The NRC has identified future
guidance activities that would need to
be completed after the part 53 proposed
rule is published for public comment to
support advanced reactor applications
and NRC reviews. For example, the NRC
recognizes that new guidance would be
needed for the implementation of
provisions in proposed § 53.620(d) and
the associated licensing provisions in
proposed subpart H that would allow
and establish requirements for the
loading of fuel into a manufactured
reactor for subsequent transport to and
use at a commercial nuclear plant that
will operate the facility pursuant to a
COL. The NRC has not yet initiated the
development of guidance documents in
this category but will engage
stakeholders during the development of
these documents to ensure common
prioritization. In addition, the NRC
works with standards development
organizations, advanced reactor
developers, DOE, and other stakeholders
to identify and facilitate new consensus
codes and standards needed for
advanced reactor development. The
NRC will continue its membership and
participation on standards development
committees and working groups to
support standards for advanced reactor
technologies, where appropriate.
XVIII. Public Meeting
The NRC will conduct a public
meeting on this proposed rule for the
purpose of describing the proposed rule
and implementation guidance to the
public and answering questions from
the public on the proposed rule and
implementation guidance.
The NRC will publish a notice of the
public meeting’s location, time, and
agenda on the NRC’s public meeting
website at least 10 calendar days before
the meeting. Stakeholders should
monitor the NRC’s public meeting
website for information about the public
meeting at: https://www.nrc.gov/publicinvolve/public-meetings/index.cfm.
XIX. Availability of Documents
The documents identified in the
following table are available to
interested persons through one or more
of the following methods, as indicated.
ADAMS accession No./Web link/
Federal Register Citation
Document
Proposed Rule Documents
Federal Register Notification, ‘‘Proposed Rule: Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors,’’ October, 2024.
‘‘Draft Environmental Assessment for the Proposed Rule—Risk Informed, Technology-Inclusive
Regulatory Framework for Advanced Reactors,’’ October, 2024.
‘‘Draft Regulatory Analysis for the Proposed Rule: Risk Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors,’’ October, 2024.
ML24095A161.
ML24095A163.
ML24095A166.
Information Collection Documents
Draft Supporting Statement for Information Collection Analysis—10 CFR Part 53 .......................
Draft Supporting Statement for Information Collection Analysis—10 CFR Part 26 .......................
Draft Supporting Statement for Information Collection Analysis—10 CFR Part 50 .......................
Draft Supporting Statement for Information Collection Analysis—10 CFR Part 73 .......................
Draft Supporting Statement for Information Collection Analysis—NRC Form 361S ......................
Draft Supporting Statement for Information Collection Analysis—NRC Form 366 ........................
Draft Supporting Statement for Information Collection Analysis—NRC Form 893 and 894 ..........
Proposed Rule—Part 26 Burden Tables for Risk Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors.
Proposed Rule—Part 50 Burden Tables for Risk Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors.
Proposed Rule—Part 53 Burden Tables for Risk Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors.
Proposed Rule—Part 73 Burden Tables for Risk Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors.
Draft NRC Form 361S, ‘‘Part 53 Plant Event Notification Worksheet’’ ..........................................
Draft NRC Form 366, ‘‘Licensee Event Report (LER)’’ ..................................................................
Draft NRC Form 366A, ‘‘Licensee Event Report (LER) Continuation Sheet’’ ................................
Draft NRC Form 366B, ‘‘Licensee Event Report (LER) (Failure Continuation)’’ ............................
Draft NRC Form 893, ‘‘10 CFR Part 26, Subpart M, Single FFD Policy Violation Form’’ .............
Draft NRC Form 894, ‘‘10 CFR Part 26, Subpart M, Annual Reporting Form for FFD Performance Information’’.
ML21162A109.
ML23030A400.
ML24220A036.
ML23030A576.
ML24220A034.
ML24220A035.
ML24220A033.
ML24240A008.
ML24220A061.
ML24220A060.
ML24240A009.
ML23032A443.
ML23032A445.
ML23032A447.
ML23032A454.
ML23032A435.
ML23032A439.
lotter on DSK11XQN23PROD with PROPOSALS2
Draft Regulatory Guidance Documents
DG–1413, ‘‘Technology-Inclusive Identification Of Licensing Events For Commercial Nuclear
Plants,’’ October, 2024.
DG–5073, ‘‘Fitness-For-Duty Programs For Commercial Nuclear Plants And Manufacturing Facilities Licensed Under 10 CFR Part 53,’’ October, 2024.
DG–5074, ‘‘Access Authorization Program for Commercial Nuclear Plants,’’ October, 2024 .......
DG–5075, ‘‘Establishing Cybersecurity Programs For Commercial Nuclear Plants Licensed
Under 10 CFR Part 53,’’ October, 2024.
DG–5076, ‘‘Guidance for Technology Inclusive Requirements for Physical Protection of Licensed Activities at Commercial Nuclear Plants,’’ October, 2024.
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ML22200A037.
ML22199A246.
ML22199A257.
ML22203A131.
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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
87013
ADAMS accession No./Web link/
Federal Register Citation
Document
DG–5078, ‘‘Fatigue Management For Nuclear Power Plant Personnel At Commercial Nuclear
Plants Licensed Under 10 CFR Part 53,’’ October, 2024.
ML22264A109.
Draft ISG Documents
Draft ISG DRO–ISG–2023–01, ‘‘Operator Licensing Programs,’’ October, 2024 ..........................
Draft ISG DRO–ISG–2023–02, ‘‘Interim Staff Guidance Augmenting NUREG–1791, ‘Guidance
for Assessing Exemption Requests from the Nuclear Power Plant Licensed Operator Staffing
Requirements Specified in 10 CFR 50.54(m),’ for Licensing Commercial Nuclear Plants
under 10 CFR Part 53,’’ October, 2024.
Draft ISG DRO–ISG–2023–03, ‘‘Development of Scalable Human Factors Engineering Review
Plans,’’ October, 2024.
ML22266A066.
ML22266A068.
ML22266A072.
Other References
American National Standards Institute/ANS–3.4–2013, ‘‘Medical Certification And Monitoring Of
Personnel Requiring Operator Licenses For Nuclear Power Plants’’.
ASME/ANS RA–S–1.4–2021, ‘‘Probabilistic Risk Assessment Standard for Advanced Non-Light
Water Reactor Nuclear Power Plants’’.
ASCE/SEI 43–19, ‘‘Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities’’.
Federal Register notification—Final policy statement, ‘‘Use of Probabilistic Risk Assessment
Methods in Nuclear Regulatory Activities; Final Policy Statement,’’ dated August 16, 1995.
Federal Register notification—Final rule, ‘‘Fitness-for-Duty Programs,’’ dated June 7, 1989 ......
Federal Register notification—Final rule, ‘‘Fitness for Duty Programs,’’ dated March 31, 2008 ..
Federal Register notification—Final rule, ‘‘Licenses, Certifications, and Approvals for Nuclear
Power Plants,’’ dated August 28, 2007.
Federal Register notification—Final rule, ‘‘Station Blackout,’’ dated June 21, 1988 ....................
Federal Register notification—Final rule, ‘‘Technical Specifications,’’ dated July 19, 1995 .........
Federal Register notification—Guidance, ‘‘Mandatory Guidelines for Federal Workplace Drug
Testing Programs,’’ dated January 23, 2017.
Federal Register notification—Guidance, ‘‘Mandatory Guidelines for Federal Workplace Drug
Testing Programs—Oral/Fluid,’’ dated October 25, 2019.
Federal Register notification—Policy Statement, ‘‘Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants,’’ dated August 8, 1985.
Federal Register notification—Policy Statement, ‘‘Safety Goals for the Operation of Nuclear
Power Plants; Policy Statement; Correction and Republication,’’ dated August 21, 1986.
Federal Register notification—Policy Statement, ‘‘Tribal Policy Statement,’’ dated January 9,
2017.
Federal Register notification—Policy Statement, ‘‘Policy Statement on the Regulation of Advanced Reactors,’’ dated October 14, 2008.
Federal Register notification—Policy Statement, ‘‘Final Safety Culture Policy Statement,’’
dated June 14, 2011.
Federal Register notification—Proposed rule, ‘‘Emergency Preparedness for Small Modular
Reactors and Other New Technologies,’’ dated May 12, 2020.
Federal Register notification—Proposed rule, ‘‘Regulatory Improvements for Production and
Utilization Facilities Transitioning to Decommissioning,’’ dated March 3, 2022.
Federal Register notification—Public meeting, ‘‘Reporting Requirements for Nonemergency
Events at Nuclear Power Plants,’’ dated November 29, 2021.
ICRP, Publication 2 ‘‘Permissible dose for internal radiation,’’ dated 1960 ...................................
ICRP, Publication 26 ‘‘Recommendations of the ICRP,’’ dated 1977 ............................................
lotter on DSK11XQN23PROD with PROPOSALS2
ICRP, Publication 30 ‘‘Limits for Intakes of Radionuclides by Workers,’’ dated 1979 ...................
Letter to Chairman Hanson, NRC, ‘‘Final Letter on Draft 10 CFR Part 53 Rulemaking Language,’’ dated November 22, 2022.
Letter to Chairman Hanson, NRC, ‘‘Fourth Interim Letter on 10 CFR Part 53 Rulemaking Language,’’ dated August 2, 2022.
Letter to Chairman Hanson, NRC, ‘‘Preliminary Proposed Rule Language For 10 CFR Part 53,
Regulation of Advanced Nuclear Reactors, Interim Report,’’ dated May 30, 2021.
Letter to Chairman Hanson, NRC, ‘‘Preliminary Rule Language For 10 CFR Part 53, Subpart F,
‘Requirements for Operations,’ Interim Report,’’ dated February 17, 2022.
Letter to Chairman Rempe, ACRS, ‘‘Response to the Advisory Committee on Reactor Safeguards, ‘Fourth Interim Letter on 10 CFR Part 53 Rulemaking Language,’’’ dated September
30, 2022.
Letter to Chairman Rempe, ACRS, ‘‘Response to the Advisory Committee on Reactor Safeguards Letter on Preliminary Rule Language for 10 CFR Part 53, Subpart F, ‘Requirements
for Operations,’ Interim Report,’’ dated March 30, 2022.
Letter to Chairman Sunseri, ACRS, ‘‘Part 53, Licensing and Regulation of Advanced Nuclear
Reactors,’’ dated November 24, 2020.
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https://webstore.ansi.org/Standards/ANSI/
ansians2013.
https://www.asme.org/codes-standards/findcodes-standards/probabilistic-risk-assessment-standard-for-advanced-non-light-waterreactor-nuclear-power-plants/2021/pdf.
https://doi.org/10.1061/9780784415405.
60 FR 42622.
54 FR 24468.
73 FR 16966.
72 FR 49352.
53 FR 23203.
60 FR 36953, 36955.
82 FR 7920.
84 FR 57554.
50 FR 32138.
51 FR 30028.
82 FR 2402.
73 FR 60612.
76 FR 34773.
85 FR 28436.
87 FR 12254.
86 FR 67669.
https://www.icrp.org/publication.asp?id=icrp%20
publication%202.
https://www.icrp.org/publication.asp?id=
ICRP%20Publication%2026.
https://www.icrp.org/publication.asp?id=
ICRP%20Publication%2030%20(Index).
ML22319A104.
ML22196A292.
ML21140A354.
ML22040A361.
ML22249A073.
ML22063A012.
ML20311A006.
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ADAMS accession No./Web link/
Federal Register Citation
Document
lotter on DSK11XQN23PROD with PROPOSALS2
Letter to Chairman Svinicki, NRC, ‘‘10 CFR Part 53, Licensing and Regulation of Advanced
Nuclear Reactors,’’ dated October 21, 2020.
Michigan v. EPA, 135 S. Ct. 2699 (2015) ......................................................................................
National Library of Medicine, National Institutes of Health, Workshop Summary, ‘‘The Evolution
of Telehealth: Where Have We Been and Where Are We Going?,’’ dated November 2012.
NEI 18–04, Rev. 1, ‘‘Risk-Informed Performance-Based Technology-Inclusive Guidance for
Non-Light Water Reactors,’’ dated August 2019.
NIA, ‘‘Clarifying ‘Major Portions’ of a Reactor Design in Support of a Standard Design Approval,’’ dated April 2017.
NRC, ‘‘A Regulatory Review Roadmap for Non-Light Water Reactors,’’ dated December 2017 ..
NRC, ‘‘Manufacturing License ML–1 for Production of Up to Eight Floating Nuclear Plants,’’
dated September 30, 1982.
NRC, ‘‘Risk-Informed and Performance-Based Human-System Considerations for Advanced
Reactors,’’ dated March 2021.
NRC Form 890, ‘‘Single Positive Test Form’’ .................................................................................
NRC Form 891, ‘‘Annual Reporting for Drug and Alcohol Tests’’ ..................................................
NRC From 892, ‘‘Annual Fatigue Reporting Form’’ ........................................................................
NUREG–0654/FEMA–REP–1, Revision 2, ‘‘Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants,’’
dated December 2019.
NUREG–0880, ‘‘Safety Goals for Nuclear Power Plant Operation,’’ dated May 1983 ..................
NUREG–1530, Revision 1, ‘‘Reassessment of NRC’s Dollar Per Person-Rem Conversion Factor Policy, Final Report,’’ dated February 2022.
NUREG/BR–0058, Revision 5, ‘‘Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory
Commission,’’ dated April 2017.
NUREG/CR–5884, ‘‘Revised Analyses of Decommissioning for the Reference Pressurized
Water Reactor Power Station,’’ dated November 1995.
NUREG/CR–6187, Volume 1, ‘‘Revised Analyses of Decommissioning for the Reference Boiling Water Reactor Power Station,’’ dated July 1996.
OMB Circular No. A–119, ‘‘Federal Participation in the Development and Use of Voluntary Consensus Standards and in Conformity Assessment Activities,’’ dated February 19, 1998.
PNNL, Technical Letter Report, ‘‘The Use of Electronic Communications to Perform Determinations of Fitness,’’ dated August 2017.
Pre-decisional DG, ‘‘Technology-Inclusive, Risk-Informed, and Performance-Based Methodology for Seismic Design of Commercial Nuclear Plants,’’ dated October 3, 2022.
Research Information Letter 2021–04, ‘‘Feasibility Study on a Potential Consequence-Based
Seismic Design Approach for Nuclear Facilities,’’ dated April 2021.
RG 1.110, Revision 1, ‘‘Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled
Nuclear Power Reactors,’’ dated October 2013.
RG 1.134, Revision 4, ‘‘Medical Assessment Of Licensed Operators Or Applicants For Operator Licenses At Nuclear Power Plants,’’ dated September 2014.
RG 1.174, ‘‘An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions
on Plant-Specific Changes to the Licensing Basis,’’ Revision 3, dated January 2018.
RG 1.208, ‘‘A Performance-Based Approach to Define the Site-Specific Earthquake Ground
Motion,’’ dated March 2007.
RG 1.232, ‘‘Guidance for Developing Principal Design Criteria for Non-Light-Water Reactors,’’
Revision 0, dated April 2018.
RG 1.233, Revision 0, ‘‘Guidance for a Technology-Inclusive, Risk-Informed, and PerformanceBased Methodology to Inform the Licensing Basis and Content of Applications for Licenses,
Certifications, and Approvals for Non-Light-Water Reactors,’’ dated June 2020.
RG 1.247, ‘‘Acceptability of Probabilistic Risk Assessment Results for Non-Light-Water Reactor
Risk-Informed Activities,’’ issued March 2022 for trial use.
RG 5.73, ‘‘Fatigue Management for Nuclear Power Plant Personnel,’’ dated March 20, 2009 ....
RG 5.77, ‘‘Insider Mitigation Program,’’ Revision 1, dated September 08, 2022 ...........................
RG 5.81, ‘‘Target Set Identification and Development for Nuclear Power Reactors,’’ Revision 1,
dated December 2019 (non-public).
SECY–18–0096, ‘‘Functional Containment Performance Criteria For Non-Light-Water-Reactors,’’ dated September 28, 2018.
SECY–19–0117, ‘‘Technology-Inclusive, Risk-Informed, and Performance-Based Methodology
to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and
Approvals for Non-Light-Water Reactors,’’ dated December 2019.
SECY–20–0032, ‘‘Rulemaking Plan on ‘Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors (RIN–3150–AK31; NRC–2019–0062,’’’ dated April 13, 2020.
SECY–20–0070, ‘‘(Redacted) Technical Evaluation of the Security Bounding Time Concept for
Operating Nuclear Power Plants,’’ dated November 8, 2021.
SECY–22–0072, ‘‘Proposed Rule: Alternative Physical Security Requirements for Advanced
Reactors (RIN 3150–AK19),’’ dated August 2, 2022.
SECY–83–293, ‘‘Amendments to 10 CFR 50 Related to Anticipated Transients Without Scram
(ATWS) Events,’’ dated July 19, 1983.
SECY–93–092, ‘‘Issues Pertaining to the Advanced Reactor (PRISM, MHTGR, and PIUS) and
CANDU 3 Designs and their Relationship to Current Regulatory Requirements,’’ dated April
8, 1993.
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ML20295A647.
https://www.ncbi.nlm.nih.gov/books/
NBK207141/.
ML19241A472.
https://www.nuclearinnovationalliance.org/clarifying-major-portions-reactor-design-supportstandard-design-approval.
ML17312B567.
ML20070J215.
ML21069A003.
ML22013B187.
ML22013B240.
ML22013B250.
ML19347D139.
ML071770230.
ML22053A025.
ML17100A480.
ML14008A187.
ML14008A186.
https://obamawhitehouse.archives.gov/omb/circulars_a119_a119fr.
ML18081A607.
ML22276A149.
ML21113A066.
ML13241A052.
ML14189A385.
ML17317A256.
ML070310619.
ML17325A611.
ML20091L698.
ML21235A008.
ML083450028.
ML16342B024.
ML13151A355.
ML18115A157.
ML18311A264 (package).
ML19340A056.
ML20126G265 (package).
ML21334A003 (package).
ML21278A823 (non-public); ML21278A994
(non-public).
ML040210725.
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ADAMS accession No./Web link/
Federal Register Citation
Document
SRM–SECY–10–0121, ‘‘Modifying the Risk-Informed Regulatory Guidance for New Reactors,’’
dated March 2, 2011.
SRM–SECY–17–0100, ‘‘Security Baseline Inspection Program Assessment Results and Recommendations for Program Efficiencies,’’ dated October 8, 2018.
SRM–SECY–20–0032, ‘‘Rulemaking Plan on ‘Risk-Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors (RIN–3150–AK31; NRC–2019–0062),’’’ dated October 2,
2020.
SRM–SECY–20–0045, ‘‘Population Related Siting Considerations for Advanced Reactors,’’
dated July 30, 2022.
SRM–SECY–98–144, ‘‘Staff Requirements—SECY–98–144—White Paper on Risk-Informed
and Performance-Based Regulations,’’ dated February 24, 1999.
SECY–23–0021, ‘‘Proposed Rule: Risk-Informed, Technology-Inclusive Regulatory Framework
for Advanced Reactors (RIN 3150–AK31),’’ March 1, 2023.
SECY–23–0021, Enclosure 1, ‘‘Draft Federal Register Notification’’ ............................................
SECY–23–0021, Enclosure 2, ‘‘Draft Environmental Assessment for the Proposed Rule—Risk
Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors’’.
SECY–23–0021, Enclosure 3, ‘‘Draft Regulatory Analysis for the Proposed Rule: Risk Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors’’.
SECY–23–0021, Enclosure 4, ‘‘Alternative Approaches Considered for Selected Topics During
the Development of 10 CFR Part 53’’.
SECY–23–0021, Enclosure 5, ‘‘Estimated Resources for The Risk-Informed, Technology-Inclusive Regulatory Framework For Advanced Reactors Rulemaking’’.
Staff Requirements—SECY–23–0021, ‘‘Proposed Rule: Risk-Informed, Technology-Inclusive
Regulatory Framework for Advanced Reactors (RIN 3150–AK31),’’ March 4, 2024.
Throughout the development of this
rule, the NRC may post documents
related to this rule, including public
comments, on the Federal rulemaking
website at https://www.regulations.gov
under Docket ID NRC–2019–0062. The
Federal rulemaking website allows you
to receive alerts when changes or
additions occur in a docket folder. To
subscribe: (1) Navigate to the docket
folder (NRC–2019–0062–0012); (2) click
the ‘‘Sign up for Email Alerts’’ link; and
(3) enter your email address and select
how frequently you would like to
receive emails (daily, weekly, or
monthly).
List of Subjects
10 CFR Part 1
lotter on DSK11XQN23PROD with PROPOSALS2
10 CFR Part 2
Administrative practice and
procedure, Antitrust, Byproduct
material, Classified information,
Confidential business information,
Freedom of information, Environmental
protection, Hazardous waste, Nuclear
energy, Nuclear materials, Nuclear
power plants and reactors, Penalties,
Reporting and recordkeeping
requirements, Sex discrimination,
Source material, Special nuclear
material, Waste treatment and disposal.
ML22194A885.
ML003753593.
ML21162A095.
ML21162A102.
ML21162A104.
ML21165A112.
ML22244A001.
ML22304A099 (non-public).
ML24064A047 (package).
10 CFR Part 26
Hazardous materials transportation,
Investigations, Nuclear energy, Nuclear
materials, Penalties, Reporting and
recordkeeping requirements, Security
measures, Special nuclear material.
Administrative practice and
procedure, Alcohol abuse, Alcohol
testing, Appeals, Drug abuse, Drug
testing, Employee assistance programs,
Fitness for duty, Management actions,
Nuclear power plants and reactors,
Privacy, Protection of information,
Radiation protection, Reporting and
recordkeeping requirements.
10 CFR Part 19
Criminal penalties, Environmental
protection, Nuclear Energy, Nuclear
materials, Nuclear power plants and
reactors, Occupational safety and
health, Penalties, Radiation protection,
Reporting and recordkeeping
requirements, Sex discrimination.
Byproduct material, Criminal
penalties, Hazardous waste, Licensed
material, Nuclear energy, Nuclear
materials, Nuclear power plants and
reactors, Occupational safety and
health, Packaging and containers,
Penalties, Radiation protection,
Reporting and recordkeeping
requirements, Source material, Special
nuclear material, Waste treatment and
disposal.
10 CFR Part 21
Nuclear power plants and reactors,
Penalties, Radiation protection,
Reporting and recordkeeping
requirements.
Administrative practice and
procedure, Classified information,
Classified information, Criminal
penalties, Investigations, Penalties,
Jkt 265001
ML20276A293.
10 CFR Part 11
10 CFR Part 25
18:06 Oct 30, 2024
ML18283A072.
Reporting and recordkeeping
requirements, Security measures.
10 CFR Part 10
VerDate Sep<11>2014
ML110610166.
Government employees, Security
measures.
10 CFR Part 20
Flags, Organization and functions
(Government Agencies), Seals and
insignia.
87015
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10 CFR Part 30
Byproduct material, Criminal
penalties, Government contracts,
Intergovernmental relations, Isotopes,
Nuclear energy, Nuclear materials,
Penalties, Radiation protection,
Reporting and recordkeeping
requirements, Whistleblowing.
10 CFR Part 40
Criminal penalties, Exports,
Government contracts, Hazardous
materials transportation, Hazardous
waste, Nuclear energy, Nuclear
materials, Penalties, Reporting and
recordkeeping requirements, Source
material, Uranium, Whistleblowing.
10 CFR Part 50
Administrative practice and
procedure, Antitrust, Backfitting,
Classified information, Criminal
penalties, Education, Emergency
planning, Fire prevention, Fire
protection, Intergovernmental relations,
Nuclear power plants and reactors,
Penalties, Radiation protection, Reactor
siting criteria, Reporting and
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10 CFR Part 75
Criminal penalties, Intergovernmental
relations, Nuclear energy, Nuclear
materials, Nuclear power plants and
reactors, Penalties, Reporting and
recordkeeping requirements, Security
measures, Treaties.
recordkeeping requirements,
Whistleblowing.
10 CFR Part 51
Administrative practice and
procedure, Environmental impact
statements, Hazardous waste, Nuclear
energy, Nuclear materials, Nuclear
power plants and reactors, Reporting
and recordkeeping requirements.
10 CFR Part 53
Administrative practice and
procedure, Antitrust, Backfitting,
Construction permit, Combined license,
Classified information, Criminal
penalties, Early site permit, Emergency
planning, Fees, Fire prevention, Fire
protection, Inspection,
Intergovernmental relations, Limited
work authorization, Manufacturing
license, Nuclear power plants and
reactors, Operating license, Penalties,
Prototype, Radiation protection, Reactor
siting criteria, Reporting and
recordkeeping requirements, Standard
design, Standard design certification,
Training programs.
10 CFR Part 70
Classified information, Criminal
penalties, Emergency medical services,
Hazardous materials transportation,
Material control and accounting,
Nuclear energy, Nuclear materials,
Packaging and containers, Penalties,
Radiation protection, Reporting and
recordkeeping requirements, Scientific
equipment, Security measures, Special
nuclear material, Whistleblowing.
10 CFR Part 72
Administrative practice and
procedure, Hazardous waste, Indians,
Intergovernmental relations, Nuclear
energy, Penalties, Radiation protection,
Reporting and recordkeeping
requirements, Security measures, Spent
fuel, Whistleblowing.
10 CFR Part 73
Criminal penalties, Exports,
Hazardous materials transportation,
Imports, Nuclear energy, Nuclear
materials, Nuclear power plants and
reactors, Penalties, Reporting and
recordkeeping requirements, Security
measures.
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10 CFR Part 74
Accounting, Criminal penalties,
Hazardous materials transportation,
Material control and accounting,
Nuclear energy, Nuclear materials,
Packaging and containers, Penalties,
Radiation protection, Reporting and
recordkeeping requirements, Scientific
equipment, Special nuclear material.
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10 CFR Part 95
Classified information, Criminal
penalties, Penalties, Reporting and
recordkeeping requirements, Security
measures.
10 CFR Part 140
Criminal penalties, Extraordinary
nuclear occurrence, Insurance,
Intergovernmental relations, Nuclear
materials, Nuclear power plants and
reactors, Penalties, Reporting and
recordkeeping requirements.
10 CFR Part 150
Criminal penalties, Hazardous
materials transportation,
Intergovernmental relations, Nuclear
energy, Nuclear materials, Penalties,
Reporting and recordkeeping
requirements, Security measures,
Source material, Special nuclear
material.
Reorganization Act of 1974, secs. 201, 203,
204, 205, 209 (42 U.S.C. 5841, 5843, 5844,
5845, 5849); Administrative Procedure Act
(5 U.S.C. 552, 553); Reorganization Plan No.
1 of 1980, 5 U.S.C. Appendix (Reorganization
Plans).
§ 1.43
PART 2—AGENCY RULES OF
PRACTICE AND PROCEDURE
3. The authority citation for part 2
continues to read as follows:
■
Authority: Atomic Energy Act of 1954,
secs. 29, 53, 62, 63, 81, 102, 103, 104, 105,
161, 181, 182, 183, 184, 186, 189, 191, 234
(42 U.S.C. 2039, 2073, 2092, 2093, 2111,
2132, 2133, 2134, 2135, 2201, 2231, 2232,
2233, 2234, 2236, 2239, 2241, 2282); Energy
Reorganization Act of 1974, secs. 201, 206
(42 U.S.C. 5841, 5846); Nuclear Waste Policy
Act of 1982, secs. 114(f), 134, 135, 141
(42 U.S.C. 10134(f), 10154, 10155, 10161);
Administrative Procedure Act (5 U.S.C. 552,
553, 554, 557, 558); National Environmental
Policy Act of 1969 (42 U.S.C. 4332); 44 U.S.C.
3504 note. Section 2.205(j) also issued under
28 U.S.C. 2461 note.
10 CFR Part 170
Byproduct material, Import and
export licenses, Intergovernmental
relations, Non-payment penalties,
Nuclear energy, Nuclear materials,
Nuclear power plants and reactors,
Source material, Special nuclear
material.
§ 2.1
10 CFR Part 171
Annual charges, Approvals,
Byproduct material, Holders of
certificates, Intergovernmental relations,
Nonpayment penalties, Nuclear
materials, Nuclear power plants and
reactors, Registrations, Source material,
Special nuclear material.
For the reasons set out in the
preamble and under the authority of the
Atomic Energy Act of 1954, as amended;
the Energy Reorganization Act of 1974,
as amended; and 5 U.S.C. 552 and 553,
the NRC is proposing the following
amendments to 10 CFR parts 1, 2, 10,
11, 19, 20, 21, 25, 26, 30, 40, 50, 51, 70,
72, 73, 74, 75, 95, 140, 150, 170, and 171
and adding 10 CFR part 53:
*
PART 1—STATEMENT OF
ORGANIZATION AND GENERAL
INFORMATION
1. The authority citation for part 1
continues to read as follows:
■
Authority: Atomic Energy Act of 1954,
secs. 23, 25, 29, 161, 191 (42 U.S.C. 2033,
2035, 2039, 2201, 2241); Energy
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[Amended]
2. In § 1.43, in paragraph (a)(2) remove
the cross reference ‘‘10 CFR parts 50, 52,
and 54’’ and add in its place the cross
reference ‘‘10 CFR parts 50, 52, 53, and
54’’.
■
[Amended]
4. In § 2.1, in paragraph (e) remove the
phrase ‘‘part 52’’ and add in its place
the phrase ‘‘part 52 or part 53’’.
■ 5. In § 2.4, revise the definitions for
‘‘Contested proceeding’’ and ‘‘Facility’’
to read as follows:
■
§ 2.4
Definitions.
*
*
*
*
Contested proceeding means—
(1) A proceeding in which there is a
controversy between the NRC staff and
the applicant for a license or permit
concerning the issuance of the license or
permit or any of the terms or conditions
thereof;
(2) A proceeding in which the NRC is
imposing a civil penalty or other
enforcement action, and the subject of
the civil penalty or enforcement action
is an applicant for or holder of a license
or permit, or is or was an applicant for
or holder of a license or permit, or is or
was an applicant for a standard design
certification under part 52 or part 53 of
this chapter; and
(3) A proceeding in which a petition
for leave to intervene in opposition to
an application for a license or permit
has been granted or is pending before
the Commission.
*
*
*
*
*
Facility means production facility or a
utilization facility as defined in §§ 50.2
and 53.020 of this chapter.
*
*
*
*
*
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§ 2.100
[Amended]
6. In § 2.100, remove the phrase
‘‘subpart E of part 52’’ and add in its
place the phrase ‘‘subpart E of part 52
or subpart H of part 53’’.
■ 7. In § 2.101, revise paragraphs
(a)(3)(i), (a)(5), (a)(9) introductory text
and paragraph (a)(9)(i) to read as
follows:
■
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§ 2.101
Filing of application.
(a) * * *
(3) * * *
(i) Submit to the Director, Office of
Nuclear Reactor Regulation, or Director,
Office of Nuclear Material Safety and
Safeguards, as appropriate, such
additional copies as the regulations in
part 50, subpart A of part 51, and part
53 of this chapter require;
*
*
*
*
*
(5) An applicant for a construction
permit under parts 50 or 53 of this
chapter or a combined license under
parts 52 or 53 of this chapter for a
production or utilization facility which
is subject to § 51.20(b) of this chapter,
and is of the type specified in
§ 50.21(b)(2) or (b)(3); or § 50.22; or part
53, as applicable, of this chapter, or is
a testing facility, may submit the
information required of applicants by
parts 50, 52, or 53 of this chapter in two
parts. One part shall be accompanied by
the information required by § 50.30(f) of
this chapter, § 52.80(b) of this chapter,
or § 53.1100(f) of this chapter, as
applicable. The other part shall include
any information required by § 50.34(a)
and, if applicable, § 50.34a of this
chapter; or §§ 52.79 and 52.80(a) of this
chapter; or §§ 53.1109, 53.1306,
53.1309, and 53.1312 of this chapter; or
§§ 53.1109, 53.1413, 53.1416, and
53.1419 of this chapter, as applicable.
One part may precede or follow other
parts by no longer than 6 months. If it
is determined that either of the parts as
described above is incomplete and not
acceptable for processing, the Director,
Office of Nuclear Reactor Regulation, or
Director, Office of Nuclear Material
Safety and Safeguards, as appropriate,
will inform the applicant of this
determination and the respects in which
the document is deficient. Such a
determination of completeness will
generally be made within a period of 30
days. Whichever part is filed first shall
also include the fee required by
§ 50.30(e) or § 53.1100(e) and § 170.21 of
this chapter and the information
required by §§ 50.33, 50.34(a)(1), and
52.79(a)(1) of this chapter; or
§§ 53.1109, 53.1309, and 53.1416 of this
chapter, as applicable, and § 50.37 or
§ 53.1115, as applicable, of this chapter.
The Director, Office of Nuclear Reactor
Regulation, or Director, Office of
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Nuclear Material Safety and Safeguards,
as appropriate, will accept for docketing
an application for a construction permit
under part 50 or part 53 of this chapter
or a combined license under parts 52 or
53 of this chapter for a production or
utilization facility which is subject to
§ 51.20(b) of this chapter, and is of the
type specified in § 50.21(b)(2) or (b)(3),
or § 50.22, or part 53, as applicable, of
this chapter or is a testing facility where
one part of the application as described
above is complete and conforms to the
requirements of part 50 of this chapter.
The additional parts will be docketed
upon a determination by the Director,
Office of Nuclear Reactor Regulation, or
Director, Office of Nuclear Material
Safety and Safeguards, as appropriate,
that it is complete.
*
*
*
*
*
(9) An applicant for a construction
permit for a utilization facility which is
subject to § 51.20(b) of this chapter and
is of the type specified in § 50.21(b)(2)
or (b)(3), or § 50.22, or part 53 of this
chapter, an applicant for or holder of an
early site permit under part 52 or part
53 of this chapter, or an applicant for a
combined license under parts 52 or 53
of this chapter, who seeks to conduct
the activities authorized under
§ 50.10(d) or § 53.1130 of this chapter
may submit a complete application
under paragraphs (a)(1) through (a)(4) of
this section which includes the
information required by § 50.10(d) or
§ 53.1130 of this chapter. Alternatively,
the applicant (other than an applicant
for or holder of an early site permit) may
submit its application in two parts:
(i) Part one must include the
information required by § 50.33(a)
through (f) or § 53.1109(a) through (e)
and § 53.1306 of this chapter, and the
information required by § 50.10(d)(2)
and (d)(3) or § 53.1130(a)(2) and (a)(3) of
this chapter, as applicable.
*
*
*
*
*
■ 8. In § 2.104, revise paragraph (a) to
read as follows:
§ 2.104
Notice of hearing.
(a) In the case of an application on
which a hearing is required by the Act
or this chapter, or in which the
Commission finds that a hearing is
required in the public interest, the
Secretary will issue a notice of hearing
to be published in the Federal Register.
The notice must be published at least 15
days, and in the case of an application
concerning a limited work
authorization, construction permit, early
site permit, or combined license for a
facility of the type described in
§ 50.21(b) or 50.22, or subpart H of part
53 of this chapter, as applicable, or a
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testing facility, at least 30 days, before
the date set for hearing in the notice.1
In addition, in the case of an application
for a limited work authorization,
construction permit, early site permit, or
combined license for a facility of the
type described in § 50.22 or subpart H
of part 53 of this chapter, as applicable,
or a testing facility, the notice must be
issued as soon as practicable after the
NRC has docketed the application. If the
Commission decides, under
§ 2.101(a)(2), to determine the
acceptability of the application based on
its technical adequacy as well as
completeness, the notice must be issued
as soon as practicable after the
application has been tendered.
*
*
*
*
*
1 If the notice of hearing concerning an
application for a limited work authorization,
construction permit, early site permit, or
combined license for a facility of the type
described in § 50.21(b) or § 50.22, or subpart
H of part 53 of this chapter, as applicable, or
a testing facility, does not specify the time
and place of initial hearing, a subsequent
notice will be published in the Federal
Register which will provide at least 30-day
notice of the time and place of that hearing.
After this notice is given, the presiding
officer may reschedule the commencement of
the initial hearing for a later date or
reconvene a recessed hearing without again
providing at least 30-day notice.
9. In § 2.105, revise paragraph (a)
introductory text and paragraphs (a)(4),
(a)(10), (a)(12), (a)(13), (b)(3)
introductory text, (b)(3)(i), (ii), and (iv)
to read as follows:
■
§ 2.105
Notice of proposed action.
(a) If a hearing is not required by the
Act or this chapter, and if the
Commission has not found that a
hearing is in the public interest, it will,
before acting thereon, publish in the
Federal Register, as applicable, or on
the NRC’s website, https://www.nrc.gov,
or both, at the Commission’s discretion,
either a notice of intended operation
under § 52.103(a) or § 53.1452(a) of this
chapter, as applicable, and a proposed
finding that inspections, tests, analyses,
and acceptance criteria for a combined
license under subpart C of part 52 or
under subpart H of part 53 of this
chapter, have been or will be met, or a
notice of proposed action with respect
to an application for:
*
*
*
*
*
(4) An amendment to an operating
license, combined license, or
manufacturing license for a facility
licensed under § 50.21(b) or § 50.22 or
under subpart H of part 53 of this
chapter, as applicable, or for a testing
facility, as follows:
(i) If the Commission determines
under § 50.58 or § 53.1515 of this
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chapter that the amendment involves no
significant hazards consideration,
though it will provide notice of
opportunity for a hearing pursuant to
this section, it may make the
amendment immediately effective and
grant a hearing thereafter; or
(ii) If the Commission determines
under §§ 50.58 and 50.91 or § 53.1515 of
this chapter, as applicable, that an
emergency situation exists or that
exigent circumstances exist and that the
amendment involves no significant
hazards consideration, it will provide
notice of opportunity for a hearing
pursuant to § 2.106 (if a hearing is
requested, it will be held after issuance
of the amendment);
*
*
*
*
*
(10) In the case of an application for
an operating license for a facility of a
type described in § 50.21(b) or § 50.22,
or part 53 of this chapter or a testing
facility, a notice of opportunity for
hearing shall be issued as soon as
practicable after the application has
been docketed; or
*
*
*
*
*
(12) An amendment to an early site
permit issued under subpart A of part
52, or under subpart H of part 53 of this
chapter, as follows:
(i) If the early site permit does not
provide authority to conduct the
activities allowed under § 50.10(e)(1) or
§ 53.1130(b)(1) of this chapter, the
amendment will involve no significant
hazards consideration, and though the
NRC will provide notice of opportunity
for a hearing under this section, it may
make the amendment immediately
effective and grant a hearing thereafter;
and
(ii) If the early site permit provides
authority to conduct the activities
allowed under § 50.10(e)(1) or
§ 53.1130(b)(1) of this chapter and the
Commission determines under §§ 50.58
and 50.91 or § 53.1515 of this chapter
that an emergency situation exists or
that exigent circumstances exist and
that the amendment involves no
significant hazards consideration, it will
provide notice of opportunity for a
hearing under § 2.106 of this chapter (if
a hearing is requested, which will be
held after issuance of the amendment).
(13) A manufacturing license under
subpart F of part 52 or subpart H of part
53 of this chapter.
(b) * * *
(3) For a notice of intended operation
under § 52.103(a) or § 53.1452(a) of this
chapter, the following information:
(i) The identification of the NRC
action as making the finding required
under § 52.103(g) or § 53.1452(g) of this
chapter;
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(ii) The manner in which the licensee
notifications under § 52.99(c) or
§ 53.1449(c), of this chapter which are
required to be made available by
§ 52.99(e)(2) or § 53.1449(e)(2), of this
chapter may be obtained and examined;
*
*
*
*
*
(iv) Any conditions, limitations, or
restrictions to be placed on the license
in connection with the finding under
§ 52.103(g) or § 53.1452(g) of this
chapter, and the expiration date or
circumstances (if any) under which the
conditions, limitations or restrictions
will no longer apply.
*
*
*
*
*
■ 10. In § 2.106, revise paragraphs (a)(2),
(a)(3), and (b)(2) introductory text to
read as follows:
§ 2.106
Notice of issuance.
(a) * * *
(2) An amendment of a license for a
facility of the type described in
§ 50.21(b) or § 50.22, or part 53 of this
chapter, as applicable, or a testing
facility, whether or not a notice of
proposed action has been previously
published; and
(3) The finding under § 52.103(g) or
§ 53.1452(g) of this chapter.
(b) * * *
(2) In the case of a finding under
§ 52.103(g) or § 53.1452(g) of this
chapter:
*
*
*
*
*
■ 11. In § 2.109, revise paragraphs (b),
(c), and (d) to read as follows:
§ 2.109 Effect of timely renewal
application.
*
*
*
*
*
(b) If the licensee of a nuclear power
plant licensed under § 50.21(b) or
§ 50.22 or under subpart H of part 53 of
this chapter files a sufficient application
for renewal of either an operating
license or a combined license at least 5
years before the expiration of the
existing license, the existing license will
not be deemed to have expired until the
application has been finally determined.
(c) If the holder of an early site permit
licensed under subpart A of part 52 or
under subpart H of part 53 of this
chapter, as applicable, files a sufficient
application for renewal under § 52.29 or
§ 53.1173 of this chapter, as applicable,
at least 12 months before the expiration
of the existing early site permit, the
existing permit will not be deemed to
have expired until the application has
been finally determined.
(d) If the licensee of a manufacturing
license under subpart F of part 52, or
under subpart H of part 53 of this
chapter files a sufficient application for
renewal under § 52.177 or § 53.1295 of
this chapter at least 12 months before
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the expiration of the existing license,
the existing license will not be deemed
to have expired until the application has
been finally determined.
*
*
*
*
*
■ 12. In § 2.110, revise paragraphs (a)(1)
and (b) to read as follows:
§ 2.110 Filing and administrative action on
submittals for standard design approval or
early review of site suitability issues.
(a)(1) A submittal for a standard
design approval under subpart E of part
52 or under subpart H of part 53 of this
chapter shall be subject to §§ 2.101(a)
and 2.390 to the same extent as if it were
an application for a permit or license.
*
*
*
*
*
(b) Upon initiation of review by the
NRC staff of a submittal for an early
review of site suitability issues under
appendix Q to part 50 of this chapter,
or for a standard design approval under
subpart E of part 52 or under subpart H
of part 53 of this chapter, the Director,
Office of Nuclear Reactor Regulation,
shall publish in the Federal Register a
notice of receipt of the submittal,
inviting comments from interested
persons within 60 days of publication or
other time as may be specified, for
consideration by the NRC staff and
ACRS in their review.
*
*
*
*
*
■ 13. In § 2.202, revise paragraph (e) to
read as follows:
§ 2.202
Orders.
*
*
*
*
*
(e)(1) If the order involves the
modification of a part 50 or a part 53
license and is a backfit, the
requirements of § 50.109 or § 53.1590 of
this chapter, as applicable, shall be
followed, unless the licensee has
consented to the action required.
(2) If the order involves the
modification of combined license under
subpart C of part 52, or subpart H of part
53 of this chapter, the requirements of
§ 52.98 or § 53.1443 of this chapter, as
applicable, shall be followed unless the
licensee has consented to the action
required.
(3) If the order involves a change to
an early site permit under subpart A of
part 52 or under subpart H of part 53 of
this chapter, the requirements of § 52.39
or § 53.1188 of this chapter, as
applicable, must be followed, unless the
applicant or licensee has consented to
the action required.
(4) If the order involves a change to
a standard design certification rule
referenced by that plant’s application,
the requirements, if any, in the
referenced design certification rule with
respect to changes must be followed, or,
in the absence of these requirements,
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the requirements of § 52.63 or § 53.1263
of this chapter, as applicable, must be
followed, unless the applicant or
licensee has consented to follow the
action required.
(5) If the order involves a change to
a standard design approval referenced
by that plant’s application, the
requirements of § 52.145 or § 53.1221 of
this chapter, as applicable, must be
followed unless the applicant or
licensee has consented to follow the
action required.
(6) If the order involves a
modification of a manufacturing license
under subpart F of part 52 or under
subpart H of part 53 of this chapter, the
requirements of § 52.171 or § 53.1288 of
this chapter, as applicable, must be
followed, unless the applicant or
licensee has consented to the action
required.
■ 14. In § 2.309, revise paragraphs (a),
(f)(1)(i), (f)(1)(vi) and (vii), (g), (h)(2),
(i)(2), (j) to read as follows:
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§ 2.309 Hearing requests, petitions to
intervene, requirements for standing, and
contentions.
(a) General requirements. Any person
whose interest may be affected by a
proceeding and who desires to
participate as a party must file a written
request for hearing and a specification
of the contentions which the person
seeks to have litigated in the hearing. In
a proceeding under § 52.103 or
§ 53.1452 of this chapter, as applicable,
the Commission, acting as the presiding
officer, will grant the request if it
determines that the requestor has
standing under the provisions of
paragraph (d) of this section and has
proposed at least one admissible
contention that meets the requirements
of paragraph (f) of this section. For all
other proceedings, except as provided in
paragraph (e) of this section, the
Commission, presiding officer, or the
Atomic Safety and Licensing Board
designated to rule on the request for
hearing and/or petition for leave to
intervene, will grant the request/petition
if it determines that the requestor/
petitioner has standing under the
provisions of paragraph (d) of this
section and has proposed at least one
admissible contention that meets the
requirements of paragraph (f) of this
section. In ruling on the request for
hearing/petition to intervene submitted
by petitioners seeking to intervene in
the proceeding on the HLW repository,
the Commission, the presiding officer,
or the Atomic Safety and Licensing
Board shall also consider any failure of
the petitioner to participate as a
potential party in the pre-license
application phase under subpart J of this
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part in addition to the factors in
paragraph (d) of this section. If a request
for hearing or petition to intervene is
filed in response to any notice of
hearing or opportunity for hearing, the
applicant/licensee shall be deemed to be
a party.
*
*
*
*
*
(f) * * *
(1) * * *
(i) Provide a specific statement of the
issue of law or fact to be raised or
controverted, provided further, that the
issue of law or fact to be raised in a
request for hearing under § 52.103(b) or
§ 53.1452(b) of this chapter, as
applicable, must be directed at
demonstrating that one or more of the
acceptance criteria in the combined
license have not been, or will not be
met, and that the specific operational
consequences of nonconformance
would be contrary to providing
reasonable assurance of adequate
protection of the public health and
safety;
*
*
*
*
*
(vi) In a proceeding other than one
under § 52.103 or § 53.1452 of this
chapter provide sufficient information
to show that a genuine dispute exists
with the applicant/licensee on a
material issue of law or fact. This
information must include references to
specific portions of the application
(including the applicant’s
environmental report and safety report)
that the petitioner disputes and the
supporting reasons for each dispute, or,
if the petitioner believes that the
application fails to contain information
on a relevant matter as required by law,
the identification of each failure and the
supporting reasons for the petitioner’s
belief; and
(vii) In a proceeding under § 52.103(b)
or § 53.1452(b) of this chapter, as
applicable, the information must be
sufficient, and include supporting
information showing, prima facie, that
one or more of the acceptance criteria in
the combined license have not been, or
will not be met, and that the specific
operational consequences of
nonconformance would be contrary to
providing reasonable assurance of
adequate protection of the public health
and safety. This information must
include the specific portion of the report
required by § 52.99(c) or § 53.1449(c) of
this chapter, as applicable, which the
requestor believes is inaccurate,
incorrect, and/or incomplete (i.e., fails
to contain the necessary information
required by § 52.99(c) or § 53.1449(c) of
this chapter, as applicable). If the
requestor identifies a specific portion of
the report under § 52.99(c) or
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87019
§ 53.1449(c) of this chapter, as
applicable, as incomplete and the
requestor contends that the incomplete
portion prevents the requestor from
making the necessary prima facie
showing, then the requestor must
explain why this deficiency prevents
the requestor from making the prima
facie showing.
*
*
*
*
*
(g) Selection of hearing procedures. A
request for hearing and/or petition for
leave to intervene may, except in a
proceeding under § 52.103 or § 53.1452
of this chapter, as applicable, also
address the selection of hearing
procedures, taking into account the
provisions of § 2.310. If a request/
petition relies upon § 2.310(d), the
request/petition must demonstrate, by
reference to the contention and the
bases provided and the specific
procedures in subpart G of this part, that
resolution of the contention necessitates
resolution of material issues of fact
which may be best determined through
the use of the identified procedures.
(h) * * *
(2) If the proceeding pertains to a
production or utilization facility (as
defined in § 50.2 or § 53.020 of this
chapter) located within the boundaries
of the State, local governmental body, or
Federally-recognized Indian Tribe
seeking to participate as a party, no
further demonstration of standing is
required. If the production or utilization
facility is not located within the
boundaries of the State, local
governmental body, or Federallyrecognized Indian Tribe seeking to
participate as a party, the State, local
governmental body, or Federallyrecognized Indian Tribe also must
demonstrate standing.
*
*
*
*
*
(i) * * *
(2) Except in a proceeding under
§ 52.103 or § 53.1452 of this chapter, as
applicable, the participant who filed the
hearing request, intervention petition, or
motion for leave to file new or amended
contentions after the deadline may file
a reply to any answer. The reply must
be filed within 7 days after service of
that answer.
*
*
*
*
*
(j) Decision on request/petition. (1) In
all proceedings other than a proceeding
under § 52.103 or § 53.1452 of this
chapter, as applicable, the presiding
officer shall issue a decision on each
request for hearing or petition to
intervene within 45 days of the
conclusion of the initial pre-hearing
conference or, if no pre-hearing
conference is conducted, within 45 days
after the filing of answers and replies
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under paragraph (i) of this section. With
respect to a request to admit amended
or new contentions, the presiding officer
shall issue a decision on each such
request within 45 days of the conclusion
of any pre-hearing conference that may
be conducted regarding the proposed
amended or new contentions or, if no
pre-hearing conference is conducted,
within 45 days after the filing of
answers and replies, if any. In the event
the presiding officer cannot issue a
decision within 45 days, the presiding
officer shall issue a notice advising the
Commission and the parties, and the
notice shall include the expected date of
when the decision will issue.
(2) The Commission, acting as the
presiding officer, shall expeditiously
grant or deny the request for hearing in
a proceeding under § 52.103 or
§ 53.1452 of this chapter, as applicable.
The Commission’s decision may not be
the subject of any appeal under § 2.311.
■ 15. Amend § 2.310 by:
■ a. In paragraphs (a) and (h)
introductory text, removing the crossreference ‘‘parts 30, 32 through 36, 39,
40, 50, 52, 54, 55, 61, 70 and 72 of this
chapter’’ and adding, in its place, the
cross reference ‘‘parts 30, 32 through 36,
39, 40, 50, 52, 53, 54, 55, 61, 70 and 72
of this chapter’’; and
■ b. Revising paragraphs (i) and (j).
The revisions read as follows.
involving a construction permit or
operating license for a facility of a type
described in §§ 50.21(b) or 50.22 or part
53 of this chapter must be held within
sixty (60) days after discovery has been
completed or any other time specified
by the Commission or the presiding
officer.
*
*
*
*
*
■ 17. In § 2.339, revise paragraph (d) to
read as follows:
§ 2.310
*
Selection of hearing procedures.
*
*
*
*
*
(i) In design certification rulemaking
proceedings under part 52 or part 53 of
this chapter, any informal hearing held
under § 52.51 or § 53.1242 of this
chapter, as applicable, must be
conducted under the procedures of
subpart O of this part.
(j) Proceedings on a Commission
finding under § 52.103(c) and (g) or
§ 53.1452(c) and (g) of this chapter, as
applicable, shall be conducted in
accordance with the procedures
designated by the Commission in each
proceeding.
*
*
*
*
*
■ 16. In § 2.329, revise paragraph (a) to
read as follows:
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§ 2.329
Prehearing conference.
(a) Necessity for prehearing
conference; timing. The Commission or
the presiding officer may, and in the
case of a proceeding on an application
for a construction permit or an operating
license for a facility of a type described
in §§ 50.21(b) or 50.22, or part 53 of this
chapter, or a testing facility, must direct
the parties or their counsel to appear at
a specified time and place for a
conference or conferences before trial. A
prehearing conference in a proceeding
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§ 2.339 Expedited decision-making
procedure.
*
*
*
*
*
(d) The provisions of this section do
not apply to an initial decision directing
the issuance of a limited work
authorization under 10 CFR 50.10 or
10 CFR 53.1130; an early site permit
under subpart A of part 52 or under
subpart H of part 53 of this chapter; a
construction permit or construction
authorization under part 50 or part 53
of this chapter; a combined license
under subpart C of part 52 or under
subpart H of part 53 of this chapter; or
a manufacturing license under subpart F
of part 52 or under subpart H of part 53.
■ 18. In § 2.340, revise paragraphs (b),
(c), (d), (f), (i), and (j) to read as follows:
§ 2.340 Initial decision in certain contested
proceedings; immediate effectiveness of
initial decisions; issuance of authorizations,
permits and licenses.
*
*
*
*
(b) Initial decision—combined license
under 10 CFR parts 52 or 53. (1) Matters
in controversy; presiding officer
consideration of matters not put in
controversy by parties. In any initial
decision in a contested proceeding on
an application for a combined license
under parts 52 or 53 of this chapter
(including an amendment to or renewal
of combined license), the presiding
officer shall make findings of fact and
conclusions of law on the matters put
into controversy by the parties and any
matter designated by the Commission to
be decided by the presiding officer. The
presiding officer shall also make
findings of fact and conclusions of law
on any matter not put into controversy
by the parties, but only to the extent that
the presiding officer determines that a
serious safety, environmental, or
common defense and security matter
exists, and the Commission approves of
an examination of and decision on the
matter upon its referral by the presiding
officer under, inter alia, the provisions
of §§ 2.323 and 2.341.
(2) Presiding officer initial decision
and issuance of permit or license.
(i) In a contested proceeding for the
initial issuance or renewal of a
combined license under parts 52 or 53
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of this chapter, or the amendment of a
combined license where the NRC has
not made a determination of no
significant hazards consideration, the
Commission or the Director, Office of
Nuclear Reactor Regulation, as
appropriate after making the requisite
findings, shall issue, deny, or
appropriately condition the permit or
license in accordance with the presiding
officer’s initial decision once that
decision becomes effective.
(ii) In a contested proceeding for the
amendment of a combined license
under parts 52 or 53 of this chapter
where the NRC has made a
determination of no significant hazards
consideration, the Commission or the
Director, Office of Nuclear Reactor
Regulation, as appropriate (appropriate
official), after making the requisite
findings and complying with any
applicable provisions of § 2.1202(a) or
§ 2.1403(a), may issue the amendment
before the presiding officer’s initial
decision becomes effective. Once the
presiding officer’s initial decision
becomes effective, the appropriate
official shall take action with respect to
that amendment in accordance with the
initial decision. If the presiding officer’s
initial decision becomes effective before
the appropriate official issues the
amendment, then the appropriate
official, after making the requisite
findings, shall issue, deny, or
appropriately condition the amendment
in accordance with the presiding
officer’s initial decision.
(c) Initial decision on findings under
10 CFR 52.103 or 10 CFR 53.1452 with
respect to acceptance criteria in nuclear
power reactor combined licenses. In any
initial decision under § 52.103(g) or
§ 53.1452(g) of this chapter with respect
to whether acceptance criteria have
been or will be met, the presiding officer
shall make findings of fact and
conclusions of law on the matters put
into controversy by the parties, and any
matter designated by the Commission to
be decided by the presiding officer.
Matters not put into controversy by the
parties but identified by the presiding
officer as matters requiring further
examination, shall be referred to the
Commission for its determination; the
Commission may, in its discretion, treat
any of these referred matters as a request
for action under § 2.206 and process the
matter in accordance with § 52.103(f) or
§ 53.1452(f) of this chapter.
(d) Initial decision—manufacturing
license under 10 CFR parts 52 or 53. (1)
Matters in controversy; presiding officer
consideration of matters not put in
controversy by parties. In any initial
decision in a contested proceeding on
an application for a manufacturing
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license under subpart C of part 52 or
subpart H of part 53 of this chapter
(including an amendment to or renewal
of a manufacturing license), the
presiding officer shall make findings of
fact and conclusions of law on the
matters put into controversy by the
parties and any matter designated by the
Commission to be decided by the
presiding officer. The presiding officer
also shall make findings of fact and
conclusions of law on any matter not
put into controversy by the parties, but
only to the extent that the presiding
officer determines that a serious safety,
environmental, or common defense and
security matter exists, and the
Commission approves of an
examination of and decision on the
matter upon its referral by the presiding
officer under, inter alia, the provisions
of §§ 2.323 and 2.341.
(2) Presiding officer initial decision
and issuance of permit or license.
(i) In a contested proceeding for the
initial issuance or renewal of a
manufacturing license under subpart C
of part 52 or subpart H of part 53 of this
chapter, or the amendment of a
manufacturing license, the Commission
or the Director, Office of Nuclear
Reactor Regulation, as appropriate, after
making the requisite findings, shall
issue, deny, or appropriately condition
the permit or license in accordance with
the presiding officer’s initial decision
once that decision becomes effective.
(ii) In a contested proceeding for the
initial issuance or renewal of a
manufacturing license under subpart C
of part 52 or subpart H of part 53 of this
chapter, or the amendment of a
manufacturing license, the Commission
or the Director, Office of Nuclear
Reactor Regulation, as appropriate
(appropriate official), may issue the
license, permit, or license amendment
in accordance with § 2.1202(a) or
§ 2.1403(a) before the presiding officer’s
initial decision becomes effective. If,
however, the presiding officer’s initial
decision becomes effective before the
license, permit, or license amendment is
issued under § 2.1202 or § 2.1403, then
the Commission or the Director, Office
of Nuclear Reactor Regulation, as
appropriate, shall issue, deny, or
appropriately condition the license,
permit, or license amendment in
accordance with the presiding officer’s
initial decision.
*
*
*
*
*
(f) Immediate effectiveness of certain
presiding officer decisions. A presiding
officer’s initial decision directing the
issuance or amendment of a limited
work authorization under § 50.10 or
§ 53.1130 of this chapter; an early site
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permit under subpart A of part 52 or
under subpart H of part 53 of this
chapter; a construction permit or
construction authorization under part
50 or part 53 of this chapter; an
operating license under part 50 or part
53 of this chapter; a combined license
under subpart C of part 52 or subpart H
or part 53 of this chapter; a
manufacturing license under subpart F
of part 52 or subpart H of part 53 of this
chapter; a renewed license under part
53 or part 54 of this chapter; or a license
under part 72 of this chapter to store
spent fuel in an independent spent fuel
storage facility (ISFSI) or a monitored
retrievable storage installation (MRS);
an initial decision directing issuance of
a license under part 61 of this chapter;
or an initial decision under § 52.103(g)
or § 53.1452(g) of this chapter that
acceptance criteria in a combined
license have been met, is immediately
effective upon issuance unless the
presiding officer finds that good cause
has been shown by a party why the
initial decision should not become
immediately effective.
*
*
*
*
*
(i) Issuance of authorizations,
permits, and licenses—production and
utilization facilities. The Commission or
the Director, Office of Nuclear Reactor
Regulation, as appropriate, shall issue a
limited work authorization under
§ 50.10 or § 53.1130 of this chapter; an
early site permit under subpart A of part
52 or subpart H of part 53 of this
chapter; a construction permit or
construction authorization under part
50 or part 53 of this chapter; an
operating license under part 50 or part
53 of this chapter; a combined license
under subpart C of part 52 or part 53 of
this chapter; or a manufacturing license
under subpart F of part 52 or part 53 of
this chapter within 10 days from the
date of issuance of the initial decision:
(1) If the Commission or the Director
has made all findings necessary for
issuance of the authorization, permit or
license, not within the scope of the
initial decision of the presiding officer;
and
(2) Notwithstanding the pendency of
a petition for reconsideration under
§ 2.345, a petition for review under
§ 2.341, or a motion for stay under
§ 2.342, or the filing of a petition under
§ 2.206.
(j) Issuance of finding on acceptance
criteria under 10 CFR 52.103 or 10 CFR
53.1452. The Commission or the
Director, Office of Nuclear Reactor
Regulation, as appropriate, shall make
the finding under § 52.103(g) or
§ 53.1452(g) of this chapter, that
acceptance criteria in a combined
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87021
license are met within 10 days from the
date of the presiding officer’s initial
decision:
(1) If the Commission or the Director
is otherwise able to make the finding
under § 52.103(g) or § 53.1452(g) of this
chapter, that the prescribed acceptance
criteria are met for those acceptance
criteria not within the scope of the
initial decision of the presiding officer;
(2) If the presiding officer’s initial
decision—with respect to contentions
that the prescribed acceptance criteria
have not been met—finds that those
acceptance criteria have been met, and
the Commission or the Director
thereafter is able to make the finding
that those acceptance criteria are met;
(3) If the presiding officer’s initial
decision—with respect to contentions
that the prescribed acceptance criteria
will not be met—finds that those
acceptance criteria will be met, and the
Commission or the Director thereafter is
able to make the finding that those
acceptance criteria are met; and
(4) Notwithstanding the pendency of
a petition for reconsideration under
§ 2.345, a petition for review under
§ 2.341, or a motion for stay under
§ 2.342, or the filing of a petition under
§ 2.206.
*
*
*
*
*
§ 2.341
[Amended]
19. In § 2.341(a)(1), remove the phrase
‘‘§ 52.103(c)’’ and add in its place the
phrase ‘‘§ 52.103(c) or § 53.1452(c)’’.
■
§ 2.400
[Amended]
20. In § 2.400, remove the phrase
‘‘parts 50 or 52’’ and add in its place the
phrase ‘‘part 50 or part 52, or
§ 53.1470’’.
■ 21. In § 2.401, revise the section
heading and paragraph (a) to read as
follows:
■
§ 2.401 Notice of hearing on construction
permit or combined license applications
pursuant to appendix N of 10 CFR parts 50,
52, or 53.
(a) In the case of applications under
appendix N of part 50 or § 53.1470 of
this chapter for construction permits for
nuclear power reactors of the type
described in § 50.22 or part 53 of this
chapter, or applications under appendix
N of part 52 or § 53.1470 of this chapter
for combined licenses, the Secretary
will issue notices of hearing pursuant to
§ 2.104.
*
*
*
*
*
■ 22. In § 2.402, revise paragraph (a) to
read as follows:
§ 2.402 Separate hearings on separate
issues; consolidation of proceedings.
(a) In the case of applications under
appendix N of part 50 or § 53.1470 of
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this chapter for construction permits for
nuclear power reactors of a type
described in 10 CFR 50.22 or part 53, or
applications pursuant to appendix N of
part 52 or § 53.1470 of this chapter for
combined licenses, the Commission or
the presiding officer may order separate
hearings on particular phases of the
proceeding, such as matters related to
the acceptability of the design of the
reactor in the context of the site
parameters postulated for the design or
environmental matters.
*
*
*
*
*
§ 2.643 Acceptance and docketing of
application for limited work authorization.
24. In § 2.404, remove the phrase
‘‘appendix N of part 50’’ and add in its
place the phrase ‘‘appendix N to part 50
or § 53.1470’’.
*
*
*
*
(b) The Director will accept for
docketing part one of an application for
a construction permit for a utilization
facility which is subject to § 51.20(b) of
this chapter and is of the type specified
in § 50.21(b)(2) or (3) or § 50.22 or part
53 of this chapter or an application for
a combined license where part one of
the application as described in
§ 2.101(a)(9) is complete. Part one will
not be considered complete unless it
contains the information required by
§ 50.10(d)(3) or § 53.1130(a)(3) of this
chapter. Upon assignment of a docket
number, the procedures in § 2.101(a)(3)
and (4) relating to formal docketing and
the submission and distribution of
additional copies of the application
must be followed.
*
*
*
*
*
§ 2.405
§ 2.645
§ 2.403
[Amended]
23. In § 2.403, remove the phrase
‘‘appendix N of part 50’’ and add in its
place the phrase ‘‘appendix N to part 50
or § 53.1470’’.
■
§ 2.404
[Amended]
■
[Amended]
§ 2.406
[Amended]
26. In § 2.406, remove the phrase
‘‘appendix N of parts 50 or 52’’ and add
in its place the phrase ‘‘appendix N to
part 50 or part 52 or § 53.1470’’.
■
§ 2.500
[Amended]
27. In § 2.500, remove the phrase
‘‘subpart F of part 52’’ and add in its
place the phrase ‘‘subpart F of part 52
or subpart H of part 53’’.
■ 28. In § 2.501, revise the section
heading and paragraph (a) introductory
text to read as follows:
■
§ 2.501 Notice of hearing on application
under 10 CFR parts 52 or 53 for a license
to manufacture nuclear power reactors.
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§ 2.649
(1) An application to construct and/or
operate a production or utilization
facility (including an application for a
limited work authorization under
§§ 50.12 or 53.1130 of this chapter, or an
application for a combined license
under subpart C of 10 CFR part 52, or
under subpart H of 10 CFR part 53;
(2) An application for an early site
permit under subpart A of 10 CFR part
52 or under subpart H of 10 CFR part
53;
(3) An application for a
manufacturing license under subpart F
of 10 CFR part 52 or under subpart H
of 10 CFR part 53;
*
*
*
*
*
(6) Production or utilization facility
licensing actions that involve significant
hazards considerations as defined in
§§ 50.92 or 53.1520 of this chapter.
*
*
*
*
*
§ 2.1301
§ 2.1403
§ 2.800
§ 2.1502
32. In § 2.800, amend paragraphs (c)
and (d) by removing the phrase ‘‘subpart
B of part 52’’ and adding in its place the
phrase ‘‘subpart B of part 52 or subpart
H of part 53’’.
■
[Amended]
33. In § 2.801, remove the phrase
‘‘subpart B of part 52’’ and add in its
place the phrase ‘‘subpart B of part 52
or subpart H of part 53’’.
§ 2.813
[Amended]
34. In § 2.813(a), remove the phrase
‘‘parts 50, 52, and 100’’ and add in its
place the phrase ‘‘parts 50, 52, 53, and
100’’.
■
§ 2.1103
[Amended]
35. In § 2.1103, remove the phrase
‘‘part 50 of this chapter’’ and add in its
place the phrase ‘‘parts 50 or 53 of this
chapter’’.
■ 36. In § 2.1202, revise paragraphs
(a)(1) through (3) and (a)(6) to read as
follows:
■
§ 2.1202
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(a) * * *
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[Amended]
39. In § 2.1500, remove the phrase
‘‘subpart B of part 52’’ and add in its
place the phrase ‘‘subpart B of part 52
or under subpart H of part 53’’.
31. In § 2.649, remove the phrase ‘‘10
CFR 50.10(d)’’ and add in its place the
phrase ‘‘10 CFR 50.10(d) or 10 CFR
53.1130(a)’’.
[Amended]
[Amended]
38. In § 2.1403, remove the phrase ‘‘10
CFR 50.92’’ and add in its place the
phrase ‘‘10 CFR 50.92 or 10 CFR
53.1520’’.
■
■
■
[Amended]
37. In § 2.1301(b), remove ‘‘part 50
and part 52’’ and add in its place ‘‘parts
50, 52, and 53’’.
■
§ 2.1500
[Amended]
■
1 The thirty-day (30) requirement of this
paragraph is not applicable to a notice of the
time and place of hearing published by the
presiding officer after notice of hearing
described in this section has been published.
18:06 Oct 30, 2024
[Amended]
30. In § 2.645, in paragraph (a),
remove the phrase ‘‘§ 50.33(a) through
(f) of this chapter’’ and add in its place
the phrase ‘‘§§ 50.33(a) through (f),
53.1109, and 53.1306(a) or 53.1413 of
this chapter, as applicable,’’.
§ 2.801
(a) In the case of an application under
subpart F of part 52 or subpart H of part
53 of this chapter for a license to
manufacture nuclear power reactors of
the type described in § 50.22 or part 53
of this chapter to be operated at sites not
identified in the license application, the
Secretary will issue a notice of hearing
to be published in the Federal Register
at least 30 days before the date set for
hearing in the notice.1 The notice shall
be issued as soon as practicable after the
application has been docketed. The
notice will state:
*
*
*
*
*
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*
■
25. In § 2.405, remove the phrase
‘‘part 52’’ and add in its place the
phrase ‘‘part 52 or part 53’’.
■
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29. In § 2.643, revise paragraph (b) to
read as follows:
■
[Amended]
40. In § 2.1502, in paragraph (a),
remove the phrase ‘‘§ 52.51(b)’’ and add
in its place the phrase ‘‘§§ 52.51(b) or
53.1242(b)(2)’’; and in paragraph (b)(1),
wherever it appears, remove the phrase
‘‘§ 52.51(a)’’ and add in its place the
phrase ‘‘§§ 52.51(a) or 53.1242(b)’’.
■
PART 10—CRITERIA AND
PROCEDURES FOR DETERMINING
ELIGIBILITY FOR ACCESS TO
RESTRICTED DATA OR NATIONAL
SECURITY INFORMATION OR AN
EMPLOYMENT CLEARANCE
41. The authority citation for part 10
continues to read as follows:
■
Authority: Atomic Energy Act of 1954,
secs. 145, 161 (42 U.S.C. 2165, 2201); Energy
Reorganization Act of 1974, sec. 201 (42
U.S.C. 5841); E.O. 10450, 18 FR 2489, 3 CFR,
1949–1953 Comp., p. 936, as amended; E.O.
10865, 25 FR 1583, 3 CFR, 1959–1963 Comp.,
p. 398, as amended; E.O. 12968, 60 FR 40245,
3 CFR, 1995 Comp., p. 391.
§ 10.1
[Amended]
42. In § 10.1, in paragraph (a)(3)
remove the phrase ‘‘under part 52’’ and
■
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add in its place the phrase ‘‘under parts
52 or 53’’.
§ 10.2
[Amended]
43. In § 10.2, in paragraph (b),
wherever it appears, remove the phrase
‘‘under part 52’’ and add in its place the
phrase ‘‘under parts 52 or 53’’.
■
PART 11—CRITERIA AND
PROCEDURES FOR DETERMINING
ELIGIBILITY FOR ACCESS TO OR
CONTROL OVER SPECIAL NUCLEAR
MATERIAL
Authority: Atomic Energy Act of 1954,
secs. 161, 223 (42 U.S.C. 2201, 2273); Energy
Reorganization Act of 1974, sec. 201 (42
U.S.C. 5841); 44 U.S.C. 3504 note. Section
11.15(e) also issued under 31 U.S.C. 9701; 42
U.S.C. 2214.
[Amended]
45. In § 11.7, in the introductory text,
remove the phrase ‘‘parts 10, 25, 50, 70,
72, 73, and 95 of this chapter’’ and add
in its place the phrase ‘‘parts 10, 25, 50,
53, 70, 72, 73, and 95 of this chapter’’.
■
PART 19—NOTICES, INSTRUCTIONS
AND REPORTS TO WORKERS:
INSPECTION AND INVESTIGATIONS
46. The authority citation for part 19
continues to read as follows:
■
Authority: Atomic Energy Act of 1954,
secs. 53, 63, 81, 103, 104, 161, 223, 234, 1701
(42 U.S.C. 2073, 2093, 2111, 2133, 2134,
2201, 2273, 2282, 2297f); Energy
Reorganization Act of 1974, secs. 201, 211,
401 (42 U.S.C. 5841, 5851, 5891); 44 U.S.C.
3504 note.
47. In § 19.2, revise paragraph (a) to
read as follows:
■
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§ 19.2
Scope.
(a) * * *
(1) All persons who receive, possess,
use, or transfer material licensed by the
NRC under the regulations in parts 30
through 36, 39, 40, 60, 61, 63, 70, or 72
of this chapter, including persons
licensed to operate a production or
utilization facility under part 50, part
52, or part 53 of this chapter, persons
licensed to possess power reactor spent
fuel in an independent spent fuel
storage installation (ISFSI) under part 72
of this chapter, and in accordance with
10 CFR 76.60 to persons required to
obtain a certificate of compliance or an
approved compliance plan under part
76 of this chapter;
(2) All applicants for and holders of
licenses (including construction permits
and early site permits) under parts 50,
52, 53, and 54 of this chapter;
(3) All applicants for and holders of
a standard design approval under
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§ 19.3
*
*
*
*
License means a license issued under
the regulations in parts 30 through 36,
39, 40, 60, 61, 63, 70, or 72 of this
chapter, including licenses to
manufacture, construct and/or operate a
production or utilization facility under
parts 50, 52, 53, or 54 of this chapter.
*
*
*
*
*
Regulated entities means any
individual, person, organization, or
corporation that is subject to the
regulatory jurisdiction of the NRC,
including (but not limited to) an
applicant for or holder of a standard
design approval under subpart E of part
52 or under subpart H of part 53 of this
chapter or a standard design
certification under subpart B of part 52
or under subpart H of part 53 of this
chapter.
*
*
*
*
*
§ 19.11
[Amended]
49. In § 19.11, in paragraph (a)
introductory text, paragraph (b)
introductory text, and paragraph (e)(1),
remove the phrase ‘‘of part 52’’
wherever may appears and add in its
place the phrase ‘‘of part 52 or under
subpart H of part 53’’.
■
§ 19.14
[Amended]
50. In § 19.14, in paragraph (a),
wherever it may appear, remove the
phrase ‘‘of part 52’’ and add in its place
the phrase ‘‘of part 52 or under subpart
H of part 53’’.
■
§ 19.20
[Amended]
51. In § 19.20, add the number ‘‘53,’’
in sequential order.
■
PART 20—STANDARDS FOR
PROTECTION AGAINST RADIATION
52. The authority citation for part 20
continues to read as follows:
■
Authority: Atomic Energy Act of 1954,
secs. 11, 53, 63, 65, 81, 103, 104, 161, 170H,
182, 186, 223, 234, 274, 1701 (42 U.S.C. 2014,
2073, 2093, 2095, 2111, 2133, 2134, 2201,
2210h, 2232, 2236, 2273, 2282, 2021, 2297f);
Energy Reorganization Act of 1974, secs. 201,
202 (42 U.S.C. 5841, 5842); Low-Level
Radioactive Waste Policy Amendments Act
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of 1985, sec. 2 (42 U.S.C. 2021b); 44 U.S.C.
3504 note.
§ 20.1002
Sfmt 4702
[Amended]
53. In § 20.1002, remove the phrase
‘‘parts 30 through 36, 39, 40, 50, 52, 60,
61, 63, 70, or 72 of this chapter’’ and
add in its place the phrase ‘‘parts 30
through 36, 39, 40, 50, 52, 53, 60, 61,
63, 70, or 72 of this chapter’’.
■ 54. In § 20.1003, revise the definition
for ‘‘License’’ to read as follows:
■
§ 20.1003
Definitions.
*
44. The authority citation for part 11
continues to read as follows:
■
§ 11.7
subpart E of part 52 or under subpart H
of part 53 of this chapter; and
(4) All applicants for a standard
design certification under subpart B of
part 52 or under subpart H of part 53 of
this chapter, and those (former)
applicants whose designs have been
certified under that subpart.
*
*
*
*
*
■ 48. In § 19.3, revise the definitions for
‘‘License’’ and ‘‘Regulated entities’’ to
read as follows:
87023
Definitions.
*
*
*
*
*
License means a license issued under
the regulations in parts 30 through 36,
39, 40, 50, 53, 60, 61, 63, 70, or 72 of
this chapter.
*
*
*
*
*
§ 20.1101
[Amended]
55. In § 20.1101, in paragraph (d),
remove the phrase ‘‘subject to § 50.34a’’
and add in its place the phrase ‘‘subject
to §§ 50.34a or 53.260 of this chapter’’.
■
§ 20.1401
[Amended]
56. Amend § 20.1401 by:
a. In paragraph (a), removing the
phrase ‘‘parts 30, 40, 50, 52, 60, 61, 63,
70, and 72 of this chapter’’, and adding
in its place the phrase ‘‘parts 30, 40, 50,
52, 53, 60, 61, 63, 70, and 72 of this
chapter’’; and
■ b. In paragraphs (a) and (c) removing
the phrase ‘‘in accordance with § 50.83’’
and adding in its place the phrase ‘‘in
accordance with §§ 50.83 or 53.1080’’.
■ 57. In § 20.1403, revise paragraph (d)
introductory text to read as follows:
■
■
§ 20.1403 Criteria for license termination
under restricted conditions.
*
*
*
*
*
(d) The licensee has submitted a
decommissioning plan or License
Termination Plan (LTP) to the
Commission indicating the licensee’s
intent to decommission in accordance
with §§ 30.36(d), 40.42(d), 50.82 (a) and
(b), subpart G of part 53, 70.38(d), or
72.54 of this chapter, and specifying
that the licensee intends to
decommission by restricting use of the
site. The licensee shall document in the
LTP or decommissioning plan how the
advice of individuals and institutions in
the community who may be affected by
the decommissioning has been sought
and incorporated, as appropriate,
following analysis of that advice.
*
*
*
*
*
■ 58. In § 20.1404, revise paragraph
(a)(4) introductory text to read as
follows:
§ 20.1404 Alternate criteria for license
termination.
(a) * * *
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(4) Has submitted a decommissioning
plan or License Termination Plan (LTP)
to the Commission indicating the
licensee’s intent to decommission in
accordance with § 30.36(d), 40.42(d),
50.82 (a) and (b), subpart G of part 53,
70.38(d), or 72.54 of this chapter, and
specifying that the licensee proposes to
decommission by use of alternate
criteria. The licensee shall document in
the decommissioning plan or LTP how
the advice of individuals and
institutions in the community who may
be affected by the decommissioning has
been sought and addressed, as
appropriate, following analysis of that
advice. In seeking such advice, the
licensee shall provide for:
*
*
*
*
*
§ 20.1406
[Amended]
59. In § 20.1406, in paragraphs (a) and
(b), wherever it appears, remove the
phrase ‘‘under part 52’’ and add in its
place the phrase ‘‘under parts 52 or 53’’.
■ 60. In § 20.1501, revise paragraph (b)
to read as follows:
■
§ 20.1501
General.
*
*
*
*
*
(b) Notwithstanding § 20.2103(a) of
this part, records from surveys
describing the location and amount of
subsurface residual radioactivity
identified at the site must be kept with
records important for decommissioning,
and such records must be retained in
accordance with § 30.35(g), § 40.36(f),
§ 50.75(g), subpart G of part 53,
§ 70.25(g), or § 72.30(d) of this chapter,
as applicable.
*
*
*
*
*
§ 20.1905
[Amended]
61. In § 20.1905, in paragraph (g)
introductory text, remove the phrase
‘‘Parts 50 or 52’’ and add in its place the
phrase ‘‘parts 50, 52, or 53’’.
■ 62. In § 20.2004, revise paragraph
(b)(1) to read as follows:
■
§ 20.2004 Treatment or disposal by
incineration.
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*
18:06 Oct 30, 2024
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(g) or 53.1640(b), (c), (d), and (e) of this
chapter, and must include the
information required by paragraph (b) of
this section. Occurrences reported
under §§ 50.73 or 53.1640 of this
chapter need not be reported by a
duplicate report under paragraph (a) of
this section.
*
*
*
*
*
§ 20.2206
[Amended]
66. In § 20.2206, in paragraph (a)(1),
remove the phrase ‘‘or § 50.22’’ and add
in its place the phrase ‘‘, § 50.22, or part
53’’.
■
PART 21—REPORTING OF DEFECTS
AND NONCOMPLIANCE
67. The authority citation for part 21
continues to read as follows:
§ 20.2201 Reports of theft or loss of
licensed material.
■
(a) * * *
(2) * * *
(i) Licensees having an installed
Emergency Notification System shall
make the reports to the NRC Operations
Center under §§ 50.72 or 53.1630 of this
chapter, and
*
*
*
*
*
(b) * * *
(2) * * *
(i) For holders of an operating license
for a nuclear power plant, the events
included in paragraph (b) of this section
must be reported under the procedures
described in §§ 50.73(b), (c), (d), (e), and
(g) or 53.1640(b), (c), (d), and (e) of this
chapter and must include the
information required in paragraph (b)(1)
of this section, and
*
*
*
*
*
(c) A duplicate report is not required
under paragraph (b) of this section if the
licensee is also required to submit a
report pursuant to §§ 30.55(c), 37.57,
37.81, 40.64(c), 50.72, 50.73, 53.1630,
53.1640, 70.52, 73.27(b), 73.67(e)(3)(vii),
73.67(g)(3)(iii), 73.1205, or 150.19(c) of
this chapter.
*
*
*
*
*
Authority: Atomic Energy Act of 1954,
secs. 53, 63, 81, 103, 104, 161, 223, 234, 1701
(42 U.S.C. 2073, 2093, 2111, 2133, 2134,
2201, 2273, 2282, 2297f); Energy
Reorganization Act of 1974, secs. 201, 206
(42 U.S.C. 5841, 5846); Nuclear Waste Policy
Act of 1982, secs. 135, 141 (42 U.S.C. 10155,
10161); 44 U.S.C. 3504 note.
§ 20.2202
*
*
*
*
(b)(1) Waste oils (petroleum derived
or synthetic oils used principally as
lubricants, coolants, hydraulic or
insulating fluids, or metalworking oils)
that have been radioactively
contaminated in the course of the
operation or maintenance of a nuclear
power reactor licensed under parts 50 or
53 of this chapter may be incinerated on
the site where generated provided that
the total radioactive effluents from the
facility, including the effluents from
such incineration, conform to the
requirements of appendix I to part 50 or
§ 53.425(d) of this chapter and the
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effluent release limits contained in
applicable license conditions other than
effluent limits specifically related to
incineration of waste oil. The licensee
shall report any changes or additions to
the information supplied under
§§ 50.34, 50.34a, or under subpart H of
part 53 of this chapter associated with
this incineration pursuant to §§ 50.71 or
53.1620 of this chapter, as appropriate.
The licensee shall also follow the
procedures of §§ 50.59 or 53.1565 of this
chapter with respect to such changes to
the facility or procedures.
*
*
*
*
*
■ 63. In § 20.2201, revise paragraphs
(a)(2)(i), (b)(2)(i), and (c) to read as
follows:
[Amended]
64. In § 20.2202, in paragraph (d)(1),
remove the phrase ‘‘10 CFR 50.72’’ and
add in its place the phrase ‘‘§§ 50.72 or
53.1630 of this chapter;’’.
■ 65. In § 20.2203, revise paragraph (c)
to read as follows:
■
§ 20.2203 Reports of exposures, radiation
levels, and concentrations of radioactive
material exceeding the constraints or limits.
*
*
*
*
*
(c) For holders of an operating license
or a combined license for a nuclear
power plant, the occurrences included
in paragraph (a) of this section must be
reported under the procedures
described in §§ 50.73(b), (c), (d), (e), and
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68. In § 21.2, revise paragraphs (a)(2)
through (4), (b), and (c) to read as
follows:
■
§ 21.2
Scope.
(a) * * *
(1) * * *
(2) Each individual, corporation,
partnership, or other entity doing
business within the United States, and
each director and responsible officer of
such an organization, that constructs a
production or utilization facility
licensed for manufacture, construction,
or operation under parts 50, 52, or 53 of
this chapter, an ISFSI for the storage of
spent fuel licensed under part 72 of this
chapter, an MRS for the storage of spent
fuel or high-level radioactive waste
under part 72 of this chapter, or a
geologic repository for the disposal of
high-level radioactive waste under parts
60 or 63 of this chapter; or supplies
basic components for a facility or
activity licensed, other than for export,
under parts 30, 40, 50, 52, 53, 60, 61, 63,
70, 71, or 72 of this chapter;
(3) Each individual, corporation,
partnership, or other entity doing
business within the United States, and
each director and responsible officer of
such an organization, applying for a
design certification rule under parts 52
or 53 of this chapter; or supplying basic
components with respect to that design
certification, and each individual,
corporation, partnership, or other entity
doing business within the United States,
and each director and responsible
officer of such an organization, whose
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application for design certification has
been granted under parts 52 or 53 of this
chapter, or who has supplied or is
supplying basic components with
respect to that design certification;
(4) Each individual, corporation,
partnership, or other entity doing
business within the United States, and
each director and responsible officer of
such an organization, applying for or
holding a standard design approval
under parts 52 or 53 of this chapter; or
supplying basic components with
respect to a standard design approval
under parts 52 or 53 of this chapter;
(b) For persons licensed to construct
a facility under either a construction
permit issued under §§ 50.23 or 53.1333
of this chapter or a combined license
under parts 52 or 53 of this chapter (for
the period of construction until the date
that the Commission makes the finding
under §§ 52.103(g) or 53.1452(g) of this
chapter), or to manufacture a facility
under parts 52 or 53 of this chapter,
evaluation of potential defects and
failures to comply and reporting of
defects and failures to comply under
§§ 50.55(e) or 53.605 of this chapter
satisfies each person’s evaluation,
notification, and reporting obligation to
report defects and failures to comply
under this part and the responsibility of
individual directors and responsible
officers of these licensees to report
defects under Section 206 of the Energy
Reorganization Act of 1974.
(c) For persons licensed to operate a
nuclear power plant under part 50, part
52, or part 53 of this chapter, evaluation
of potential defects and appropriate
reporting of defects under §§ 50.72,
50.73, 53.1630, 53.1640, or 73.1200 and
73.1205 of this chapter, satisfies each
person’s evaluation, notification, and
reporting obligation to report defects
under this part, and the responsibility of
individual directors and responsible
officers of these licensees to report
defects under Section 206 of the Energy
Reorganization Act of 1974.
*
*
*
*
*
■ 69. In § 21.3, revise the definitions for
‘‘Basic component’’, ‘‘Commercial grade
item’’, ‘‘Critical characteristics’’,
‘‘Dedicating entity’’, ‘‘Dedication’’,
‘‘Defect’’, and ‘‘Substantial safety
hazard’’ to read as follows:
§ 21.3
Definitions.
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*
*
*
*
*
Basic component. (1)(i) When applied
to nuclear power plants licensed under
part 53 of this chapter, basic component
means a safety-related structure, system,
or component (SSC), or part thereof, and
when applied to nuclear power plants
licensed under parts 50 or 52, of this
chapter, basic component means an
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SSC, or part thereof that affects its safety
function necessary to assure:
(A) The integrity of the reactor coolant
pressure boundary;
(B) The capability to shut down the
reactor and maintain it in a safeshutdown condition; or
(C) The capability to prevent or
mitigate the consequences of accidents
which could result in potential offsite
exposures comparable to those referred
to in §§ 50.34(a)(1), 50.67(b)(2), or
100.11 of this chapter, as applicable.
(ii) Basic components are items
designed and manufactured under a
quality assurance program complying
with appendix B to part 50 of this
chapter, or commercial grade items
which have successfully completed the
dedication process.
(2) When applied to standard design
certifications and approvals under part
53 of this chapter, basic component
means the design or procurement
information approved or to be approved
within the scope of the design
certification or approval for a safetyrelated SSC, or part thereof. When
applied to standard design certifications
under subpart B of part 52 of this
chapter and standard design approvals
under part 52 of this chapter, basic
component means the design or
procurement information approved or to
be approved within the scope of the
design certification or approval for an
SSC, or part thereof, that affects its
safety function necessary to assure:
(i) The integrity of the reactor coolant
pressure boundary;
(ii) The capability to shut down the
reactor and maintain it in a safeshutdown condition; or
(iii) The capability to prevent or
mitigate the consequences of accidents
which could result in potential offsite
exposures comparable to those referred
to in §§ 50.34(a)(1), 50.67(b)(2), or
100.11 of this chapter, as applicable.
(3) When applied to other facilities
and other activities licensed under 10
CFR parts 30, 40, 50 (other than nuclear
power plants), 60, 61, 63, 70, 71, or 72
of this chapter, basic component means
a structure, system, or component, or
part thereof, that affects their safety
function, that is directly procured by the
licensee of a facility or activity subject
to the regulations in this part and in
which a defect or failure to comply with
any applicable regulation in this
chapter, order, or license issued by the
Commission could create a substantial
safety hazard.
(4) In all cases, basic component
includes safety-related design, analysis,
inspection, testing, fabrication,
replacement of parts, or consulting
services that are associated with the
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87025
component hardware, design
certification, design approval, or
information in support of an early site
permit application under part 52 or part
53 of this chapter, whether these
services are performed by the
component supplier or others.
Commercial grade item. (1) When
applied to nuclear power plants
licensed under parts 50 or 53 of this
chapter, commercial grade item means
an SSC, or part thereof that affects its
safety function, that was not designed
and manufactured as a basic
component. Commercial grade items do
not include items where the design and
manufacturing process require inprocess inspections and verifications to
ensure that defects or failures to comply
are identified and corrected (i.e., one or
more critical characteristics of the item
cannot be verified).
(2) When applied to facilities and
activities licensed pursuant to parts 30,
40, 50 (other than nuclear power
plants), 60, 61, 63, 70, 71, or 72 of this
chapter, commercial grade item means
an item that is:
(i) Not subject to design or
specification requirements that are
unique to those facilities or activities;
(ii) Used in applications other than
those facilities or activities; and
(iii) To be ordered from the
manufacturer/supplier on the basis of
specifications set forth in the
manufacturer’s published product
description (for example, a catalog).
*
*
*
*
*
Critical characteristics. When applied
to nuclear power plants licensed under
parts 50, 52, or 53 of this chapter,
critical characteristics are those
important design, material, and
performance characteristics of a
commercial grade item that, once
verified, will provide reasonable
assurance that the item will perform its
intended safety function.
Dedicating entity. When applied to
nuclear power plants licensed under
parts 50, 52, or 53 of this chapter,
dedicating entity means the
organization that performs the
dedication process. Dedication may be
performed by the manufacturer of the
item, a third-party dedicating entity, or
the licensee itself. The dedicating entity,
under § 21.21(c) of this part, is
responsible for identifying and
evaluating deviations, reporting defects
and failures to comply for the dedicated
item, and maintaining auditable records
of the dedication process.
Dedication. (1) When applied to
nuclear power plants licensed pursuant
to 10 CFR parts 30, 40, 50, 53, or 60,
dedication is an acceptance process
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undertaken to provide reasonable
assurance that a commercial grade item
to be used as a basic component will
perform its intended safety function
and, in this respect, is deemed
equivalent to an item designed and
manufactured under a 10 CFR part 50,
appendix B, quality assurance program.
This assurance is achieved by
identifying the critical characteristics of
the item and verifying their
acceptability by inspections, tests, or
analyses performed by the purchaser or
third-party dedicating entity after
delivery, supplemented as necessary by
one or more of the following:
commercial grade surveys; product
inspections or witness at holdpoints at
the manufacturer’s facility, and analysis
of historical records for acceptable
performance. In all cases, the dedication
process must be conducted under the
applicable provisions of 10 CFR part 50,
appendix B. The process is considered
complete when the item is designated
for use as a basic component.
(2) When applied to facilities and
activities licensed pursuant to 10 CFR
parts 30, 40, 50 (other than nuclear
power plants), 60, 61, 63, 70, 71, or 72,
dedication occurs after receipt when
that item is designated for use as a basic
component.
Defect means:
(1) A deviation in a basic component
delivered to a purchaser for use in a
facility or an activity subject to the
regulations in this part if, on the basis
of an evaluation, the deviation could
create a substantial safety hazard;
(2) The installation, use, or operation
of a basic component containing a
defect as defined in this section;
(3) A deviation in a portion of a
facility subject to the early site permit,
standard design certification, standard
design approval, construction permit,
combined license or manufacturing
licensing requirements of parts 50, 52,
or 53 of this chapter, provided the
deviation could, on the basis of an
evaluation, create a substantial safety
hazard and the portion of the facility
containing the deviation has been
offered to the purchaser for acceptance;
(4) A condition or circumstance
involving a basic component that could
contribute to the exceeding of a safety
limit, as defined in the technical
specifications of a license for operation
issued under part 50, part 52, or part 53
of this chapter; or
(5) An error, omission or other
circumstance in a design certification,
or standard design approval that, on the
basis of an evaluation, could create a
substantial safety hazard.
*
*
*
*
*
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Substantial safety hazard means a
loss of safety function to the extent that
there is a major reduction in the degree
of protection provided to public health
and safety for any facility or activity
licensed or otherwise approved or
regulated by the NRC, other than for
export, under part 30, 40, 50, 52, 53, 60,
61, 63, 70, 71, or 72 of this chapter.
*
*
*
*
*
§ 25.35
§ 21.21
Authority: Atomic Energy Act of 1954,
secs. 53, 103, 104, 107, 161, 223, 234, 1701
(42 U.S.C. 2073, 2133, 2134, 2137, 2201,
2273, 2282, 2297f); Energy Reorganization
Act of 1974, secs. 201, 202 (42 U.S.C. 5841,
5842); 44 U.S.C. 3504 note.
[Amended]
70. Amend § 21.21 by:
■ a. In paragraph (a)(3), removing the
phrase ‘‘under part 52’’ and add in its
place the phrase ‘‘under parts 52 or 53’’;
and
■ b. In paragraphs (d)(1)(i) and (ii)
removing the phrase ‘‘parts 30, 40, 50,
52, 60, 61, 63, 70, 71, or 72 of this
chapter’’ and adding in its place the
phrase ‘‘parts 30, 40, 50, 52, 53, 60, 61,
63, 70, 71, or 72 of this chapter’’.
■
§ 21.51
[Amended]
71. In § 21.51, in paragraphs (a)(4) and
(5) remove the phrase ‘‘of part 52’’ and
add in its place the phrase ‘‘of part 52
or under subpart H of part 53’’.
■
§ 21.61
[Amended]
72. In § 21.61, in paragraph (b) remove
the phrase ‘‘under part 52’’ and add in
its place the phrase ‘‘under parts 52 or
53’’.
■
PART 25—ACCESS AUTHORIZATION
73. The authority citation for part 25
continues to read as follows:
■
Authority: Atomic Energy Act of 1954,
secs. 145, 161, 223, 234 (42 U.S.C. 2165,
2201, 2273, 2282); Energy Reorganization Act
of 1974, sec. 201 (42 U.S.C. 5841); 44 U.S.C.
3504 note; E.O. 10865, 25 FR 1583, as
amended, 3 CFR, 1959–1963 Comp., p. 398;
E.O. 12829, 58 FR 3479, 3 CFR, 1993 Comp.,
p. 570; E.O. 13526, 75 FR 707, 3 CFR, 2009
Comp., p. 298; E.O. 12968, 60 FR 40245, 3
CFR, 1995 Comp., p. 391. Section 25.17(f)
and Appendix A also issued under 31 U.S.C.
9701; 42 U.S.C. 2214.
74. In § 25.5, revise the definition for
‘‘License’’ to read as follows:
■
§ 25.5
*
*
*
*
License means a license issued
pursuant to 10 CFR parts 50, 52, 53, 60,
63, 70, or 72.
*
*
*
*
*
[Amended]
75. In § 25.17, in paragraph (a),
remove the phrase ‘‘under 10 CFR parts
50, 52, 54, 60, 63, 70, 72, or 76’’ and add
in its place the phrase ‘‘under 10 CFR
parts 50, 52, 53, 54, 60, 63, 70, 72, or
76’’.
■
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PART 26—FITNESS FOR DUTY
PROGRAMS
77. The authority citation for part 26
continues to read as follows:
■
78. In § 26.3, revise paragraph (d) and
add paragraph (f) to read as follows:
■
§ 26.3
Scope.
*
*
*
*
*
(d) Contractor/vendors (C/Vs) who
implement FFD programs or program
elements, to the extent that the licensees
and other entities specified in
paragraphs (a) through (c) and (f) of this
section rely on those C/V FFD programs
or program elements to meet the
requirements of this part, shall comply
with the requirements of this part.
*
*
*
*
*
(f) No later than the start of
construction activities, licensees and
other entities that have applied for or
have been issued a license under part 53
of this chapter, other than a
manufacturing license (ML), must
implement the requirements in subpart
M of this part or all the requirements of
this part except subparts K and M.
Holders of an ML under part 53 of this
chapter must implement the
requirements in subpart M or all the
requirements of this part except
subparts K and M, before commencing
activities that assemble a manufactured
reactor.
■ 79. In § 26.4, revise paragraphs (a)
introductory text, (a)(1), (a)(4), (b), (c),
(e) introductory text, (e)(4), (f), (g)
introductory text, and (h) to read as
follows:
§ 26.4 FFD program applicability to
categories of individuals.
Definitions.
*
§ 25.17
[Amended]
76. In § 25.35, in paragraph (a),
wherever it appears, remove the phrase
‘‘under part 52’’ and add in its place the
phrase ‘‘under parts 52 or 53’’.
■
Sfmt 4702
(a) All persons who are granted
unescorted access to nuclear power
reactor protected areas by the licensees
in § 26.3(a) and, as applicable, (c) and
perform the following duties shall be
subject to an FFD program that meets all
of the requirements of this part, except
subpart K of this part, and those persons
who are granted unescorted access to
either nuclear power reactor protected
areas or remote facilities where safetysignificant systems or components may
be operated within the design basis of
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a licensed commercial nuclear plant, by
the licensees and other entities in
§ 26.3(f) and perform the following
duties must be subject to an FFD
program that satisfies the requirements
in subpart M of this part, unless the
licensee or other entity subjects these
individuals to an FFD program that
satisfies all of the requirements of this
part except for those requirements in
subparts K and M:
(1) For persons who are granted
unescorted access by the licensees in
§ 26.3(a) and, as applicable, (c),
operating or onsite directing of the
operation of systems and components
that a risk-informed evaluation process
has shown to be significant to public
health and safety; for those persons who
are granted unescorted access by the
licensees and other entities in § 26.3(f),
operating or directing of the operation of
systems and components that a riskinformed evaluation process has shown
to be significant to public health and
safety;
*
*
*
*
*
(4) For persons who are granted
unescorted access to nuclear power
reactor protected areas by the licensees
in § 26.3(a) and, as applicable, (c),
performing maintenance or onsite
directing of the maintenance of SSCs
that a risk-informed evaluation process
has shown to be significant to public
health and safety; for those persons who
are granted unescorted access to nuclear
power reactor protected areas by the
licensees and other entities in § 26.3(f),
performing maintenance or directing of
the maintenance of SSCs that a riskinformed evaluation process has shown
to be significant to public health and
safety; and
*
*
*
*
*
(b) All persons who are granted
unescorted access to nuclear power
reactor protected areas by the licensees
in § 26.3(a) and, as applicable, (c) and
who do not perform the duties
described in paragraph (a) of this
section shall be subject to an FFD
program that meets all of the
requirements of this part, except
§§ 26.205 through 26.209 and subpart K
of this part. All persons who are granted
unescorted access to a facility licensed
under part 53 of this chapter, and who
do not perform or direct the
performance of the duties described in
§ 26.4(a), must be subject to the
requirements in subpart M of this part,
unless the licensee or other entity
implements an FFD program that
satisfies all of the requirements of this
part, except §§ 26.205 through 26.209
and subparts K and M.
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(c) All persons who are required by a
licensee in § 26.3(a) and, as applicable,
(c) to physically report to the licensee’s
Technical Support Center or Emergency
Operations Facility by licensee
emergency plans and procedures shall
be subject to an FFD program that meets
all of the requirements of this part,
except §§ 26.205 through 26.209 and
subpart K of this part. Also, for licensees
or other entities in § 26.3(f), all persons
without unescorted access to the facility
who make decisions and/or direct
actions regarding plant safety and
security, and all persons who
participate remotely in emergency
response activities or physically report
to the Technical Support Center or
Emergency Operations Facility (or an
equivalent facility), must be subject to
an FFD program that satisfies all of the
requirements described in subpart M of
this part, unless the licensee or other
entity implements an FFD program that
satisfies all of the requirements of this
part, except §§ 26.205 through 26.209
and subparts K and M.
*
*
*
*
*
(e) When construction activities, as
defined in § 26.5, begin, any individual
whose duties for the licensees and other
entities in § 26.3(c) require him or her
to have the following types of access or
perform the following activities at the
location where the nuclear power plant
will be constructed and operated shall
be subject to an FFD program that meets
all of the requirements of this part,
except subparts I, K, and M of this part,
and for any individual whose duties for
the licensees and other entities in
§ 26.3(f) require him or her to have the
following types of access, perform
construction activities as defined in
§ 26.5, or perform the following
activities must be subject to an FFD
program as described in subpart M or an
FFD program that satisfies all of the
requirements of this part, except
subparts I, K, and M:
*
*
*
*
*
(4) Witnesses or determines
inspections, tests, and analyses
certification required under part 52 or
part 53 of this chapter;
*
*
*
*
*
(f) Any individual who is constructing
or directing the construction of safetyor security-related SSCs shall be subject
to an FFD program that meets the
requirements of subpart K, or, if
applicable, subpart M of this part,
unless the licensee or other entity
subjects these individuals to an FFD
program that meets all of the
requirements of this part, except for
subparts I, K, and M of this part.
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87027
(g) All FFD program personnel who
are involved in the day-to-day
operations of the program, as defined by
the procedures of the licensees and
other entities in § 26.3(a) through (c),
and, as applicable, (d) and whose duties
require them to have the following types
of access or perform the following
activities shall be subject to an FFD
program that meets all of the
requirements of this part, except
subparts I, K, and M of this part, and,
at the licensee’s or other entity’s
discretion, subpart C of this part. All
personnel whose duties require them to
have the following types of access or
perform the following activities at
facilities licensed under part 53 of this
chapter must be subject to the
requirements in subpart M or an FFD
program that satisfies all of the
requirements of this part, except
subparts I, K, and M, and, at the
licensee’s or other entity’s discretion,
subpart C of this part:
*
*
*
*
*
(h) Individuals who have applied for
authorization to have the types of access
or perform the activities described in
paragraphs (a) through (d) of this section
shall be subject to §§ 26.31(c)(1),
26.35(b), 26.37, 26.39, and the
applicable requirements of subparts C, E
through H, and M of this part.
*
*
*
*
*
■ 80. Amend § 26.5 by:
■ a. Adding the definitions for
‘‘Biological marker’’ and ‘‘Change’’;
■ b. Revising the definitions for
‘‘Constructing or construction
activities’’ ‘‘Contractor/vendor (C/V)’’;
■ c. Adding the definition of ‘‘Illicit
substance’’;
■ d. Revising the definitions of ‘‘Other
entity’’ and ‘‘Questionable validity’’;
■ e. Adding the definitions of
‘‘Reduction in FFD program
effectiveness’’;
■ f. Revising the definitions of
‘‘Reviewing official’’, ‘‘Safety-related
structures, systems, and components
(SSCs)’’, and ‘‘Security-related SSCs’’;
■ g. Adding the definitions of ‘‘Special
nuclear material’’; and
■ h. Revising the definition of ‘‘Unit
outage’’.
The additions and revisions read as
follows:
§ 26.5
Definitions.
*
*
*
*
*
Biological marker means, for a part 53
licensee implementing subpart M of this
part, an endogenous substance that is
used to validate that the biological
specimen collected for testing was
produced by the donor.
*
*
*
*
*
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Change as used in § 26.603(e) means
an action that results in a modification
of, addition to, or removal from the
licensee’s or other entity’s FFD program.
*
*
*
*
*
Constructing or construction activities
means, for the purposes of this part, the
tasks involved in building a nuclear
power plant that are performed at the
location where the nuclear power plant
will be constructed and operated. These
tasks include fabricating, erecting,
integrating, and testing safety- and
security-related SSCs, and the
installation of their foundations,
including the placement of concrete. For
a licensee or other entity described in
§ 26.3(f), construction is defined in
§ 53.020 of this chapter.
Contractor/vendor (C/V) means any
company, or any individual not
employed by a licensee or other entity
specified in § 26.3(a) through (c) and (f),
who is providing work or services to a
licensee or other entity covered in
§ 26.3(a) through (c) and (f), either by
contract, purchase order, oral
agreement, or other arrangement.
*
*
*
*
*
Illicit substance means a substance
that causes impairment and possible
addiction but is not an illegal drug as
defined in § 26.5.
*
*
*
*
*
Other entity means any corporation,
firm, partnership, limited liability
company, association, C/V, or other
organization who is subject to this part
under § 26.3(a) through (c) and (f) but is
not licensed by the NRC.
*
*
*
*
*
Questionable validity means the
results of validity screening or initial
validity tests at a licensee testing facility
indicating that a urine specimen may be
adulterated, substituted, dilute, or
invalid. For a part 53 licensee or other
entity, questionable validity means the
results of validity screening or initial
validity tests indicating that a biological
specimen obtained from an individual
pursuant to subpart M of this part may
be adulterated, substituted, dilute, or
invalid.
Reduction in FFD program
effectiveness means, for a part 53
licensee or other entity implementing
subpart M of this part, a change or series
of changes to an element of the FFD
program that reduces or eliminates the
licensee’s ability to satisfy or maintain
site-specific FFD program performance
when compared to historical sitespecific performance, the licensee’s
fleet-level program performance, or
industry performance.
*
*
*
*
*
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Reviewing official means an employee
of a licensee or other entity specified in
§ 26.3(a) through (c) and (f), who is
designated by the licensee or other
entity to be responsible for reviewing
and evaluating any potentially
disqualifying FFD information about an
individual, including, but not limited
to, the results of a determination of
fitness, as defined in § 26.189, in order
to determine whether the individual
may be granted or maintain
authorization.
Safety-related structures, systems, and
components (SSCs) means, for part 50 or
part 52 licensees and other entities
described in § 26.3(a) through (d), those
SSCs that are relied on to remain
functional during and following design
basis events to ensure the integrity of
the reactor coolant pressure boundary,
the capability to shut down the reactor
and maintain it in a safe shutdown
condition, or the capability to prevent or
mitigate the consequences of accidents
that could result in potential offsite
exposure comparable to the guidelines
in § 50.34(a)(1) of this chapter. For part
53 licensees and other entities described
in § 26.3(d) and (f), safety-related has
the same meaning as that in § 53.020 of
this chapter.
Security-related SSCs means, for the
purposes of this part, those structures,
systems, and components that the
licensee will rely on to implement the
licensee’s physical security and
safeguards contingency plans that either
are required under part 73 of this
chapter if the licensee is a construction
permit applicant or holder or an early
site permit holder, as described in
§ 26.3(c)(3) through (c)(5), respectively,
or are included in the licensee’s
application if the licensee is a combined
license applicant or holder, as described
in § 26.3(c)(1) and (c)(2), respectively, or
a licensee or other entity described in
§ 26.3(d) or (f).
*
*
*
*
*
Special nuclear material (SNM) has
the same meaning as that in § 70.4 of
this chapter.
*
*
*
*
*
Unit outage means, for the purposes
of this part, for electricity-generation
units, that the reactor unit is
disconnected from the electrical grid.
Unit outage means, for the purposes of
this part, for non-electricity-generation
units, that the reactor unit is
disconnected from the loads to which
its output is supplied under normal
operating conditions.
*
*
*
*
*
■ 81. In § 26.8, revise paragraph (b) to
read as follows:
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§ 26.8 Information collection
requirements: OMB approval.
*
*
*
*
*
(b) The approved information
collection requirements contained in
this part appear in §§ 26.9, 26.27, 26.29,
26.31, 26.33, 26.35, 26.37, 26.39, 26.41,
26.53, 26.55, 26.57, 26.59, 26.61, 26.63,
26.65, 26.67, 26.69, 26.75, 26.77, 26.85,
26.87, 26.89, 26.91, 26.93, 26.95, 26.97,
26.99, 26.101, 26.103, 26.107, 26.109,
26.111, 26.113, 26.115, 26.117, 26.119,
26.125, 26.127, 26.129, 26.135, 26.137,
26.139, 26.153, 26.157, 26.159, 26.163,
26.165, 26.167, 26.168, 26.169, 26.183,
26.185, 26.187, 26.189, 26.202, 26.203,
26.205, 26.207, 26.211, 26.401, 26.403,
26.405, 26.406, 26.407, 26.411, 26.413,
26.415, 26.417, 26.603, 26.604, 26.605,
26.606, 26.607, 26.608, 26.609, 26.611,
26.613, 26.617, 26.619, 26.711, 26.713,
26.715, 26.717, 26.719, and 26.821.
■ 82. Revise § 26.21 to read as follows:
§ 26.21
Fitness-for-duty program.
The licensees and other entities
specified in § 26.3(a) through (c) and (f)
(for those licensees and other entities
that do not implement the requirements
in subparts M and K of this part) shall
establish, implement, and maintain FFD
programs that, at a minimum, comprise
the program elements contained in this
subpart. The individuals specified in
§ 26.4(a) through (e) and (g), and, at the
licensee’s or other entity’s discretion,
§ 26.4(f), and, if necessary, § 26.4(j) shall
be subject to these FFD programs.
Licensees and other entities may rely on
the FFD program or program elements of
a C/V, as defined in § 26.5, if the C/V’s
FFD program or program elements
satisfy the applicable requirements of
this part.
■ 83. Revise § 26.51 to read as follows:
§ 26.51
Applicability.
The requirements in this subpart
apply to the licensees and other entities
identified in § 26.3(a), (b), and, as
applicable, (c) for the categories of
individuals in § 26.4(a) through (d), and,
at the licensee’s or other entity’s
discretion, in § 26.4(g) and, if necessary,
§ 26.4(j). The requirements in this
subpart also apply to the licensees and
other entities specified in § 26.3(c), as
applicable, for the categories of
individuals in § 26.4(e). At the
discretion of a licensee or other entity
in § 26.3(c), the requirements of this
subpart also may be applied to the
categories of individuals identified in
§ 26.4(f). In addition, the requirements
in this subpart apply to the entities in
§ 26.3(d) to the extent that a licensee or
other entity relies on the C/V to satisfy
the requirements of this subpart. Certain
requirements in this subpart also apply
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to the individuals specified in § 26.4(h).
The requirements in this subpart apply
to the FFD programs of licensees and
other entities identified in § 26.3(f) that
elect not to implement the requirements
in subpart M for the categories of
individuals in § 26.4 and those licensees
and other entities that elect to
implement the requirements in § 26.605.
§ 26.53
[Amended]
84. Amend § 26.53 by:
a. In paragraph (e), wherever it
appears, remove the phrase ‘‘§ 26.3(a)
through (c)’’ and add in its place the
phrase ‘‘§ 26.3(a) through (c) and (f)’’;
and
■ b. In paragraphs (g), (h), and (i),
wherever it appears, remove the phrase
‘‘(c) and (d)’’ and add in its place the
phrase ‘‘(c), (d), and (f)’’.
■
■
§ 26.63
[Amended]
85. In § 26.63, in paragraph (d) remove
the phrase ‘‘§ 26.3(a) through (d)’’ and
add in its place the phrase ‘‘§ 26.3(a)
through (d) and (f)’’.
■ 86. Revise § 26.73 to read as follows:
■
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§ 26.73
Applicability.
The requirements in this subpart
apply to the licensees and other entities
identified in § 26.3(a), (b), and, as
applicable, (c) for the categories of
individuals specified in § 26.4(a)
through (d) and (g). The requirements in
this subpart also apply to the licensees
and other entities specified in § 26.3(c),
as applicable, for the categories of
individuals in § 26.4(e). At the
discretion of a licensee or other entity
in § 26.3(c), the requirements of this
subpart also may be applied to the
categories of individuals identified in
§ 26.4(f). In addition, the requirements
in this subpart apply to the entities in
§ 26.3(d) to the extent that a licensee or
other entity relies on the C/V to satisfy
the requirements of this subpart. The
regulations in this subpart also apply to
the individuals specified in § 26.4(h)
and (j), as appropriate. The
requirements in this subpart apply to
the FFD programs of licensees and other
entities identified in § 26.3(f) that elect
not to implement the requirements in
subpart M for the categories of
individuals in § 26.4 and those licensees
and other entities that elect to
implement the requirements in
§ 26.605(b).
■ 87. Revise § 26.81 to read as follows:
§ 26.81
Purpose and applicability.
This subpart contains requirements
for collecting specimens for drug testing
and conducting alcohol tests by or on
behalf of the licensees and other entities
in § 26.3(a) through (d) for the categories
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of individuals specified in § 26.4(a)
through (d) and (g). At the discretion of
a licensee or other entity in § 26.3(c),
specimen collections and alcohol tests
must be conducted either under this
subpart for the individuals specified in
§ 26.4(e) and (f) or the licensee or other
entity may rely on specimen collections
and alcohol tests conducted under the
requirements of 49 CFR part 40 for the
individuals specified in § 26.4(e) and (f).
The requirements of this subpart do not
apply to specimen collections and
alcohol tests that are conducted under
the requirements of 49 CFR part 40, as
permitted in this paragraph and under
§§ 26.4(j) and 26.31(b)(2) and subpart K.
The requirements in this subpart apply
to the FFD programs of licensees and
other entities identified in § 26.3(f) that
elect not to implement the requirements
in subpart M for the categories of
individuals in § 26.4 and those licensees
and other entities that elect to
implement the requirements in § 26.605.
■ 88. Revise § 26.201 to read as follows:
§ 26.201
Applicability.
(a) The requirements in this subpart,
with the exception of § 26.202, apply to
the licensees and other entities
identified in § 26.3(a); if applicable, (c),
(d), and (f), for licensees and other
entities not implementing the
requirements in subparts K and M. For
the licensees and other entities to whom
the requirements in this subpart, with
the exception of § 26.202, apply, the
requirements in §§ 26.203 and 26.211
apply to the individuals identified in
§ 26.4(a) through (c). In addition, the
requirements in §§ 26.205 through
26.209 apply to the individuals
identified in § 26.4(a).
(b) The requirements in this subpart,
with the exception of § 26.203, apply to
the licensees or other entities identified
in § 26.3(f) implementing this subpart
under §§ 26.604 and 26.605. For these
licensees and other entities, the
requirements in §§ 26.202 and 26.211
apply to the individuals identified in
§ 26.4(a) through (c) and any person
licensed to operate under 10 CFR part
53; and the requirements in §§ 26.205
through 26.209 apply to the individuals
identified in § 26.4(a).
■ 89. Add § 26.202 to read as follows:
§ 26.202 General provisions for facilities
licensed under part 53.
(a) Policy. Licensees must establish a
policy for the management of fatigue for
all individuals who are subject to the
licensee’s FFD program and incorporate
it into the written policy required in
§ 26.606(a).
(b) Procedures. In addition to the
procedures required in § 26.606(b),
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licensees must develop, implement, and
maintain procedures that—
(1) Describe the process to be
followed when any individual
identified in § 26.4(a) through (c) makes
a self-declaration that he or she is not
fit to safely and competently perform
his or her duties for any part of a
working tour as a result of fatigue. The
procedure must—
(i) Describe the individual’s and
licensee’s rights and responsibilities
related to self-declaration;
(ii) Describe requirements for
establishing controls and conditions
under which an individual may be
permitted or required to perform work
after that individual declares that he or
she is not fit due to fatigue; and
(iii) Describe the process to be
followed if the individual disagrees
with the results of a fatigue assessment
that is required under § 26.211(a)(2);
(2) Describe the process for
implementing the controls required
under § 26.205 for the individuals who
are performing the duties listed in
§ 26.4(a);
(3) Describe the process to be
followed in conducting fatigue
assessments under § 26.211; and
(4) Describe the disciplinary actions
that the licensee may impose on an
individual following a fatigue
assessment, and the conditions and
considerations for taking those
disciplinary actions.
(c) Training and assessments.
Licensees must include the following
KAs in the content of the training and
trainee assessments required in
§ 26.608:
(1) Knowledge of the contributors to
worker fatigue, circadian variations in
alertness and performance, indications
and risk factors for common sleep
disorders, shiftwork strategies for
obtaining adequate rest, and the
effective use of fatigue countermeasures;
and
(2) Ability to identify symptoms of
worker fatigue and contributors to
decreased alertness in the workplace.
(d) Recordkeeping. Licensees must
retain the following records for at least
3 years or until the completion of all
related legal proceedings, whichever is
later:
(1) Records of work hours for
individuals who are subject to the work
hour controls in § 26.205;
(2) For licensees implementing the
requirements of § 26.205(d)(3), records
of shift schedules and shift cycles, or,
for licensees implementing the
requirements of § 26.205(d)(7), records
of shift schedules and records showing
the beginning and end times and dates
of all averaging periods, of individuals
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who are subject to the work hour
controls in § 26.205;
(3) The documentation of waivers that
is required in § 26.207(a)(4), including
the bases for granting the waivers;
(4) The documentation of work hour
reviews that is required in § 26.205(e)(3)
and (e)(4); and
(5) The documentation of fatigue
assessments that is required in
§ 26.211(g).
(e) Reporting. Licensees must include
the following information in a standard
format in the annual FFD program
performance report required under
§ 26.617(b)(2):
(1) A summary for each nuclear power
plant site of all instances during the
previous calendar year when the
licensee waived one or more of the work
hour controls specified in § 26.205(d)(1)
through (d)(5)(i) and (d)(7) for
individuals described in § 26.4(a). The
summary must include only those
waivers under which work was
performed. If it was necessary to waive
more than one work hour control during
any single extended work period, the
summary of instances must include
each of the work hour controls that were
waived during the period. For each
category of individuals specified in
§ 26.4(a), the licensee must report—
(i) The number of instances when
each applicable work hour control
specified in § 26.205(d)(1)(i) through
(iii), (d)(2)(i) and (ii), (d)(3)(i) through
(v), and (d)(7) was waived for
individuals not working on outage
activities;
(ii) The number of instances when
each applicable work hour control
specified in § 26.205(d)(1)(i) through
(iii), (d)(2)(i) and (ii), (d)(3)(i) through
(v), (d)(4) and (d)(5)(i), and (d)(7) was
waived for individuals working on
outage activities; and
(iii) A summary that shows the
distribution of waiver use among the
individuals applicable within each
category of individuals identified in
§ 26.4(a) (e.g., a table that shows the
number of individuals who received
only one waiver during the reporting
period, the number of individuals who
received a total of two waivers during
the reporting period).
(2) A summary of corrective actions,
if any, resulting from the analyses of
these data, including fatigue
assessments.
(f) Audits. Licensees must audit the
management of worker fatigue under
§ 26.615.
■ 90. In § 26.205, revise paragraphs
(d)(7)(iii) and (d)(8) to read as follows:
§ 26.205
*
*
Work Hours.
*
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*
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(d) * * *
(7) * * *
(iii) Each licensee shall state, in its
FFD policy and procedures required by
either §§ 26.27 and 26.203(a) and (b) or
§§ 26.202(a) and (b) and 26.606, the
work hour counting system in
§ 26.205(d)(7)(ii) the licensee is using.
(8) Each licensee shall state, in its
FFD policy and procedures required by
either §§ 26.27 and 26.203(a) and (b) or
§§ 26.202(a) and (b) and 26.606, the
requirements with which the licensee is
complying: the minimum days off
requirements in § 26.205(d)(3) or
maximum average work hours
requirements in § 26.205(d)(7).
*
*
*
*
*
■ 91. In § 26.207, revise paragraph
(a)(1)(ii) to read as follows:
§ 26.207
Waivers and exceptions.
(a) * * *
(1) * * *
(ii) A supervisor assesses the
individual face to face and determines
that there is reasonable assurance that
the individual will be able to safely and
competently perform his or her duties
during the additional work period for
which the waiver will be granted. The
supervisor performing the assessment
shall be trained as required by either
§§ 26.29 and 26.203(c) or §§ 26.202(c)
and 26.608 and shall be qualified to
direct the work to be performed by the
individual. If there is no supervisor on
site who is qualified to direct the work,
the assessment may be performed by a
supervisor who is qualified to provide
oversight of the work to be performed by
the individual. At a minimum, the
assessment must address the potential
for acute and cumulative fatigue
considering the individual’s work
history for at least the past 14 days, the
potential for circadian degradations in
alertness and performance considering
the time of day for which the waiver
will be granted, the potential for fatiguerelated degradations in alertness and
performance to affect risk-significant
functions, and whether any controls and
conditions must be established under
which the individual will be permitted
to perform work. For licensees and other
entities in § 26.3(f), the assessment may
be performed remotely using electronic
communications. In such instances, the
assessment must be supported by
someone who is present in-person with
the individual whose alertness may be
impaired, and that supporting person
must be trained under the requirements
of either § 26.29 and § 26.203(c) or
§ 26.202(c) and § 26.608.
*
*
*
*
*
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92. In § 26.211, revise paragraphs
(a)(1) and (3) and paragraph (b)
introductory text to read as follows:
■
§ 26.211
Fatigue assessments.
(a) * * *
(1) For-cause. In addition to any other
test or determination of fitness that may
be required under §§ 26.31(c), 26.77,
26.607(b), and 26.619, a fatigue
assessment must be conducted in
response to an observed condition of
impaired individual alertness creating a
reasonable suspicion that an individual
is not fit to safely and competently
perform his or her duties, except if the
condition is observed during an
individual’s break period. If the
observed condition is impaired alertness
with no other behaviors or physical
conditions creating a reasonable
suspicion of possible substance abuse,
then the licensee need only conduct a
fatigue assessment. If the licensee has
reason to believe that the observed
condition is not due to fatigue, the
licensee need not conduct a fatigue
assessment;
*
*
*
*
*
(3) Post-event. A fatigue assessment
must be conducted in response to events
requiring post-event drug and alcohol
testing as specified in § 26.31(c) or postevent tests in § 26.607(b)(4). Licensees
may not delay necessary medical
treatment in order to conduct a fatigue
assessment; and
*
*
*
*
*
(b) Only supervisors and FFD program
personnel who are trained under either
§§ 26.29 and 26.203(c) or §§ 26.202(c)
and 26.608 may conduct a fatigue
assessment. The fatigue assessment
must be conducted face to face with the
individual whose alertness may be
impaired. For licensees and other
entities in § 26.3(f), a fatigue assessment
may be performed remotely using
electronic communications. In such
instances, the fatigue assessment must
be supported by someone who is
present in-person with the individual
whose alertness may be impaired, and
that supporting person must be trained
in accordance with the requirements of
either §§ 26.29 and 26.203(c) or
§§ 26.202(c) and 26.608.
*
*
*
*
*
■ 93. Add Subpart M, consisting of
§§ 26.601 through 26.619, to read as
follows:
Subpart M—Fitness for Duty Programs
for Facilities Licensed Under 10 CFR
Part 53
Sec.
26.601
26.603
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General provisions.
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26.604 FFD program requirements for
facilities that satisfy the § 26.603(c)
criterion.
26.605 FFD program requirements for
facilities that do not implement § 26.604.
26.606 Written policy and procedures.
26.607 Drug and alcohol testing.
26.608 FFD program training.
26.609 Behavioral observation.
26.610 Sanctions.
26.611 Protection of information.
26.613 Appeals process.
26.615 Audits.
26.617 Recordkeeping and reporting.
26.619 Suitability and fitness
determinations.
§ 26.601
Applicability.
A licensee or other entity in § 26.3(f),
at its discretion, may establish,
implement, and maintain a fitness-forduty (FFD) program that satisfies the
requirements of this subpart for those
categories of individuals in § 26.4, as
applicable, and any person licensed to
operate under 10 CFR part 53. If a
licensee or other entity in § 26.3(f) does
not elect to implement an FFD program
that satisfies the requirements of this
subpart, then those categories of
individuals in § 26.4, as applicable, and
any person licensed to operate under 10
CFR part 53 must be subject to an FFD
program that satisfies all part 26
requirements, except for those
requirements in subparts K and M.
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§ 26.603
General provisions.
(a) FFD program description. An
applicant’s description of the FFD
program in its Final Safety Analysis
Report, required by subpart H of part 53
of this chapter, must include—
(1) If the applicant performed the
analysis under paragraph (c) of this
section, a summary of the analysis,
including the assumptions,
methodology, conclusion, and
references;
(2) A statement whether the FFD
program will be implemented pursuant
to § 26.604 or § 26.605, or will satisfy all
part 26 requirements, except for the
requirements in subparts K and M;
(3) A discussion of the applicability of
the FFD program to those individuals
described in § 26.4 and how the
program will be implemented offsite at
a U.S. Nuclear Regulatory Commission
(NRC)-licensed facility authorized to
assemble or test a manufactured reactor,
if applicable;
(4) A description of the drug and
alcohol testing and fitness
determination process to be
implemented through the licensee’s or
other entity’s procedures, including the
collection and testing facilities to be
used, biological specimens to be
collected, and sanctions to be imposed
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upon a confirmed FFD policy violation;
and
(5) A summary of the FFD
performance monitoring and review
program (PMRP), including the
measures and thresholds required by
paragraph (d)(1) of this section.
(b) FFD program implementation and
availability. For the licensees and other
entities in § 26.3(f), other than the
holder of a manufacturing license (ML),
the FFD program must be implemented
no later than the start of construction
activities, as defined in § 26.5, and
maintained until the NRC’s docketing of
the license holder’s certifications
described in § 53.1070 of this chapter.
For holders of an ML, the FFD program
must be implemented no later than the
start of activities that assemble the
manufactured reactor and maintained
until expiration of the ML.
(c) Criterion and analysis for an FFD
program. For a licensee or other entity
to implement an FFD program under
§ 26.604, the licensee or other entity
must perform a site-specific analysis to
demonstrate that the facility and its
operation satisfy the criterion in
§ 53.860(a)(2) of this chapter. The
licensee or other entity must maintain
the analysis, including updates to reflect
changes made to the staffing, FFD
programs, or offsite support resources
described in the analysis, to show that
the facility and its operation continue to
satisfy the criterion, until permanent
cessation of operations under § 53.1070
of this chapter.
(d) FFD performance monitoring and
review. A licensee or other entity must
establish performance measures and
associated thresholds as described in
paragraph (d)(1) of this section and
monitor the effectiveness of its FFD
program by comparing performance data
against these performance measures and
thresholds, in a manner sufficient to
satisfy the § 26.23 performance
objectives.
(1) PRMP elements. The PMRP must
be documented and maintained and
include the following program elements:
(i) Performance measures.
Performance measures must be
identified and designed to monitor FFD
program performance.
(A) If the licensee or other entity is
subject to the requirements in § 26.604,
then the monitoring program must
include performance measures for the
following: the behavioral observation
program; occurrence of FFD policy
violations categorized by licensee
employee, contractor/vendor, and labor
category; and occurrence of individuals
with potentially disqualifying
information or who possessed FFD
prohibited items.
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87031
(B) If the licensee or other entity is
subject to the requirements in § 26.604
and has implemented a drug testing
program at its discretion or is subject to
the requirements of § 26.605, then the
monitoring program must include
performance measures identified in
paragraph (d)(1)(i)(A) of this section.
This monitoring program must also
include performance measures for the
pre-access and random positive testing
rates, random testing rate for licensee
employees and contractor/vendors, and
the number of subversion attempts
categorized by licensee employee,
contractor/vendor, and labor category.
(ii) Thresholds. Licensee- or other
entity-specific thresholds for its sitespecific performance measures must be
established and used to facilitate
corrective actions to maintain FFD
program performance. Initial thresholds
must be based on FFD performance data
from comparable facilities subject to
part 26, the licensee’s or other entity’s
fleet-level program performance if
applicable, and industry FFD
performance data.
(iii) Monitoring program. Licensees
and other entities must monitor the
performance of their FFD programs
against licensee- or other entityestablished performance measures and
thresholds as FFD performance data is
received to determine whether a
threshold has been exceeded. Licensees
and other entities must perform year-toyear comparisons of site-specific
performance; site-specific performance
to the licensee’s or other entity’s fleetlevel program performance, if
applicable; and site-specific to industry
performance.
(iv) Quantitative and qualitative
reviews. The PMRP must include a
documented review of the elements in
paragraph (d)(1)(i) through (iii) of this
section and the following qualitative
elements.
(A) Worker protections. The review
must include a documented assessment
of the licensee’s or other entity’s
implementation of the protections
described in §§ 26.606(b)(1)(iii), 26.611,
and 26.613.
(B) Laboratory test results and
Medical Review Officer performance.
The review must include a documented
assessment of whether the actions taken
by the Medical Review Officer (MRO)
met the requirements in § 26.185 based
on the laboratory test results reported
under § 26.169. This review must
include a comparative analysis between
the point of collection testing and
assessment (POCTA) screening result(s)
and the corresponding specimen test
results obtained from the U.S.
Department of Health and Human
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Services (HHS)-certified laboratory if
the POCTA indicated a positive,
adulterated, substituted, or invalid
screening result or discrepant biological
marker, to assess the effectiveness of the
POCTA and to inform MRO decisions
under § 26.185 or § 26.607(m)(6).
(C) Change control. The review must
include a documented assessment of the
changes made under paragraph (e) of
this section to verify that the summation
of program changes has not resulted in
a reduction in FFD program
effectiveness.
(2) Corrective actions. Corrective
actions must be implemented to address
when FFD performance meets a
licensee-established performance
threshold or to resolve a finding
resulting from a qualitative review or
audit in a manner that restores
performance and corrects root causes,
contributing causes, or both.
(3) Program review periodicity. The
documented review in paragraph
(d)(1)(iv) of this section must be
conducted biennially to assess and
modify licensee or other entity
implementation of its FFD program.
This documented review must
demonstrate that the performance
measures and thresholds are appropriate
and adjusted as necessary based on sitelevel and licensee’s or other entity’s
fleet-level, if applicable, program
performance, and industry performance.
(i) Identified program weaknesses and
corrective actions must be summarized
in the annual reporting requirement
described in § 26.617(b)(2) or § 26.717,
as applicable.
(ii) The program review must be
completed and approved by the licensee
or other entity to support the reporting
of PMRP weaknesses and corrective
actions as required in paragraph (d)(3)(i)
of this section every odd-numbered
year, and the implementation of
corrective actions before May 15 of that
odd-numbered year.
(e) FFD program change control. (1)
The licensee or other entity may make
changes to its FFD program under this
subpart if—
(i) The licensee or other entity
performs and retains an analysis
demonstrating that the changes do not
reduce the effectiveness of the FFD
program; or
(ii) The change was necessitated or
justified by a change to part 26,
laboratory processes or procedures, or
guidance issued by the HHS or NRC, as
implemented by the licensee or other
entity though its procedures.
(2) A licensee or other entity desiring
to make a change that decreases FFD
program effectiveness must implement a
mitigating strategy so the FFD program,
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as revised, will continue to satisfy the
performance objectives in § 26.23 and
not result in a reduction in program
effectiveness.
(3) Except for phencyclidine, and
notwithstanding paragraph (e)(1)(ii) of
this section, the change control process
may not be used to reduce the minimum
panel of drugs to be tested in
§ 26.607(c)(1).
(4) The licensee must retain a record
of each change made under this section
for a period of at least 5 years from the
date the change was implemented and
summarize this change in its annual
FFD performance report required by
§ 26.617(b)(2) or § 26.717, as applicable.
§ 26.604 FFD program requirements for
facilities that satisfy the § 26.603(c)
criterion.
(a) FFD program. A licensee or other
entity with an analysis that
demonstrates that its facility and
operation satisfy the criterion in
§ 26.603(c) may elect to establish,
implement, and maintain an FFD
program under this section. That FFD
program must contain the following
elements:
(1) Applies to those individuals
described in § 26.4, as applicable; and
(2) Implements the following
requirements and subparts in this part:
(i) § 26.23, Performance objectives;
(ii) § 26.603, General provisions;
(iii) § 26.606, Written policies and
procedures, (a) and, if applicable (b);
(iv) § 26.608, FFD program training;
(v) § 26.609, Behavioral observation;
(vi) § 26.610, Sanctions;
(vii) § 26.611, Protection of
information;
(viii) § 26.613, Appeals process;
(ix) § 26.615, Audits;
(x) § 26.617, Recordkeeping and
reporting;
(xi) § 26.619, Suitability and fitness
determinations;
(xii) Subpart A—Administrative
Provisions;
(xiii) Subpart I—Managing Fatigue;
and
(xiv) Subpart O—Inspections,
Violations, and Penalties.
(b) [Reserved]
§ 26.605 FFD program requirements for
facilities that do not implement § 26.604.
(a) Licensees and other entities who
satisfy the criterion in § 26.603(c), at
their discretion, and licensees and other
entities who do not satisfy the criterion
in § 26.603(c), must establish,
implement, and maintain an FFD
program under this section either during
construction activities as defined in
§ 26.5, or during activities performed
under an ML that allows the assembly,
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testing, or both of a manufactured
reactor, as applicable. This FFD program
must contain the following elements:
(1) Applies to those individuals
described in § 26.4, as applicable; and,
(2) Implements the following
requirements and subparts in this part—
(i) § 26.23, Performance objectives;
(ii) § 26.603, General provisions;
(iii) § 26.606, Written policy and
procedures;
(iv) § 26.607, Drug and alcohol testing;
(v) § 26.608, FFD program training;
(vi) § 26.609, Behavioral observation;
(vii) § 26.610, Sanctions;
(viii) § 26.611, Protection of
information;
(ix) § 26.613, Appeals process;
(x) § 26.615, Audits;
(xi) § 26.617, Recordkeeping and
reporting;
(xii) § 26.619, Suitability and fitness
determinations;
(xiii) Subpart A—Administrative
Provisions;
(xiv) Subpart I—Managing Fatigue, in
the case of holders of an ML that allows
the assembly, testing, or both of a
manufactured reactor; and
(xv) Subpart O—Inspections,
Violations, and Penalties.
(b) Licensees and other entities who
satisfy the criterion in § 26.603(c), at
their discretion, and licensees and other
entities who do not satisfy the criterion
in § 26.603(c), before the loading of fuel
onsite into a reactor vessel; before
receiving a manufactured reactor; or
before individuals subject to part 26
operate, test, perform maintenance of, or
direct the maintenance or surveillance
of security-related equipment or
equipment that a risk-informed
evaluation process has shown to be
significant to public health and safety,
must establish, implement, and
maintain an FFD program that—
(1) Applies to those individuals
described in § 26.4, as applicable; and,
(2) Implements the following
requirements and subparts—
(i) § 26.23, Performance objectives;
(ii) § 26.603, General provisions;
(iii) § 26.606, Written policy and
procedures;
(iv) § 26.607, Drug and alcohol testing;
(v) § 26.608, FFD program training;
(vi) § 26.609, Behavioral observation;
(vii) § 26.611, Protection of
information;
(viii) § 26.613, Appeals process;
(ix) § 26.615, Audits;
(x) Subpart A—Administrative
Provisions;
(xi) Subpart C—Granting and
Maintaining Authorization;
(xii) Subpart D—Management Actions
and Sanctions to be Imposed;
(xiii) Subpart H—Determining
Fitness-for-Duty Policy Violations and
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Determining Fitness, unless using the
HHS Guidelines for MRO evaluation of
drug test results, and determining
fitness;
(xiv) Subpart I—Managing Fatigue;
(xv) Subpart N—Recordkeeping and
Reporting Requirements; and
(xvi) Subpart O—Inspections,
Violations, and Penalties.
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§ 26.606
Written policy and procedures.
(a) Licensees and other entities that
implement an FFD program under this
subpart must ensure that—
(1) A written FFD policy statement is
provided to each individual who is
subject to the program before the
individual is subject to behavioral
observation, drug and alcohol testing, or
both.
(2) The FFD policy statement
describes the performance objectives in
§ 26.23.
(3) The FFD policy statement
describes the minimum days off
requirements in § 26.205(d)(3) or
maximum average work hours
requirements in § 26.205(d)(7).
(4) The FFD policy statement must be
written in sufficient detail to provide
affected individuals with information
on what is expected of them and what
consequences may result from a lack of
adherence to the policy, including those
elements described in § 26.606(b), part
26-required sanctions, and required
medical/clinical treatment and followup testing for FFD policy violations.
(5) The FFD policy statement
describes the individual’s
responsibilities to report for work in a
physiological and psychological
condition that enables the safe and
competent performance of assigned
duties and responsibilities and inform a
licensee- or other entity-designated
representative when the individual
determines that this cannot be
accomplished.
(b) Licensees and other entities must
establish, implement, and maintain
written procedures that address the
following topics:
(1) If implementing a drug and
alcohol testing program under this
subpart,
(i) The methods and techniques to
collect and test for drugs and alcohol
and for the shipping and temporary
storage of biological specimens used for
drug testing at HHS-certified
laboratories,
(ii) The urine specimen volumes,
techniques for split specimen
collections, and the acceptability of a
urine specimen as described in § 26.111
or as described in the HHS Guidelines,
(iii) Protecting the privacy of an
individual who provides a specimen,
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protecting the integrity of the specimen,
and ensuring that the test results are
valid and attributable to the correct
individual, and
(iv) If the licensee or other entity
elects to use the HHS Guidelines, the
name of the specific HHS Guideline and
revision being implemented by the
licensee or other entity and a
description of the specific sections in
the guideline that are being
implemented in the procedure,
including specimen collections, drug
testing, and evaluation of test results.
(2) The immediate and follow-up
actions that will be taken, and the
procedures to be used, in those cases in
which individuals who are subject to
the FFD program:
(i) Have been involved in the use,
sale, or possession of illegal substances,
illegal drugs, or illicit substances;
(ii) Are impaired by any illegal
substances, illegal drugs, or illicit
substances or the consumption of
alcohol as determined by behavioral
observation or a test that measures
blood alcohol concentration;
(iii) If drug and alcohol testing is
conducted, attempted to subvert the
testing process by adulterating or
diluting specimens (in vivo or in vitro),
substituting specimens, or by any other
means;
(iv) If drug and alcohol testing is
conducted, refused to provide a
specimen for analysis or follow
instructions provided by FFD program
personnel;
(v) Had legal action taken relating to
drug or alcohol use; or
(vi) Demonstrated character or actions
indicating that the individual cannot be
trusted or relied upon to perform those
duties and responsibilities or maintain
access to NRC-licensed facilities, special
nuclear material (SNM), or sensitive
information.
(3) The process, including the duties
and responsibilities of FFD program
personnel, to be followed if an
individual’s behavior or condition raises
a concern regarding the possible use,
sale, or possession of illegal drugs onor offsite; the possible use or possession
of alcohol on the NRC-licensed facility;
impairment from any cause that in any
way could adversely affect the
individual’s ability to safely and
competently perform the individual’s
duties; or the receipt of credible
information indicating that the
individual cannot be trusted or relied on
to perform those duties and
responsibilities making the individual
subject to this part.
(4) Operation and oversight of an
onsite or offsite collection facility.
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(5) The fatigue management
requirements in §§ 26.202(b) and either
26.205(d)(3) or (d)(7).
(6) Measures to prevent subversion of
drug and alcohol tests conducted onsite
and offsite.
§ 26.607
Drug and alcohol testing.
Licensees and other entities
implementing § 26.604, at their
discretion, and licensees and other
entities implementing § 26.605 must
perform drug and alcohol testing that
complies with the following
requirements—
(a) Split specimens. Split specimen
collections of oral fluid or urine must be
used for the test conditions described in
paragraph (b) of this section. A split
specimen collection need not be used if
the licensee or other entity elects to use
a POCTA device for a screening test
conducted during random testing under
paragraphs (b)(2) and (h) of this section
or a protected area portal monitor
indication that drugs or alcohol were
detected under paragraph (j) of this
section. Testing of the split specimen
(specimen B) requires the donor’s
permission unless ordered by the MRO
to resolve an invalid test result obtained
for specimen A.
(b) Test conditions. Individuals
identified in § 26.4 must be subject to
drug and alcohol testing under the
following conditions:
(1) Pre-access. A pre-access test must
be conducted for drugs and alcohol
before performing or directing the
conduct of roles and responsibilities
making the individual subject to this
subpart or being granted unescorted
access to the protected areas of the NRClicensed facility. A pre-access test must
have been conducted no more than 14
days before the individual is granted
unescorted access.
(2) Random. Random testing for drugs
and alcohol must—
(i) Be administered in a manner that
provides reasonable assurance that
individuals are unable to predict the
time periods during which specimens
will be collected;
(ii) Require individuals who are
selected for random testing to report to
the onsite collection site as soon as
reasonably practicable after notification,
within the time period specified in the
FFD program procedure;
(iii) Ensure that all individuals in the
population that is subject to random
testing on a given day have an equal
probability of being selected and tested;
(iv) Ensure that an individual
completing a test is immediately eligible
for another random test; and
(v) Ensure that the sampling process
used to select individuals for random
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testing provides that the number of
random tests performed annually is
equal to at least 50 percent for licensee
employees and 50 percent for
contractor/vendors at the NRC-licensed
site.
(3) For-cause. A for-cause drug test,
alcohol test, or both, must be conducted
onsite in response to an individual’s
observed behavior or physical condition
indicating possible substance abuse or
after receiving credible information that
an individual is engaging in substance
abuse, as defined in § 26.5;
(4) Post-event. A post-event test for
drugs and alcohol must be conducted—
(i) As soon as practical after an event
involving a human error that was
committed by an individual specified in
§ 26.4, where the human error may have
caused or contributed to the event. This
test must be conducted onsite unless the
individual requires offsite medical care.
The licensee or other entity must test
the individual(s) who committed or
directed the error and need not test
individuals who were affected by the
event and whose actions likely did not
cause or contribute to the event. The
licensee or other entity must describe in
its procedures what constitutes a human
error.
(ii) Within 4 hours of an event unless
immediate medical intervention
precludes the conduct of the test on the
individual(s) who caused or contributed
to the accident(s), if the event results
in—
(A) An illness or personal injury to
any individual which results in death,
days away from work, restricted work,
transfer to another job, medical
treatment beyond first aid, loss of
consciousness, or other significant
illness or injury, as diagnosed by a
licensee- or other entity-designated
physician or other licensed health care
professional, even if the illness or injury
does not result in death, days away from
work, restricted work or job transfer,
medical treatment beyond first aid, or
loss of consciousness; or
(B) Damage to any safety- or securityrelated structures, systems, and
components; and
(5) Follow-up. An individual subject
to part 26 who has violated the FFD
policy for substance use or abuse, or the
sale, use, or possession of illegal drugs
must be subject to a follow-up series of
tests for drugs, alcohol, or both to verify
an individual’s continued abstinence
from substance abuse.
(c) Urine and oral fluid specimens. (1)
All urine or oral fluid specimens must
be subject to validity testing, including
an adulterant and biological marker, and
tested for the substances listed in
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§ 26.31(d)(1), except as allowed by
§ 26.603(e)(3).
(2) For the use of urine as the
biological specimen to be tested, the
following requirements must be
implemented—
(i) § 26.115, Collecting a urine
specimen under direct observation;
(ii) § 26.119, Determining ‘‘shy’’
bladder; and
(iii) § 26.163, Cutoff levels for drugs
and drug metabolites, (a)(2) regarding
special analysis testing.
(3) For alcohol testing onsite, the
following requirements must be
implemented—
(i) § 26.91, Acceptable devices for
conducting initial and confirmatory
tests for alcohol and methods of use;
(ii) § 26.93, Preparing for alcohol
testing;
(iii) § 26.95, Conducting an initial test
for alcohol using a breath specimen;
(iv) § 26.97, Collecting oral fluid
specimens for alcohol and drug testing;
(v) § 26.99, Determining the need for
a confirmatory test for alcohol;
(vi) § 26.101, Conducting a
confirmatory test for alcohol; and,
(vii) § 26.103, Determining a
confirmed positive test result for
alcohol.
(4) For all test conditions in paragraph
(b) of this section, except for the use of
a POCTA screening device in paragraph
(h) of this section, and for MRO-directed
tests under § 26.185, drug testing must
be performed at an HHS-certified
laboratory for the specific biological
specimen to be tested. Only HHScertified laboratory test results from
urine and oral fluid specimens may be
used for the issuance of a part 26required sanction. The licensee or other
entity must establish and maintain a
contract with a primary and a back-up
HHS-certified laboratory (with a
different Certifying Scientist) for the
specimen(s) to be tested. These
contracts must stipulate that the
laboratories are subject to inspection or
audit by the licensee or other entity and
that records and documents must be
provided and/or able to be photocopied
and removed from the premises to
support the inspection or audit.
(d) Privacy and integrity. The
specimen collection and drug and
alcohol testing procedures of FFD
programs must protect the donor’s
privacy and the integrity of the
specimen and implement quality
controls to ensure that test results are
valid and attributable to the correct
individual.
(e) Offsite collection facilities. At the
licensee’s or other entity’s discretion,
specimen collections and alcohol testing
may be conducted at a local hospital or
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other facility licensed to conduct
specimen collections and perform
alcohol testing and audited by the State
or a State-designated entity. The
licensee or other entity must audit these
facilities, if used, before their initial use
and then on a biennial basis to confirm
that the facility procedures are
comparable to those described in
subpart E of this part or the HHS
Guidelines for urine and oral fluid.
(f) Initial testing. A licensee or other
entity subject to this subpart performing
an initial test must use an
immunoassay, or an alternative
technology established in its FFD
program through § 26.603(e), that
satisfies the requirements of the U.S.
Food and Drug Administration (FDA)
for commercial distribution. Specimens
that yield positive, adulterated,
substituted, or invalid initial validity or
drug test results or discrepant biological
markers must be subject to confirmatory
testing by an HHS-certified laboratory,
certified for that biological specimen,
except for invalid specimens that cannot
be tested.
(g) Oral fluid testing. If the licensee or
other entity elects to use oral fluid for
drug or alcohol testing, the collection,
packaging, and temporary storage of the
drug or alcohol test device, and
shipment of an oral fluid specimen to an
HHS-certified laboratory or the
collection of an oral fluid specimen for
alcohol testing must be performed in
accordance with licensee- or other
entity-established procedures based
either on the requirements in part 26 or
the procedures in HHS Guidelines
identified by the licensee or other entity
in § 26.606(b)(1)(iv). The device must
have received premarket approval from
the FDA and must not expire before
laboratory testing. The drugs, drug
metabolites, initial and confirmatory
testing cutoffs, and biological markers, if
applicable, must be those established by
HHS for oral fluid testing and the
alcohol cutoffs in this part or, if not
established by HHS or the NRC for the
panel of drugs and drug metabolites to
be tested, as determined and
documented by a forensic toxicologist
review conducted pursuant to
§ 26.31(d)(1)(i)(D).
(h) Point of collection testing and
assessment. (1) If the licensee or other
entity elects to use a POCTA device,
then it may only be used for pre-access
and random drug and alcohol initial
testing in paragraph (b) of this section,
the alcohol testing process in paragraph
(c)(3) of this section, and the portal area
screening process in paragraph (j) of this
section. Before the licensee or other
entity uses a POCTA device, a forensic
toxicologist must review and document
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their evaluation that the validity and
accuracy of the device for alcohol and/
or the drugs and drug metabolites listed
in § 26.31(d) are comparable to the
performance achieved by initial testing
conducted using a similar technology at
an HHS-certified laboratory. For initial
testing of drugs and drug metabolites
using a POCTA device, this review must
include a documented evaluation of
POCTA device performance against the
requirements in § 26.161(b) for a urine
specimen or the procedures in the HHS
Guidelines for urine or oral fluid, as
implemented by the licensee or other
entity through its procedures.
(2) If the performance of the POCTA
device is not comparable to that
achieved from initial testing conducted
by an HHS-certified laboratory as
determined by the forensic toxicologist,
then the licensee or other entity must
implement a mitigating strategy to
maintain program effectiveness under
§ 26.603(e)(2), as applicable.
(3) The licensee and other entity must
implement procedures for the use of a
POCTA that ensures the effectiveness of
the collection process, assessment of the
screening results, and prevention of
subversion attempts.
(4) If the use of a POCTA device
indicates a discrepant biological marker
or that a test result exceeds the initial
test cutoff, the specimen is invalid, or
the individual subverted the drug or
alcohol test, then the individual must be
immediately removed from duties,
responsibilities, and access making the
individual subject to this subpart.
(i) The individual must be subject to
an immediate drug and alcohol test
using the alcohol testing process in
paragraph (c)(3) of this section for a
positive alcohol screen and either oral
fluid or urine by a collection kit that is
not a POCTA device, but of the same
type of biological specimen collected by
the POCTA, for validity, if required, and
initial and confirmatory testing by an
HHS-certified laboratory.
(ii) If this individual shows any signs
of impairment, the individual’s
authorization must be temporarily
removed until the MRO reviews the
laboratory test result(s), interviews the
individual, and performs a
determination of fitness under § 26.189
or § 26.619, as applicable, that enables
the restoration of authorization.
(i) Hair testing. The testing of hair
specimens may only be used to inform
a licensee’s or other entity’s
determination of whether the individual
is trustworthy and reliable under the
test condition in paragraph (b)(1) of this
section to supplement the information
gained from a pre-access test using oral
fluid or urine as the test specimen and
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must be conducted at an HHS-certified
laboratory certified for hair specimens.
(1) If used, this process must be
described in the licensee’s or other
entity’s FFD policy and described in
detail in its procedure. The panel of
drugs and drug metabolites to be
evaluated must only include those listed
as Schedule I or II of section 202 of the
Controlled Substances Act [21 U.S.C.
812]. The collection, packaging, and
temporary storage of a hair specimen
and shipment of the specimen to an
HHS-certified laboratory must be
conducted in accordance with the HHS
Guidelines. The test kit must be FDA
cleared, and licensee- or other entitydesignated FFD program personnel must
conduct the collection, packaging,
temporary storage, shipping, and
custody and control of the specimen.
(2) Before the licensee or other entity
begins to conduct hair testing, the initial
and confirmatory testing cutoffs must be
the cutoffs established by HHS for hair
testing or, if not established by HHS or
the NRC, as determined by a forensic
toxicologist review conducted pursuant
to § 26.31(d)(1)(i)(D).
(3) Confirmed positive test results
must be considered potentially
disqualifying FFD information until
proven otherwise by a review under
§ 26.613. Sanctions under this subpart
must not be issued for any FFD policy
violation involving a drug test using a
hair specimen unless the licensee or
other entity determines that the
individual subverted, as defined in
§ 26.5, the hair test.
(j) Portal area screening. A noninvasive point of collection testing
instrument may be used to screen
individuals for drugs, drug metabolites,
and alcohol before the individuals’
entry into or exit from a protected or
vital area.
(1) If a licensee or other entity uses
such an instrument, then before such
use, a forensic toxicologist must review
the instrument and document an
evaluation that the instrument and
setpoints used in the instrument are
acceptable for use for the detection and
screening of the drugs and drug
metabolites selected for screening from
the panel of drugs and drug metabolites
to be tested under the FFD program and
alcohol and its metabolites.
(2) The instrument must be operated
in accordance with the manufacturer’s
specifications. If screening detects the
presence of drugs, drug metabolites, or
alcohol at or above the instrument set
point(s), the individual screened by the
instrument must be subject to a POCTA
screening test using the process
described in paragraph (h) of this
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87035
section or an oral fluid or urine test that
is sent to an HHS-certified laboratory.
(3) A part 26 sanction may not be
issued to an individual based solely on
a portal area screening instrument
detection that drugs or alcohol exceed
the instrument’s established setpoint.
(k) Blood testing. The testing of blood
specimens may only be conducted
under the order of the licensee- or other
entity-designated MRO for a valid
medical reason as confirmed by the
MRO pursuant to § 26.31(d)(5). This
specimen must be subject to testing by
a laboratory that satisfies quality control
requirements that are comparable to
those required for certification by the
HHS.
(l) Custody-and-control form. For the
collection and packaging of urine, oral
fluid, and hair specimens, the licensee
or other entity must use a custody-andcontrol form approved by the U.S.
Office of Management and Budget. For
the use of a POCTA device, the licensee
or other entity must implement a
licensee- or other entity-approved and
-maintained procedure that ensures the
reliability of the tracking, handling, and
storage of a specimen from the point of
specimen collection to the final
disposition of the specimen and the
reliability of an identification system to
uniquely assign the specimen to the
donor.
(m) Medical Review Officer. Licensees
or other entities must—
(1) Require their designated MRO to
review positive, adulterated,
substituted, and dilute confirmatory
drug and validity test results and test
results of questionable validity to
determine whether the donor has
violated the FFD policy for urine and
oral fluid specimens. The review must
be completed before reporting the
results to the individual designated by
the licensee or other entity to assess
authorization or perform the suitability
and fitness determinations required
under § 26.619, or, if required, that are
described in subpart H of this part.
(2) Require their MRO to satisfy the
requirements in § 26.183 and, prior to
conducting any activities under this
part, attend and pass a medical- or
clinical-based training session to
improve his/her knowledge of MRO
duties and responsibilities, drug and
alcohol testing processes and
procedures, and evaluation of drug
testing results. This training session
must be conducted by a nationally
recognized MRO training and
certification organization that has been
assessed by the licensee’s or other
entity’s FFD program personnel to
include the technical elements an MRO
must implement under § 26.185. An
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MRO who performed the duties and
responsibilities in §§ 26.185 and 26.187
for at least 3 continuous years in the last
10 years prior to being hired or
contracted by the licensee or other
entity satisfies the requirements in this
paragraph.
(3) Require their MRO to attend a
medical- or clinical-based training
session on a triennial basis to improve
his/her knowledge of changes in drug
and alcohol testing processes and
procedures and evaluation of drug
testing results.
(4) Require their MRO to determine
whether a biological specimen is
positive, adulterated, substituted, dilute
or of questionable validity by
implementing the requirements in
§ 26.185 or the HHS Guidelines through
the licensee’s or other entity’s
procedures.
(i) If § 26.185 or the HHS Guidelines,
as used by the licensee or other entity
in its procedures, are insufficient to
make this determination, then guidance
issued by a State agency in the state in
which the NRC-licensed facility is
located, Federal agencies, or nationally
recognized MRO training and
certification organizations may be used
to inform an MRO determination.
(ii) An MRO need not review a
confirmed alcohol positive test result
determined by an evidentiary breath
testing device under paragraphs
(c)(3)(vi) and (vii) of this section.
(5) Require their MRO to determine
and approve the use of oral fluid or
urine as an alternative biological
specimen when the donor cannot
provide a specimen for testing. This
determination and the retest must be
documented and completed as soon as
reasonably practicable.
(6) Require the MRO to review all
specimens screened and tested
associated with a drug-related FFD
policy violation. This review includes
POCTA, split specimens, and all
specimens taken to resolve a discrepant
condition, such as a possible subversion
attempt, impairment without a known
cause, or a donor-requested or MROdirected re-test. To resolve a discrepant
condition, the MRO is authorized to test
a specimen for a biological marker,
adulterants, or additional drugs.
(n) Limitations of screening and
testing. Specimens collected under NRC
regulations may only be designated or
approved for screening and testing as
described in this part and may not be
used to conduct any other analysis or
test without the written permission of
the donor. Analyses, screens, and tests
that may not be conducted include, but
are not limited to, DNA testing,
serological typing, or any other medical
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or genetic test used for diagnostic or
specimen identification purposes. No
biological specimens may be passively
sampled and analyzed in a manner
different than described in this subpart.
(o) Specimen collectors. All onsite
specimen collections, except a
collection by a portal area screening
instrument in paragraph (j) of this
section, must be conducted by licenseeor other entity-designated and -trained
personnel.
§ 26.608
FFD program training.
(a) FFD program training. (1)
Individuals must be trained in the FFD
policy and procedure, including fatigue
management, and their FFD program
responsibilities. Individuals who collect
specimens for testing or screening must
also be trained in specimen collector
duties and responsibilities, including, at
a minimum, specimen collection,
custody and control, identification and
response to subversion attempts, and
privacy. For licensees and other entities
of commercial nuclear plants, the FFD
program training program must use a
systems approach to training as defined
in § 53.725 of this chapter and described
in § 53.830 of this chapter for those
individuals in § 26.4.
(2) FFD program training must
include training on the behavioral
observation program. The behavioral
observation program training must
include the detection of physiological
behaviors or conditions that may
indicate—
(i) Possible use, sale, or possession of
illegal drugs or illicit drugs, or
substance abuse on- or offsite;
(ii) Use or possession of alcohol onsite
or use while on duty offsite;
(iii) Impairment from fatigue or any
cause that, if left unattended, could
result in inattentiveness or human
errors; and
(iv) Any individual’s inability to
safely and competently perform
assigned duties and responsibilities or
act in a trustworthy and reliable manner
while having access to protected areas,
SNM, or sensitive information.
(3) Training must explain that an
individual’s FFD policy violation will—
(i) Subject the individual to an FFD
program-required sanction designed to
preclude recurrence of an FFD policy
violation;
(ii) Contribute to the licensee’s or
other entity’s assessment of whether the
individual can be trusted and relied
upon to safely and competently perform
the assigned duties and responsibilities
making the individual subject to this
subpart;
(iii) Be used to inform the licensee’s
or other entity’s insider mitigation and
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access authorization programs under
§§ 73.55, 73.56, 73.100 or 73.120 of this
chapter; and
(iv) Be used to inform other NRC
licensees and other entities subject to
part 26 when FFD program information
is requested to support authorization
determinations under subpart C of this
part or §§ 73.56 or 73.120 of this
chapter.
(b) Training and assessments.
Training and a trainee assessment must
be conducted before pre-access testing,
and refresher training and trainee
assessments must be conducted
periodically thereafter.
(c) Training program review. The
licensee or other entity must
periodically evaluate its FFD training
program and revise it as appropriate to
reflect industry experience as well as
applicable changes to the regulations in
this part, the HHS Guidelines, if used,
and specimen collection and testing
processes implemented by the licensee
or other entity.
§ 26.609
Behavioral observation.
(a) Licensees and other entities must
ensure that the individuals who are
subject to this subpart are subject to
behavioral observation and that
behavioral observation is performed by
all individuals subject to this subpart.
(b) Licensees and other entities must
require all individuals subject to the
FFD program to report to the licenseeor other entity-designated representative
any onsite or offsite behaviors or
activities by individuals subject to this
part that may constitute an
unreasonable risk to the safety or
security of the NRC-licensed facility or
SNM or may cause harm to others. This
reporting must include any information
relating to character or reputation of the
individual indicating that the individual
cannot be trusted or relied upon to
perform those duties and
responsibilities or maintain access to
NRC-licensed facilities, SNM, or
sensitive information that makes them
subject to part 26.
(c) Behavioral observation must be
performed visually, in-person, and,
when necessary, remotely by live video
and audible streaming and capture, to
observe the behavior of individuals in
the workforce subject to the
requirements in this subpart.
(d) Not withstanding paragraph (c) of
this section, for a reactor facility where
individual task loading does not allow
for the effective conduct of behavior
observation in addition to assigned
operational tasks, the licensee or other
entity must implement a live video and
audible streaming and capture system to
conduct behavioral observation of
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§ 26.613
persons licensed to operate under 10
CFR part 53 who manipulate the
controls of any commercial nuclear
plant licensed under 10 CFR part 53.
§ 26.610
Sanctions.
Licensees and other entities that
implement an FFD program under this
subpart must establish sanctions for
FFD policy violations that, at a
minimum, prohibit the individuals
specified in § 26.4 from being assigned
to perform or direct those duties and
responsibilities or maintaining
authorization making them subject to
this subpart. The severity of the
sanction must escalate with the number
of occurrences and severity of the FFD
policy violation. The sanction must be
long enough to act as a deterrent and,
if the individual is retained as a licensee
employee or contractor/vendor,
facilitate the individual to complete
counseling or treatment. The sanctions
must include a minimum 5-year denial
of access to the NRC-licensed facility for
any individual who is determined to
have been involved in the sale, use, or
possession of illegal drugs or the
consumption of alcohol within a
protected area of any facility licensed
under part 53 of this chapter or within
a transporter’s facility or vehicle used in
the conveyance of formula quantities of
strategic SNM while the individual is
subject to this subpart, and a permanent
denial of access to the NRC-licensed
facility for three FFD policy violations
or any subversion attempt of any drug
or alcohol test or screening process,
including subversion attempts at any
licensee or other entity subject to this
part.
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§ 26.611
Protection of information.
(a) Licensees and other entities that
collect personal information about an
individual for the purpose of complying
with this subpart must establish and
maintain a system of files and
procedures to prevent unauthorized
disclosure.
(b) Licensees and other entities must
obtain a signed consent that documents
the individual’s acceptance of being
subject to the FFD program and
authorizes the disclosure of the personal
information collected and maintained
under this subpart, except for
disclosures to the individuals and
entities specified in § 26.37(b)(1)
through (b)(6), (b)(8), and persons
deciding matters under review in
§ 26.613. This signed and dated consent
must be obtained before making the
individual subject to the FFD program.
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Appeals process.
Licensees and other entities that
implement an FFD program under this
subpart must establish and implement
procedures for the review of a
determination that an individual in
§ 26.4 has violated the FFD policy. The
procedure must provide for an objective
and impartial review of the facts related
to the determination that the individual
has violated the FFD policy and a
schedule for the completion of the
review.
§ 26.615
Audits.
(a) Licensees and other entities that
implement an FFD program under this
subpart must audit their programs at a
frequency that ensures the continuing
effectiveness of their FFD program, FFD
program elements that are provided by
C/Vs, and the FFD programs of C/Vs
that are accepted by the licensee or
other entity. Corrective actions must be
as soon as reasonably practicable to
resolve any problems identified in an
audit and preclude recurrence.
(b) The subject matter, scope, and
frequency of audits must be revised as
necessary to improve or maintain
program performance based on findings
resulting from licensee or other entity
implementation of its FFD PMRP in
§ 26.603(d).
(c) Licensees and other entities may
conduct joint audits or accept audits of
C/Vs so long as the audit addresses the
relevant services of the C/Vs.
(d) Licensees and other entities must
audit HHS-certified laboratories unless
the licensee’s or other entity’s panel of
drugs and drug metabolites to be tested
is equivalent to the panel by which the
laboratory is certified by HHS or is
subject to the standards and procedures
for drug testing and evaluation used by
the laboratory under the HHS
Guidelines. Licensees and other entities
must audit any hospital or other facility
licensed by the State (or Statedesignated entity) if used to conduct
specimen collections and perform
alcohol testing under this part on a
biennial basis to confirm that the facility
procedures are comparable to those
described in subpart E of this part, for
urine and oral fluid.
§ 26.617
Recordkeeping and reporting.
(a) Licensees and other entities that
implement FFD programs under this
subpart must ensure that records
pertaining to the administration of their
program, which may be stored and
archived electronically, are maintained
so that they are available for NRC
inspection purposes and for any legal
proceedings resulting from the
administration of the program. Records
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87037
pertaining to the administration of the
FFD program and FFD performance data
required by § 26.717 must be retained
until license termination.
(b) Licensees and other entities must
make the following reports:
(1) Reports to the NRC Operations
Center by telephone within 24 hours
after the licensee or other entity
discovers any intentional act that casts
doubt on the integrity of the FFD
program and any programmatic failure,
degradation, or discovered vulnerability
of the FFD program that may permit
undetected drug or alcohol use or abuse
by individuals who are subject to this
subpart. These events must be reported
under this subpart, rather than under
the provisions of § 73.1200 of this
chapter; and
(2) Annual program performance
reports for the FFD program, including
the FFD program performance data
listed in § 26.717(b), as applicable.
Licensees and other entities must
submit FFD program performance data
(for January through December) to the
NRC annually, before March 1 of the
following year and must use unexpired
NRC-provided forms for the electronic
submission of FFD information to the
NRC.
(c) Licensees and other entities
subject to this subpart must describe in
sufficient detail to support an
authorization determination, an
individual’s FFD policy violation (while
protecting privacy information under
§ 26.611) and FFD program weakness to
NRC, licensees, and other entities
subject to this part when requested to
support authorization determinations
under subpart C of this part or § 73.120
of this chapter, as applicable, or to
support licensee or other entity
performance monitoring.
§ 26.619 Suitability and fitness
determinations.
Licensees and other entities that
implement FFD programs under this
subpart must develop, implement, and
maintain procedures for evaluating
whether to assign individuals to
perform or direct those duties and
responsibilities making them subject to
this subpart. A suitability or fitness
determination conducted for cause must
be performed face to face. A suitability
or fitness determination conducted for
cause may be performed remotely using
electronic communications only when
supported by someone who is present
in-person with the individual being
assessed, and that supporting person
must be trained in accordance with the
requirements of either §§ 26.29 or
26.608.
■ 94. Revise § 26.709 to read as follows:
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Applicability.
(a) The requirements of this subpart
apply to the FFD programs of licensees
and other entities specified in § 26.3(a)
through (d), except for FFD programs
that are implemented under subpart K
of this part.
(b) The requirements in this subpart
apply to the FFD programs of licensees
and other entities specified in § 26.3(f)
that elect not to implement the
requirements in subpart M or elect to
implement the requirements in
§ 26.605(b).
§ 26.711
[Amended]
[Amended]
96. In § 26.825, in paragraph (b) add
remove the phrase ‘‘§§ 26.1, 26.3, 26.5,
26.7, 26.8, 26.9, 26.11, 26.51, 26.81,
26.121, 26.151, 26.181, 26.201, 26.823,
and 26.825’’ and add in its place the
phrase ‘‘§§ 26.1, 26.3, 26.5, 26.7, 26.8,
26.9, 26.11, 26.51, 26.81, 26.121, 26.151,
26.181, 26.201, 26.601, 26.823, and
26.825’’.
■
PART 30—RULES OF GENERAL
APPLICABILITY TO DOMESTIC
LICENSING OF BYPRODUCT
MATERIAL
Authority: Atomic Energy Act of 1954,
secs. 11, 81, 161, 181, 182, 183, 184, 186,
187, 223, 234, 274 (42 U.S.C. 2014, 2111,
2201, 2231, 2232, 2233, 2234, 2236, 2237,
2273, 2282, 2021); Energy Reorganization Act
of 1974, secs. 201, 202, 206, 211 (42 U.S.C.
5841, 5842, 5846, 5851); 44 U.S.C. 3504 note.
98. In § 30.4, revise the definition for
‘‘Utilization facility’’ to read as follows:
■
Definitions.
*
*
*
*
*
Utilization facility means a utilization
facility as defined in the regulations
contained in part 50 or part 53 of this
chapter;
■ 99. In § 30.50, revise paragraph (c)(3)
to read as follows:
Reporting requirements.
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*
*
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(c) * * *
(3) The provisions of this section do
not apply to licensees subject to the
notification requirements in §§ 50.72 or
53.1630 of this chapter. They do apply
to those part 50 licensees possessing
material licensed under this part, who
are not subject to the notification
requirements in § 50.72 of this chapter.
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101. In § 40.60, revise paragraph (c)(3)
to read as follows:
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Reporting requirements.
*
*
*
*
*
(c) * * *
(3) The provisions of this section do
not apply to licensees subject to the
notification requirements in §§ 50.72 or
53.1630 of this chapter. They do apply
to those part 50 licensees possessing
material licensed under this part who
are not subject to the notification
requirements in § 50.72 of this chapter.
PART 50—DOMESTIC LICENSING OF
PRODUCTION AND UTILIZATION
FACILITIES
102. The authority citation for part 50
continues to read as follows:
97. The authority citation for part 30
continues to read as follows:
§ 30.50
Authority: Atomic Energy Act of 1954,
secs. 62, 63, 64, 65, 69, 81, 83, 84, 122, 161,
181, 182, 183, 184, 186, 187, 193, 223, 234,
274, 275 (42 U.S.C. 2092, 2093, 2094, 2095,
2099, 2111, 2113, 2114, 2152, 2201, 2231,
2232, 2233, 2234, 2236, 2237, 2243, 2273,
2282, 2021, 2022); Energy Reorganization Act
of 1974, secs. 201, 202, 206, 211 (42 U.S.C.
5841, 5842, 5846, 5851); Uranium Mill
Tailings Radiation Control Act of 1978, sec.
104 (42 U.S.C. 7914); 44 U.S.C. 3504 note.
■
■
§ 30.4
100. The authority citation for part 40
continues to read as follows:
■
■
95. In § 26.711, in paragraphs (c) and
(d), remove the phrase ‘‘(c) and (d),’’ and
add in its place the phrase ‘‘(c), (d), and
(f),’’.
■
§ 26.825
PART 40—DOMESTIC LICENSING OF
SOURCE MATERIAL
Authority: Atomic Energy Act of 1954,
secs. 11, 101, 102, 103, 104, 105, 108, 122,
147, 149, 161, 181, 182, 183, 184, 185, 186,
187, 189, 223, 234 (42 U.S.C. 2014, 2131,
2132, 2133, 2134, 2135, 2138, 2152, 2167,
2169, 2201, 2231, 2232, 2233, 2234, 2235,
2236, 2237, 2239, 2273, 2282); Energy
Reorganization Act of 1974, secs. 201, 202,
206, 211 (42 U.S.C. 5841, 5842, 5846, 5851);
Nuclear Waste Policy Act of 1982, sec.
306(42 U.S.C. 10226); National
Environmental Policy Act of 1969 (42 U.S.C.
4332); 44 U.S.C. 3504 note; Sec. 109, Pub. L.
96–295, 94 Stat. 783.
103. In § 50.47, revise paragraphs
(a)(1) and (e) to read as follows:
■
§ 50.47
Emergency plans.
(a)(1)(i) Except as provided in
paragraph (d) of this section, no initial
operating license for a nuclear power
reactor will be issued under this part or
under part 53 of this chapter unless a
finding is made by the NRC that there
is reasonable assurance that adequate
protective measures can and will be
taken in the event of a radiological
emergency. No finding under this
section is necessary for issuance of a
renewed nuclear power reactor
operating license.
(ii) No initial combined license under
parts 52 or 53 of this chapter will be
issued unless a finding is made by the
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NRC that there is reasonable assurance
that adequate protective measures can
and will be taken in the event of a
radiological emergency. No finding
under this section is necessary for
issuance of a renewed combined
license.
(iii) If an application for an early site
permit under subpart A of part 52 of this
chapter includes complete and
integrated emergency plans under
§ 52.17(b)(2)(ii) of this chapter or an
application for an early site permit
under subpart H of part 53 of this
chapter includes complete and
integrated emergency plans under
§ 53.1146(b)(2)(ii) of this chapter, no
early site permit will be issued unless
a finding is made by the NRC that the
emergency plans provide reasonable
assurance that adequate protective
measures can and will be taken in the
event of a radiological emergency.
(iv) If an application for an early site
permit proposes major features of the
emergency plans under §§ 52.17(b)(2)(i)
or 53.1146(b)(2)(i) of this chapter, no
early site permit will be issued unless
a finding is made by the NRC that the
major features are acceptable in
accordance with the applicable
standards of either § 50.47 and appendix
E to this part or the applicable
requirements of § 50.160, within the
scope of emergency preparedness
matters addressed in the major features.
*
*
*
*
*
(e) Notwithstanding the requirements
of paragraph (b) of this section and the
provisions of § 52.103 or § 53.1452 of
this chapter, a holder of a combined
license under part 52 or part 53 of this
chapter, as applicable, that is complying
with the requirements of § 50.47(b) and
appendix E to this part may not load
fuel or operate except as provided in
accordance with appendix E to this part
and § 50.54(gg), and a holder of a
combined license under part 52 or part
53 of this chapter that is complying with
the requirements of § 50.160 may not
load fuel or operate except as provided
in accordance with § 50.160(c)(2) and
§ 50.54(gg).
*
*
*
*
*
■ 104. In § 50.54, revise paragraphs
(q)(2), (q)(4), and (gg)(1) introductory
text to read as follows:
§ 50.54
Conditions of licenses.
*
*
*
*
*
(q) * * *
(2)(i) Except as provided in paragraph
(q)(2)(ii) of this section, a holder of a
license under this part, or a combined
license under parts 52 or 53 of this
chapter after the Commission makes the
finding under §§ 52.103(g) or 53.1452(g)
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of this chapter, as applicable, shall
follow and maintain the effectiveness of
an emergency plan that meets the
requirements in appendix E to this part
and, for nuclear power reactor licensees,
the planning standards of § 50.47(b).
(ii) A holder of a license under this
part for a non-power production or
utilization facility, a holder of a license
under this part or part 53 of this chapter
for a small modular reactor or a nonlight-water reactor, or a holder of a
combined license under parts 52 or 53
of this chapter after the Commission
makes the finding under §§ 52.103(g) or
53.1452(g) of this chapter, as applicable,
for a small modular reactor or a nonlight-water reactor, shall follow and
maintain the effectiveness of either an
emergency plan that meets the
requirements in § 50.160 or an
emergency plan that meets the
requirements in appendix E to this part
and, for nuclear power reactor licensees,
the planning standards of § 50.47(b).
*
*
*
*
*
(4) The changes to a licensee’s
emergency plan that reduce the
effectiveness of the plan as defined in
paragraph (q)(1)(iv) of this section may
not be implemented without prior
approval by the NRC. A licensee
desiring to make such a change shall
submit an application for an
amendment to its license. In addition to
the filing requirements of §§ 50.90 and
50.91 or §§ 53.1510 and 53.1515 of this
chapter, as applicable, the request must
include all emergency plan pages
affected by that change and must be
accompanied by a forwarding letter
identifying the change, the reason for
the change, and the basis for concluding
that the licensee’s emergency plan, as
revised, will continue to meet either the
requirements in § 50.160 or the
requirements in appendix E to this part
and, for nuclear power reactor licensees,
the planning standards of § 50.47(b).
*
*
*
*
*
(gg)(1) Notwithstanding §§ 52.103 or
53.1452 of this chapter, if following the
conduct of the exercise required by
paragraph IV.f.2.a of appendix E to this
part or § 50.160(c)(2), as applicable,
FEMA identifies one or more
deficiencies in the state of offsite
emergency preparedness, the holder of a
combined license under 10 CFR part 52
or under 10 CFR part 53, as applicable,
may operate at up to 5 percent of rated
thermal power only if the Commission
finds that the state of onsite emergency
preparedness provides reasonable
assurance that adequate protective
measures can and will be taken in the
event of a radiological emergency. The
NRC will base this finding on its
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assessment of the applicant’s onsite
emergency plans against the pertinent
standards in either § 50.47 and
appendix E to this part, or § 50.160, as
applicable. Review of the applicant’s
emergency plans will include the
following standards with offsite aspects:
*
*
*
*
*
■ 105. In § 50.160, revise paragraphs
(b)(3) and (c)(2) to read as follows:
§ 50.160 Emergency preparedness for
small modular reactors, non-light-water
reactors, and non-power production or
utilization facilities.
*
*
*
*
*
(b) * * *
(3) Emergency planning zone. For an
applicant whose analysis required by
§ 50.33(g)(2) or § 53.1109(g)(2) of this
chapter meets the criteria in
§ 50.33(g)(2)(i) or § 53.1109(g)(2)(i) of
this chapter, as applicable, determine
and describe the boundary and physical
characteristics of the EPZ in the
emergency plan.
*
*
*
*
*
(c) * * *
(2) A holder of a combined license
issued under parts 52 or 53 of this
chapter before the Commission has
made the finding under §§ 52.103(g) or
53.1452(g) of this chapter, as applicable,
must establish, implement, and
maintain an emergency preparedness
program that meets the requirements of
paragraph (b) of this section, as
described in the approved emergency
plan and license, and conduct an initial
exercise to demonstrate this compliance
within 2 years before the scheduled date
for initial loading of fuel (or, for a fueled
manufactured reactor, within 2 years
before the scheduled date for initiating
the physical removal of any one of the
independent physical mechanisms to
prevent criticality required under
§ 53.620(d)(1) of this chapter).
■ 106. In appendix B to part 50, revise
the first paragraph in the Introduction
section, the first paragraph of section III,
Design Control, and section IV,
Procurement Document Control, to read
as follows:
Appendix B to Part 50—Quality
Assurance Criteria for Nuclear Power
Plants and Fuel Reprocessing Plants
Introduction. Every applicant for a
construction permit is required by the
provisions of § 50.34 or § 53.1309 of this
chapter to include in its preliminary safety
analysis report a description of the quality
assurance program to be applied to the
design, fabrication, construction, and testing
of the structures, systems, and components of
the facility. Every applicant for an operating
license is required by the provisions of
§ 50.34 or § 53.1369 of this chapter to
include, in its final safety analysis report,
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87039
information pertaining to the managerial and
administrative controls to be used to assure
safe operation. Every applicant for a
combined license is required by the
provisions of §§ 52.79 or 53.1416 of this
chapter to include in its final safety analysis
report a description of the quality assurance
applied to the design, and to be applied to
the fabrication, construction, and testing of
the structures, systems, and components of
the facility and to the managerial and
administrative controls to be used to assure
safe operation. For applications submitted
after September 27, 2007, every applicant for
an early site permit is required by the
provisions of §§ 52.17 or 53.1146 of this
chapter to include in its site safety analysis
report a description of the quality assurance
program applied to site activities related to
the design, fabrication, construction, and
testing of the structures, systems, and
components of a facility or facilities that may
be constructed on the site. Every applicant
for a design approval is required by the
provisions of §§ 52.137 or 53.1209 of this
chapter to include in its final safety analysis
report a description of the quality assurance
program applied to the design of the
structures, systems, and components of the
facility. Every applicant for a design
certification is required by the provisions of
§§ 52.47 or 53.1239 of this chapter to include
in its final safety analysis report a description
of the quality assurance program applied to
the design of the structures, systems, and
components of the facility. Every applicant
for a manufacturing license is required by the
provisions of §§ 52.157 or 53.1279 of this
chapter to include in its final safety analysis
report a description of the quality assurance
program applied to the design, and to be
applied to the manufacture of, the structures,
systems, and components of the reactor.
Nuclear power plants and fuel reprocessing
plants include structures, systems, and
components that prevent or mitigate the
consequences of postulated accidents that
could cause undue risk to the health and
safety of the public. This appendix
establishes quality assurance requirements
for the design, manufacture, construction,
and operation of those structures, systems,
and components. The pertinent requirements
of this appendix apply to all activities
affecting the safety-related functions of those
structures, systems, and components; these
activities include designing, purchasing,
fabricating, handling, shipping, storing,
cleaning, erecting, installing, inspecting,
testing, operating, maintaining, repairing,
refueling, and modifying.
*
*
*
*
*
III. Design Control
Measures shall be established to assure that
applicable regulatory requirements and the
design bases, as defined in § 50.2 and as
specified in the license application, or the
functional design criteria, as defined in
§ 53.020 of this chapter and as specified in
the license application, for those structures,
systems, and components to which this
appendix applies are correctly translated into
specifications, drawings, procedures, and
instructions. These measures shall include
provisions to assure that appropriate quality
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standards are specified and included in
design documents and that deviations from
such standards are controlled. Measures shall
also be established for the selection and
review for suitability of application of
materials, parts, equipment, and processes
that are essential to the safety-related
functions of the structures, systems and
components.
*
*
*
*
*
IV. Procurement Document Control
Measures shall be established to assure that
applicable regulatory requirements, design
bases or functional design criteria, and other
requirements which are necessary to assure
adequate quality are suitably included or
referenced in the documents for procurement
of material, equipment, and services, whether
purchased by the applicant or by its
contractors or subcontractors. To the extent
necessary, procurement documents shall
require contractors or subcontractors to
provide a quality assurance program
consistent with the pertinent provisions of
this appendix.
*
*
*
*
*
PART 51—ENVIRONMENTAL
PROTECTION REGULATIONS FOR
DOMESTIC LICENSING AND RELATED
REGULATORY FUNCTIONS
107. The authority citation for part 51
continues to read as follows:
■
Authority: Atomic Energy Act of 1954,
secs. 161, 193 (42 U.S.C. 2201, 2243); Energy
Reorganization Act of 1974, secs. 201, 202
(42 U.S.C. 5841, 5842); National
Environmental Policy Act of 1969 (42 U.S.C.
4332, 4334, 4335); Nuclear Waste Policy Act
of 1982, secs. 144(f), 121, 135, 141, 148 (42
U.S.C. 10134(f), 10141, 10155, 10161, 10168);
44 U.S.C. 3504 note.
108. In § 51.20, revise paragraphs
(b)(1) and (2) to read as follows:
■
§ 51.20 Criteria for and identification of
licensing and regulatory actions requiring
environmental impact statements.
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*
*
*
*
(b) * * *
(1) Issuance of a limited work
authorization or a permit to construct a
nuclear power reactor, testing facility, or
fuel reprocessing plant under part 50 of
this chapter, issuance of an early site
permit under part 52 of this chapter, or
issuance of a limited work
authorization, construction permit, or
early site permit under part 53 of this
chapter.
(2) Issuance or renewal of a full power
or design capacity license to operate a
nuclear power reactor, testing facility, or
fuel reprocessing plant under parts 50 or
53 of this chapter, or a combined license
under parts 52 or 53 of this chapter.
*
*
*
*
*
■ 109. In § 51.22, revise paragraphs
(c)(3) introductory text, (c)(9)
introductory text, (c)(12) introductory
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text, (c)(17), (c)(22) and (23) to read as
follows:
§ 51.22 Criterion for categorical exclusion;
identification of licensing and regulatory
actions eligible for categorical exclusion or
otherwise not requiring environmental
review.
*
*
*
*
*
(c) * * *
(3) Amendments to parts 20, 30, 31,
32, 33, 34, 35, 37, 39, 40, 50, 51, 52, 53,
54, 60, 61, 63, 70, 71, 72, 73, 74, 81, and
100 of this chapter which relate to—
*
*
*
*
*
(9) Issuance of an amendment to a
permit or license for a reactor under part
50, part 52, or part 53 of this chapter
that changes a requirement or issuance
of an exemption from a requirement,
with respect to installation or use of a
facility component located within the
restricted area, as defined in part 20 of
this chapter; or the issuance of an
amendment to a permit or license for a
reactor under part 50, part 52, or part 53
of this chapter that changes an
inspection or a surveillance
requirement; provided that:
*
*
*
*
*
(12) Issuance of an amendment to a
license under parts 50, 52, 53, 60, 61,
63, 70, 72, or 75 of this chapter relating
solely to safeguards matters (i.e.,
protection against sabotage or loss or
diversion of special nuclear material) or
issuance of an approval of a safeguards
plan submitted under parts 50, 52, 53,
70, 72, and 73 of this chapter, provided
that the amendment or approval does
not involve any significant construction
impacts. These amendments and
approvals are confined to—
*
*
*
*
*
(17) Issuance of an amendment to a
permit or license under part 30, part 40,
part 50, part 52, part 53, or part 70 of
this chapter which deletes any limiting
condition of operation or monitoring
requirement based on or applicable to
any matter subject to the provisions of
the Federal Water Pollution Control Act.
*
*
*
*
*
(22) Issuance of a standard design
approval under part 52 or part 53 of this
chapter.
(23) The Commission finding for a
combined license under § 52.103(g) or
§ 53.1452(g) of this chapter.
*
*
*
*
*
§ 51.26
[Amended]
110. In § 51.26, in paragraph (d)
remove the phrase ‘‘under part 52’’ and
add in its place the phrase ‘‘under 10
CFR parts 52 or 53,’’.
■ 111. In § 51.30, revise paragraph (a)
introductory text and paragraphs (d) and
(e) to read as follows:
■
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§ 51.30
Environmental assessment.
(a) An environmental assessment for
proposed actions, other than those for a
standard design certification under 10
CFR parts 52 or 53, or a manufacturing
license under 10 CFR parts 52 or 53,
shall identify the proposed action and
include:
*
*
*
*
*
(d) An environmental assessment for
a standard design certification under
subpart B of part 52 of this chapter, or
under subpart H of part 53 of this
chapter must identify the proposed
action and will be limited to the
consideration of the costs and benefits
of severe accident mitigation design
alternatives and the bases for not
incorporating severe accident mitigation
design alternatives in the design
certification. An environmental
assessment for an amendment to a
design certification will be limited to
the consideration of whether the design
change which is the subject of the
proposed amendment renders a severe
accident mitigation design alternative
previously rejected in the earlier
environmental assessment to become
cost beneficial, or results in the
identification of new severe accident
mitigation design alternatives, in which
case the costs and benefits of new severe
accident mitigation design alternatives
and the bases for not incorporating new
severe accident mitigation design
alternatives in the design certification
must be addressed.
(e) An environmental assessment for a
manufacturing license under subpart F
of part 52 of this chapter or under
subpart H of part 53 of this chapter must
identify the proposed action and will be
limited to the consideration of the costs
and benefits of severe accident
mitigation design alternatives and the
bases for not incorporating severe
accident mitigation design alternatives
in the manufacturing license. An
environmental assessment for an
amendment to a manufacturing license
will be limited to consideration of
whether the design change which is the
subject of the proposed amendment
either renders a severe accident
mitigation design alternative previously
rejected in an environmental assessment
to become cost beneficial, or results in
the identification of new severe accident
mitigation design alternatives, in which
case the costs and benefits of new severe
accident mitigation design alternatives
and the bases for not incorporating new
severe accident mitigation design
alternatives in the manufacturing
license must be addressed. In either
case, the environmental assessment will
not address the environmental impacts
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associated with manufacturing the
reactor under the manufacturing license.
§ 51.31
[Amended]
112. In § 51.31, in paragraph (a)
remove the phrase ‘‘under part 52’’ and
add in its place the phrase ‘‘under parts
52 or 53’’.
■
§ 51.32
[Amended]
113. In § 51.32, in paragraphs (b)(1)
and (3) remove the phrase ‘‘of part 52 of
this chapter’’ and add in its place the
phrase ‘‘of part 52 of this chapter or
subpart H of part 53 of this chapter’’.
■
§ 51.49
[Amended]
114. In § 51.49, in paragraph (c)
introductory text, remove the phrase ‘‘of
part 52 of this chapter’’ and add in its
place the phrase ‘‘of part 52 of this
chapter or under subpart H of part 53 of
this chapter’’.
■
§ 51.50
[Amended]
115. In § 51.50, wherever it appears,
remove the phrase ‘‘in accordance with
§ 50.36b of this chapter’’ and add in its
place the phrase ‘‘in accordance with
§§ 50.36b or 53.1112 of this chapter’’.
■
§ 51.53
[Amended]
116. In § 51.53, in paragraph (d)
remove the phrase ‘‘under § 50.82 of this
chapter’’ and add in its place the phrase
‘‘under §§ 50.82 or 53.1080 of this
chapter’’.
■
§ 51.54
[Amended]
117. In § 51.54, in paragraph (a),
remove the phrase ‘‘of part 52 of this
chapter’’ and add in its place the phrase
‘‘of part 52 of this chapter or under
subpart H of part 53 of this chapter’’.
■
§ 51.55
[Amended]
118. In § 51.55, in paragraph (a)
remove the phrase ‘‘of part 52 of this
chapter’’ and add in its place the phrase
‘‘of part 52 of this chapter or under
subpart H of part 53 of this chapter’’.
■ 119. In § 51.58, revise paragraph (b) to
read as follows:
■
§ 51.58 Environmental report—number of
copies; distribution.
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*
*
*
*
*
(b) Each applicant for a license to
manufacture a nuclear power reactor, or
for an amendment to a license to
manufacture, seeking approval of the
final design of the nuclear power reactor
under subpart F of part 52 of this
chapter or under subpart H of part 53 of
this chapter, shall submit to the
Commission an environmental report or
any supplement to an environmental
report in the manner specified in §§ 52.3
or 53.040 of this chapter. The applicant
shall maintain the capability to generate
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additional copies of the environmental
report or any supplement to the
environmental report for subsequent
distribution to parties and Boards in the
NRC proceeding; Federal, State, and
local officials; and any affected Indian
Tribes, in accordance with written
instructions issued by the Director,
Office of Nuclear Reactor Regulation.
■ 120. In § 51.77, revise paragraph (a)
introductory text to read as follows:
§ 51.77 Distribution of draft environmental
impact statement.
(a) In addition to the distribution
authorized by § 51.74, a copy of a draft
environmental statement for a licensing
action for a production or utilization
facility, except an action authorizing
issuance, amendment, or renewal of a
license to manufacture a nuclear power
reactor pursuant to 10 CFR part 52,
subpart F or 10 CFR part 53, subparts H
or I will also be distributed to:
*
*
*
*
*
§ 51.92
[Amended]
121. In § 51.92, in paragraph (b),
wherever it may appear, remove the
phrase ‘‘10 CFR part 52’’ and add in its
place the phrase ‘‘10 CFR parts 52 or
53’’.
■
§ 51.95
[Amended]
122. In § 51.95, in paragraph (c)
introductory text remove the phrase
‘‘under 10 CFR parts 52 or 54’’ and add
in its place the phrase ‘‘under 10 CFR
parts 52, 53, or 54’’.
■ 123. In § 51.101, revise paragraph
(a)(2) to read as follows:
■
§ 51.101
Limitations on actions.
(a) * * *
(2) Any action concerning the
proposal taken by an applicant which
would—
(i) Have an adverse environmental
impact, or
(ii) Limit the choice of reasonable
alternatives that may be grounds for
denial of the license. In the case of an
application covered by §§ 30.32(f),
40.31(f), 50.10(c), 53.1130, 70.21(f), or
72.16 and 72.34 of this chapter, the
provisions of this paragraph will be
applied in accordance with
§ 30.33(a)(5), 40.32(e), 50.10(c), 53.1130,
70.23(a)(7), or 72.40(b) of this chapter,
as appropriate.
*
*
*
*
*
§ 51.103
[Amended]
124. In § 51.103, in paragraph (a)(6)
remove the phrase ‘‘under 10 CFR
50.10’’ and add in its place the phrase
‘‘under §§ 50.10 or 53.1130 of this
chapter’’.
■
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125. In § 51.105, revise paragraph
(c)(1) introductory text to read as
follows:
■
§ 51.105 Public hearings in proceedings
for issuance of construction permits or
early site permits; limited work
authorizations.
*
*
*
*
*
(c)(1) In addition to complying with
the applicable provisions of § 51.104, in
any proceeding for the issuance of a
construction permit for a nuclear power
plant or an early site permit under parts
52 or 53 of this chapter, where the
applicant requests a limited work
authorization under §§ 50.10(d) or
53.1130 of this chapter, the presiding
officer will—
*
*
*
*
*
■ 126. In § 51.107, revise paragraphs (a)
introductory text, (b) introductory text,
and (d)(1) introductory text to read as
follows:
§ 51.107 Public hearings in proceedings
for issuance of combined licenses; limited
work authorizations.
(a) In addition to complying with the
applicable requirements of § 51.104, in
a proceeding for the issuance of a
combined license for a nuclear power
reactor under parts 52 or 53 of this
chapter, the presiding officer will:
*
*
*
*
*
(b) If a combined license application
references an early site permit, then the
presiding officer in the combined
license hearing must not admit any
contention proffered by any party on
environmental issues that have been
accorded finality under §§ 52.39 or
53.1188 of this chapter, unless the
contention:
*
*
*
*
*
(d)(1) In any proceeding for the
issuance of a combined license where
the applicant requests a limited work
authorization under §§ 50.10(d) or
§ 53.1130(a) of this chapter, the
presiding officer, in addition to
complying with any applicable
provision of § 51.104, will:
*
*
*
*
*
■ 127. Revise § 51.108 to read as
follows:
§ 51.108 Public hearings on Commission
findings that inspections, tests, analyses,
and acceptance criteria of combined
licenses are met.
In any public hearing requested under
§§ 52.103(b) or 53.1452(b) of this
chapter, the Commission will not admit
any contentions on environmental
issues, the adequacy of the
environmental impact statement for the
combined license issued under subpart
C of part 52 of this chapter or under
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subpart H of part 53 of this chapter, or
the adequacy of any other
environmental impact statement or
environmental assessment referenced in
the combined license application. The
Commission will not make any
environmental findings in connection
with the finding under § 52.103(g) or
§ 53.1452(g) of this chapter.
■ 128. Add part 53, consisting of
§§ 53.000 through 53.9010, to read as
follows:
PART 53—RISK-INFORMED,
TECHNOLOGY-INCLUSIVE
REGULATORY FRAMEWORK FOR
COMMERCIAL NUCLEAR PLANTS
Sec.
53.000
Purpose.
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Subpart B—Technology-Inclusive Safety
Requirements
53.210 Safety criteria for design-basis
accidents.
53.220 Safety criteria for licensing-basis
events other than design-basis accidents.
53.230 Safety functions.
53.240 Licensing-basis events.
53.250 Defense in depth.
53.260 Normal operations.
53.270 Protection of plant workers.
Subpart C—Design and Analysis
Requirements
53.400 Design features for licensing-basis
events.
53.410 Functional design criteria for
design-basis accidents.
53.415 Protection against external hazards.
53.420 Functional design criteria for
licensing-basis events other than designbasis accidents.
53.425 Design features and functional
design criteria for normal operations.
53.430 Design features and functional
design criteria for protection of plant
workers.
53.440 Design requirements.
53.450 Analysis requirements.
53.460 Safety categorization and treatments.
53.470 Maintaining analytical safety
margins used to justify operational
flexibilities.
53.480 Earthquake engineering.
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53.500
53.510
53.520
53.530
53.540
General siting and siting assessment.
External hazards.
Site characteristics.
Population-related considerations.
Siting interfaces.
Subpart E—Construction and
Manufacturing Requirements
53.600 Construction and manufacturing—
scope and purpose.
53.605 Reporting of defects and
noncompliance.
53.610 Construction.
53.620 Manufacturing.
Subpart F—Requirements for Operation
Subpart A—General Provisions
53.015 Scope.
53.020 Definitions.
53.030 Reserved.
53.040 Written communications.
53.050 Deliberate misconduct.
53.060 Employee protection.
53.070 Completeness and accuracy of
information.
53.080 Specific exemptions.
53.090 Standards for review.
53.100 Jurisdictional limits.
53.110 Attacks and destructive acts.
53.115 Rights related to special nuclear
material.
53.117 License suspension and rights of
recapture.
53.120 Information collection requirements:
OMB approval.
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Subpart D—Siting Requirements
53.700 Operational objectives.
53.710 Maintaining capabilities and
availability of structures, systems, and
components.
53.715 Maintenance, repair, and inspection
programs.
53.720 Response to seismic events.
53.725 General staffing, training, personnel
qualifications, and human factors
requirements.
53.726 Communications.
53.728 Completeness and accuracy of
information.
53.730 Defining, fulfilling, and maintaining
the role of personnel in ensuring safe
operations.
53.735 General exemptions.
53.740 Facility licensee requirements—
General.
53.745 Operator license requirements.
53.760 Operator licensing.
53.765 Medical requirements.
53.770 Incapacitation because of disability
or illness.
53.775 Applications for operators and
senior operators.
53.780 Training, examination, and
proficiency program.
53.785 Conditions of operator and senior
operator licenses.
53.790 Issuance, modification, and
revocation of operator and senior
operator licenses.
53.795 Expiration and renewal of operator
and senior operator licenses.
53.800 Facility licensees for self-reliantmitigation facilities.
53.805 Facility licensee requirements
related to generally licensed reactor
operators.
53.810 Generally licensed reactor operators.
53.815 Generally licensed reactor operator
training, examination, and proficiency
programs.
53.820 Cessation of individual
applicability.
53.830 Training and qualification of
commercial nuclear plant personnel.
53.845 Programs.
53.850 Radiation protection.
53.855 Emergency preparedness.
53.860 Security programs.
53.865 Quality assurance.
53.870 Integrity assessment programs.
53.875 Fire protection.
53.880 Inservice inspection and inservice
testing.
53.910 Procedures and guidelines.
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Subpart G—Decommissioning
Requirements
53.1000 Scope and purpose.
53.1010 Financial assurance for
decommissioning.
53.1020 Cost estimates for
decommissioning.
53.1030 Annual adjustments to cost
estimates for decommissioning.
53.1040 Methods for providing financial
assurance for decommissioning.
53.1045 Limitations on the use of
decommissioning trust funds.
53.1050 NRC oversight.
53.1060 Reporting and recordkeeping
requirements.
53.1070 Termination of license.
53.1075 Program requirements during
decommissioning.
53.1080 Release of part of a commercial
nuclear plant or site for unrestricted use.
Subpart H—Licenses, Certifications, and
Approvals
53.1100 Filing of application for licenses,
certifications, or approvals; oath or
affirmation.
53.1101 Requirement for license.
53.1103 Combining applications and
licenses.
53.1106 Elimination of repetition.
53.1109 Contents of applications; general
information.
53.1112 Environmental conditions.
53.1115 Agreement limiting access to
classified information.
53.1118 Ineligibility of certain applicants.
53.1120 Exceptions and exemptions from
licensing requirements.
53.1121 Public inspection of applications.
53.1124 Relationship between sections.
53.1130 Limited work authorizations.
53.1140 Early site permits.
53.1143 Filing of applications.
53.1144 Contents of applications for early
site permits; general information.
53.1146 Contents of applications for early
site permits; technical information.
53.1149 Review of applications.
53.1155 Referral to the Advisory Committee
on Reactor Safeguards.
53.1158 Issuance of early site permit.
53.1161 Extent of activities permitted.
53.1164 Duration of permit.
53.1167 Limited work authorization after
issuance of early site permit.
53.1170 Transfer of early site permit.
53.1173 Application for renewal.
53.1176 Criteria for renewal.
53.1179 Duration of renewal.
53.1182 Use of site for other purposes.
53.1188 Finality of early site permit
determinations.
53.1200 Standard design approvals.
53.1203 Filing of applications.
53.1206 Contents of applications for
standard design approvals; general
information.
53.1209 Contents of applications for
standard design approvals; technical
information.
53.1210 Contents of applications for
standard design approvals; other
application content.
53.1212 Standards for review of
applications.
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53.1215 Referral to the Advisory Committee
on Reactor Safeguards.
53.1218 Staff approval of design.
53.1221 Finality of standard design
approvals; information requests.
53.1230 Standard design certifications.
53.1233 Filing of applications.
53.1236 Contents of applications for
standard design certifications; general
information.
53.1239 Contents of applications for
standard design certifications; technical
information.
53.1241 Contents of applications for
standard design certifications; other
application content.
53.1242 Review of applications.
53.1245 Referral to the Advisory Committee
on Reactor Safeguards.
53.1248 Issuance of standard design
certification.
53.1251 Duration of certification.
53.1254 Application for renewal.
53.1257 Criteria for renewal.
53.1260 Duration of renewal.
53.1263 Finality of standard design
certifications.
53.1270 Manufacturing licenses.
53.1273 Filing of applications.
53.1276 Contents of applications for
manufacturing licenses; general
information.
53.1279 Contents of applications for
manufacturing licenses; technical
information.
53.1282 Contents of applications for
manufacturing licenses; other
application content.
53.1285 Review of applications.
53.1286 Referral to the Advisory Committee
on Reactor Safeguards.
53.1287 Issuance of manufacturing licenses.
53.1288 Finality of manufacturing licenses.
53.1291 Duration of manufacturing
licenses.
53.1293 Transfer of manufacturing licenses.
53.1295 Renewal of manufacturing licenses.
53.1300 Construction permits.
53.1306 Contents of applications for
construction permits; general
information.
53.1309 Contents of applications for
construction permits; technical
information.
53.1312 Contents of applications for
construction permits; other application
content.
53.1315 Review of applications.
53.1318 Finality of referenced NRC
approvals, permits, and certifications.
53.1324 Referral to the Advisory Committee
on Reactor Safeguards.
53.1327 Authorization to conduct limited
work authorization activities.
53.1330 Exemptions, departures, and
variances.
53.1333 Issuance of construction permits.
53.1336 Finality of construction permits.
53.1342 Duration of construction permits.
53.1345 Transfer of construction permits.
53.1348 Termination of construction
permits.
53.1360 Operating licenses.
53.1366 Contents of applications for
operating licenses; general information.
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53.1369 Contents of applications for
operating licenses; technical
information.
53.1372 Contents of applications for
operating licenses; other application
content.
53.1375 Review of applications.
53.1381 Referral to the Advisory Committee
on Reactor Safeguards.
53.1384 Exemptions, departures, and
variances.
53.1387 Issuance of operating licenses.
53.1390 Backfitting of operating licenses.
53.1396 Duration of operating licenses.
53.1399 Transfer of an operating license.
53.1402 Application for renewal.
53.1405 Continuation of an operating
license.
53.1410 Combined licenses.
53.1413 Contents of applications for
combined licenses; general information.
53.1416 Contents of applications for
combined licenses; technical
information.
53.1419 Contents of applications for
combined licenses; other application
content.
53.1422 Review of applications.
53.1425 Finality of referenced NRC
approvals.
53.1431 Referral to the Advisory Committee
on Reactor Safeguards.
53.1434 Authorization to conduct limited
work authorization activities.
53.1437 Exemptions, departures, and
variances.
53.1440 Issuance of combined licenses.
53.1443 Finality of combined licenses.
53.1449 Inspection during construction.
53.1452 Operation under a combined
license.
53.1455 Duration of combined license.
53.1456 Transfer of a combined license.
53.1458 Application for renewal.
53.1461 Continuation of combined license.
53.1470 Standardization of commercial
nuclear plant designs: licenses to
construct and operate nuclear power
reactors of identical design at multiple
sites.
Subpart I—Maintaining and Revising
Licensing-Basis Information
53.1500 Licensing-basis information.
53.1502 Specific terms and conditions of
licenses.
53.1505 Changes to licensing-basis
information requiring prior NRC
approval.
53.1510 Application for amendment of
license.
53.1515 Public notices; State consultation.
53.1520 Issuance of amendment.
53.1525 Revising certification information
within a design certification rule.
53.1530 Revising design information within
a manufacturing license.
53.1535 Amendments during construction.
53.1540 Updating licensing-basis
information and determining the need
for NRC approval.
53.1545 Updating Final Safety Analysis
Reports.
53.1550 Evaluating changes to facility as
described in Final Safety Analysis
Reports.
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53.1560 Updating program documents
included in licensing-basis information.
53.1565 Evaluating changes to programs
included in licensing-basis information.
53.1570 Transfer of licenses.
53.1575 Termination of licenses.
53.1580 Information requests.
53.1585 Revocation, suspension,
modification of licenses and approvals
for cause.
53.1590 Backfitting.
53.1595 Renewal.
Subpart J—Reporting and Other
Administrative Requirements
53.1600 General information.
53.1610 Unfettered access for inspections.
53.1620 Maintenance of records, making of
reports.
53.1630 Immediate notification
requirements for operating commercial
nuclear plants.
53.1640 Licensee event report system.
53.1645 Reports of radiation exposure to
members of the public.
53.1650 Facility information and
verification.
53.1660 Financial requirements.
53.1670 Financial qualifications.
53.1680 Annual financial reports.
53.1690 Licensee’s change of status;
financial qualifications.
53.1700 Creditor regulations.
53.1710 Financial protection.
53.1720 Insurance required to stabilize and
decontaminate plant following an
accident.
53.1730 Financial protection requirements.
Subpart M—Enforcement
53.9000 Violations.
53.9010 Criminal penalties.
Authority: Atomic Energy Act of 1954,
secs. 11, 101, 103, 108, 122, 147, 161, 181,
182, 183, 184, 185, 186, 187, 189, 223, 234
(42 U.S.C. 2014, 2131, 2132, 2133, 2134,
2135, 2138, 2152, 2167, 2169, 2201, 2231,
2232, 2233, 2234, 2235, 2236, 2237, 2239,
2273, 2282); Energy Reorganization Act of
1974, secs. 201, 202, 206, 211 (42 U.S.C.
5841, 5842, 5846, 5851); Nuclear Waste
Policy Act of 1982, sec. 306 (42 U.S.C.
10226); National Environmental Policy Act of
1969 (42 U.S.C. 4332); 44 U.S.C. 3504 note;
Sec. 109, Pub. L. 96–295, 94 Stat. 783; Pub.
L. 115–439, 132 Stat. 5571.
§ 53.000
Purpose.
This part provides an optional
technology-inclusive, performancebased framework for the issuance,
amendment, renewal, and termination
of licenses, permits, certifications, and
approvals for commercial nuclear plants
licensed under section 103 of the
Atomic Energy Act of 1954, as amended
(the Act)(68 Stat. 919), and Title II of the
Energy Reorganization Act of 1974, as
amended (88 Stat. 1242). Also, this part
gives notice to all persons who
knowingly provide to any holder of or
applicant for an approval, certification,
permit, or license, or to a contractor,
subcontractor, or consultant of any of
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them, components, equipment,
materials, or other goods or services that
relate to the activities of a holder of or
applicant for an approval, certification,
permit, or license, subject to this part,
that they may be individually subject to
U.S. Nuclear Regulatory Commission
enforcement action for violation of the
provisions in § 53.050.
Subpart A—General Provisions
§ 53.015
Scope.
Subpart A provides general provisions
applicable to all applicants and
licensees subject to the rules of this part.
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§ 53.020
Definitions.
For the purpose of this part:
Anticipated event sequence means
event sequences expected to occur one
or more times during the life of a
commercial nuclear plant. Anticipated
event sequences take into account the
expected response of all structures,
systems, and components (SSCs) within
the plant, regardless of safety
classification.
Applicant means a person applying
for a license, permit, or other form of
Commission permission or approval
under this part.
Certified fuel handler means, for a
commercial nuclear plant, either—
(1) A non-licensed operator who has
qualified in accordance with a fuel
handler training program approved by
the Commission; or
(2) A non-licensed operator who
demonstrates compliance with the
following criteria:
(i) Has qualified in accordance with a
fuel handler training program that
demonstrates compliance with the same
requirements as training programs for
non-licensed operators required by
§ 53.830, and
(ii) Is responsible for decisions on—
(A) Safe conduct of decommissioning
activities,
(B) Safe handling and storage of spent
fuel, and
(C) Appropriate response to plant
emergencies.
Combined license (COL) means a
combined construction permit (CP) and
operating license (OL) with conditions
for a commercial nuclear plant issued
under this part.
Commercial nuclear plant means a
facility consisting of one or more
commercial nuclear reactors and
associated co-located support facilities,
including the collection of buildings,
radionuclide sources, and SSCs for
which a license, certification, or
approval is being sought under this part,
that is or will be used for producing
power for commercial electric power or
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other commercial purposes. For the
purposes of requirements in this part
that reference requirements in part 50 of
this chapter, a commercial nuclear plant
is equivalent to a nuclear power plant.
Commercial nuclear reactor means an
apparatus, other than an atomic
weapon, designed or used to sustain
nuclear fission. For the purposes of
requirements in this part that reference
requirements in 10 CFR part 50, a
commercial nuclear reactor is
equivalent to a nuclear reactor as
defined in 10 CFR 50.2.
Commission means the U.S. Nuclear
Regulatory Commission (NRC) or its
duly authorized representatives.
Consensus code or standard means
any technical standard that is—
(1) Developed or adopted by a
voluntary consensus standard body
under procedures that assure that
persons having interests within the
scope of the standard that are affected
by the provisions of the standard have
reached substantial agreement on its
adoption;
(2) Formulated in a manner that
afforded an opportunity for diverse
views to be considered; and
(3) Designated by the standards body
as a consensus code or standard.
Construction means the activities in
paragraph (1) below and does not mean
the activities in paragraph (2) below.
(1) Activities constituting
construction are those activities credited
or relied upon for demonstrating
compliance with the safety criteria
defined in subpart B of this part which
are conducted on-site to build the
commercial nuclear plant, including the
driving of piles; subsurface preparation;
placement of backfill, concrete, or
permanent retaining walls within an
excavation; installation of foundations;
or in-place assembly, erection,
fabrication, or testing, which are for—
(i) Safety-related (SR) and non-safetyrelated but safety-significant (NSRSS)
SSCs of a facility;
(ii) SSCs necessary to comply with 10
CFR part 73; or
(iii) Onsite emergency facilities
necessary to comply with § 53.855.
(2) Construction does not include—
(i) Changes for temporary use of the
land for public recreational purposes;
(ii) Site exploration, including
necessary borings to determine
foundation conditions or other
preconstruction monitoring to establish
background information related to the
suitability of the site, the environmental
impacts of construction or operation, or
the protection of environmental values;
(iii) Preparation of a site for
construction of a facility, including
clearing of the site, grading, installation
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of drainage, erosion, and other
environmental mitigation measures, and
construction of temporary roads and
borrow areas;
(iv) Erection of fences and other
access control measures;
(v) Excavation;
(vi) Erection of support buildings
(such as construction equipment storage
sheds, warehouse and shop facilities,
utilities, concrete mixing plants,
docking and unloading facilities, and
office buildings) for use in connection
with the construction of the facility;
(vii) Building of service facilities
(such as paved roads, parking lots,
railroad spurs, exterior utility and
lighting systems, potable water systems,
sanitary sewage treatment facilities, and
transmission lines);
(viii) Procurement or fabrication of
components or portions of the proposed
facility occurring at locations other than
the final, in-place location at the
facility; or
(ix) Manufacture of a nuclear power
reactor under a manufacturing license
(ML) under subpart H of this part to be
installed at the proposed site and to be
part of the proposed facility.
Custom combined license (custom
COL) means a COL that does not
reference a standard design approval or
design certification.
Decommission or decommissioning
means to remove a plant or site safely
from service and reduce residual
radioactivity to a level that permits—
(1) Release of the property for
unrestricted use and termination of the
license; or
(2) Release of the property under
restricted conditions and termination of
the license.
Defense in depth means inclusion of
two or more independent and
redundant layers of defense in the
design of a facility and its operating
procedures to compensate for
uncertainties such that no single layer of
defense, no matter how robust, is
exclusively relied upon. Defense in
depth includes, but is not limited to, the
use of access controls, physical barriers,
redundant and diverse safety functions,
and emergency response measures.
Design-basis accidents (DBAs) means
postulated event sequences that are
used to set functional design criteria
and performance objectives for the
design of SR SSCs through deterministic
analyses. Design-basis accidents are a
type of licensing-basis event and are
based on the capabilities and
reliabilities of SR SSCs needed to
mitigate and prevent event sequences,
respectively.
Design-basis external hazard level
means the level of severity or intensity
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of an external hazard for which the SR
SSCs are protected against or designed
to withstand without losing their
capability to perform their safety
functions.
Design features means the active and
passive SSCs and the inherent
characteristics of those SSCs that
contribute to limiting the total effective
dose equivalent to individual members
of the public during normal operations
and prevent or mitigate the
consequences of event sequences.
Electric utility means any entity that
generates or distributes electricity and
that recovers the cost of this electricity,
either directly or indirectly, through
rates established by the entity itself or
by a separate regulatory authority.
Investor-owned utilities, including
generation or distribution subsidiaries,
public utility districts, municipalities,
rural electric cooperatives, and State
and Federal agencies, including
associations of any of the foregoing, are
included within the meaning of
‘‘electric utility.’’
Event sequence means a postulated
initiating event defined for a set of
initial plant conditions followed by
system, safety function, and operator
successes or failures, and terminating in
a specified end state depending on the
system, safety function, and operator
successes and failures (e.g., prevention
of release of radioactive material or
release in one of the reactor-specific
release categories). An event sequence
may include many unique variations of
events that are similar in terms of
results or end states.
Exclusion area means that area
surrounding the reactor, in which the
reactor licensee has the authority to
determine all activities including
exclusion or removal of personnel and
property from the area. This area may be
traversed by a highway, railroad, or
waterway, provided these are not so
close to the facility as to interfere with
normal operations of the facility and
provided appropriate and effective
arrangements are made to control traffic
on the highway, railroad, or waterway,
in case of emergency, to protect the
public health and safety. Residence
within the exclusion area must normally
be prohibited. In any event, residents
must be subject to ready removal in case
of necessity. Activities unrelated to
operation of the reactor may be
permitted in an exclusion area under
appropriate limitations, provided that
no significant hazards to the public
health and safety will result.
Fission product release means the
amount and composition of radioactive
material released to the environment,
after accounting for any retention of
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radionuclides provided by reactor
design features.
Fuel means special nuclear material
(SNM) or source material, discrete
elements that physically contain SNM
or source material, and homogeneous
mixtures that contain SNM or source
material, intended to or used to create
power in a commercial nuclear plant.
Functional design criteria means
metrics for the performance of SSCs. For
SR SSCs, these criteria define
performance metrics necessary to
demonstrate compliance with the safety
criteria in § 53.210. For NSRSS SSCs,
these criteria define performance
metrics necessary to demonstrate
compliance with the safety criteria in
§ 53.220.
License, when used in the context of
a facility, means a limited work
authorization, CP, OL, early site permit,
COL, or ML under this part, or a
renewed license issued by the
Commission under this part. When used
in the context of a license authorizing
an individual to manipulate the controls
of a facility, license means a license
issued by the Commission to perform
the function of an operator, senior
operator, or generally licensed reactor
operator as defined in this part.
Licensee means a person who is
authorized to conduct activities under a
license issued under this part by the
Commission.
Licensing-basis events means a
collection of event sequences
considered in the design and licensing
of the commercial nuclear plant.
Licensing-basis events are unplanned
events and include anticipated event
sequences, unlikely event sequences,
very unlikely event sequences, and
DBAs.
Licensing-basis information means the
information contained in regulations,
orders, licenses, certifications, or
approvals issued by the NRC for a
commercial nuclear plant licensed
under this part and that information
submitted to the NRC by an applicant or
licensee in a Safety Analysis Report,
program description, or other licensingrelated document required under this
part.
Low-population zone means the area
immediately surrounding the exclusion
area which contains residents, the total
number and density of which are such
that there is a reasonable probability
that appropriate protective measures
could be taken on their behalf in the
event of a serious accident. A
permissible population density or total
population within this zone is not
included in this definition because the
situation may vary from case to case.
Whether a specific number of people
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can, for example, be evacuated from a
specific area or instructed to take shelter
on a timely basis, will depend on many
factors such as location, number and
size of highways, scope and extent of
advance planning, and actual
distribution of residents within the area.
Major decommissioning activity
means, for a commercial nuclear plant,
any activity that results in permanent
removal of major radioactive
components, permanently modifies the
structure of the containment, if
applicable, or results in dismantling
components for shipment containing
greater than class C waste in accordance
with 10 CFR 61.55.
Major feature of the emergency plans
means an aspect of those plans
necessary to:
(1) Address in whole or part either
one or more of the 16 standards in 10
CFR 50.47(b) or the requirements of 10
CFR 50.160(b), as applicable; or
(2) Describe the emergency planning
zones as required in § 53.1109(g).
Manufactured reactor means the
essential portions of a nuclear reactor
that are manufactured under an ML and
subsequently transported and
incorporated into a commercial nuclear
plant under a COL.
Manufacturing license means a
license issued under this part that
authorizes the manufacture of
manufactured reactors but not its
construction, installation, or operation.
Non-Safety-Related but SafetySignificant (NSRSS) SSCs means those
SSCs which are not SR but are relied on
to achieve adequate defense in depth or
perform risk-significant functions and
warrant special treatment.
Non-Safety-Significant SSCs means
those SSCs that are not SR or NSRSS,
are not relied on to achieve adequate
defense in depth or to perform risksignificant functions, and do not
warrant special treatment.
Person means—
(1) any individual, corporation,
partnership, firm, association, trust,
estate, public or private institution,
group, government agency other than
the Commission or the Department,
except that the Department shall be
considered a person to the extent that its
facilities are subject to the licensing and
related regulatory authority of the
Commission pursuant to section 202 of
the ERA, any State or any political
subdivision of, or any political entity
within a State, any foreign government
or nation or any political subdivision of
any such government or nation, or other
entity; and
(2) any legal successor, representative,
agent, or agency of the foregoing.
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Population center distance means the
distance from the reactor to the nearest
boundary of a densely populated center
containing more than about 25,000
residents.
Probabilistic risk assessment means a
quantitative assessment of the risk
associated with plant operation and
maintenance that is measured in terms
of event sequence occurrence
frequencies and consequences.
Programmatic controls means
administrative procedures that govern
human action in implementing
programs and operating, monitoring,
and maintaining SSCs and equipment of
a commercial nuclear plant.
Programmatic controls considered to be
licensing basis information are specified
in an application for a requested activity
of the Commission.
Quality assurance (QA) means all
those planned and systematic actions
necessary to ensure that a structure,
system, or component will perform
satisfactorily in service. Quality
assurance includes quality control,
which comprises those QA actions
related to the physical characteristics of
a material, structure, component, or
system which provide a means to
control the quality of the material,
structure, component, or system to
predetermined requirements.
Safety criteria means performancebased metrics that establish a level of
safety provided in requirements in
§§ 53.210 and 53.220.
Safety-related structures, systems, or
components means those SSCs that are
relied upon to demonstrate compliance
with the safety criteria in § 53.210 and
warrant special treatment.
Small modular reactor means a power
reactor, which may be of modular
design as defined in 10 CFR 52.1,
licensed under this part to produce heat
energy up to 1,000 megawatts thermal
per module.
Site characteristics means the actual
physical, environmental, and
demographic features of a site. Site
characteristics are specified in an early
site permit or in a Preliminary or Final
Safety Analysis Report for a limited
work authorization, CP, or COL, as
applicable.
Site parameters are the postulated
physical, environmental, and
demographic features of an assumed
site. Site parameters are specified in a
standard design approval, standard
design certification, or ML.
Source material means source
material as defined in subsection 11z. of
the Atomic Energy Act of 1954, as
amended, (the Act) and in the
regulations contained in part 40 of this
chapter.
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Special nuclear material (SNM)
means:
(1) Plutonium, uranium-233, uranium
enriched in the isotope-233 or in the
isotope-235, and any other material
which the Commission, pursuant to the
provisions of section 51 of the Act,
determines to be SNM, but does not
include source material; or
(2) Any material artificially enriched
by any of the foregoing, but does not
include source material.
Special treatment means those
requirements, such as QA and
programmatic controls, that ensure that
SR and NSRSS SSCs will provide
defense in depth or perform risksignificant functions. The requirements
also ensure that the SSCs will perform
under the service conditions and with
the reliability assumed in the analysis
performed under § 53.450 to
demonstrate compliance with the safety
criteria in §§ 53.210 and 53.220.
Standard design means a design
which is sufficiently detailed and
complete to support certification or
approval in accordance with subpart H
of this part, and which is usable under
of this part for a multiple number of
units or at a multiple number of sites
without reopening or repeating the
review.
Standard design approval or design
approval means an NRC staff approval,
issued under subpart H of this part, of
a final standard design for a commercial
nuclear plant. The approval may be for
either the final design for the entire
reactor facility or the final design of
major portions thereof.
Standard design certification or
design certification means a
Commission approval, issued under
subpart H of this part, of a final standard
design for a nuclear power facility. This
design may be referred to as a certified
standard design.
Total effective dose equivalent means
the sum of the effective dose equivalent
(for external exposures) and the
committed effective dose equivalent (for
internal exposures).
Utilization facility means any
commercial nuclear reactor other than
one designed or used primarily for the
formation of plutonium or uranium-233.
Unlikely event sequences means event
sequences that are not expected to occur
in the life of a commercial nuclear plant
and are less likely than anticipated
event sequences, but are infrequent
rather than rare. Unlikely event
sequences take into account the
expected response of all SSCs within
the plant regardless of safety
classification.
Very unlikely event sequences means
event sequences that are not expected to
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occur in the life of a commercial nuclear
plant, are less likely than an unlikely
event sequence, and are rare. Very
unlikely event sequences take into
account the expected response of all
SSCs within the plant regardless of
safety classification.
§ 53.030
[Reserved]
§ 53.040
Written communications.
(a) General requirements. All
correspondence, reports, applications,
and other written communications from
the applicant or licensee to the NRC
concerning the regulations in this part
or individual license conditions must be
sent either by mail addressed: ATTN:
Document Control Desk, U.S. Nuclear
Regulatory Commission, Washington,
DC 20555–0001; by hand delivery to the
NRC’s offices at 11555 Rockville Pike,
Rockville, Maryland, between the hours
of 8:15 a.m. and 4 p.m. eastern time; or,
where practicable, by electronic
submission, for example, via Electronic
Information Exchange, email, or CD–
ROM. Electronic submissions must be
made in a manner that enables the NRC
to receive, read, authenticate, distribute,
and archive the submission, and process
and retrieve it a single page at a time.
Detailed guidance on making electronic
submissions can be obtained by visiting
the NRC’s website at https://
www.nrc.gov/site-help/esubmittals.html; by email to
MSHD.Resource@nrc.gov; or by writing
the Office of the Chief Information
Officer, U.S. Nuclear Regulatory
Commission, Washington, DC 20555–
0001. The guidance discusses, among
other topics, the formats the NRC can
accept, the use of electronic signatures,
and the treatment of nonpublic
information. If the communication is on
paper, the signed original must be sent.
If a submission due date falls on a
Saturday, Sunday, or Federal holiday,
the next Federal working day becomes
the official due date.
(b) Distribution requirements. Copies
of all correspondence, reports, and other
written communications concerning the
regulations in this part or individual
license conditions, or the terms and
conditions of an early site permit or
standard design approval, must be
submitted to the persons listed below
(addresses for the NRC Regional Offices
are listed in appendix D to 10 CFR part
20).
(1) Applications for amendment of
permits and licenses, reports, and other
communications. All written
communications (including responses to
generic letters, bulletins, information
notices, regulatory information
summaries, inspection reports, and
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miscellaneous requests for additional
information) that are required of holders
of licenses, permits, and design
approvals issued pursuant to this part,
must be submitted as follows, except as
otherwise specified in paragraphs (b)(2)
through (7) of this section: to the NRC’s
Document Control Desk (if on paper, the
signed original), with a copy to the
appropriate Regional Office, and a copy
to the appropriate NRC Resident
Inspector if one has been assigned to the
site of the facility or the place of
manufacture of a reactor licensed under
this part.
(2) Applications for permits and
licenses, and amendments to
applications. Applications for licenses,
permits, and design approvals and
amendments to any of these types of
applications must be submitted to the
NRC’s Document Control Desk, with a
copy to the appropriate Regional Office,
and a copy to the appropriate NRC
Resident Inspector if one has been
assigned to the facility or the place of
manufacture of a reactor licensed under
this part, except as otherwise specified
in paragraphs (b)(3) through (9) of this
section. If the application or amendment
is on paper, the submission to the
Document Control Desk must be the
signed original.
(3) Acceptance review application.
Written communications required for an
application for determination of
suitability for docketing must be
submitted to the NRC’s Document
Control Desk, with a copy to the
appropriate Regional Office. If the
communication is on paper, the
submission to the Document Control
Desk must be the signed original.
(4) Security plan and related
submissions. Written communications,
as defined in paragraphs (b)(4)(i)
through (v) of this section, must be
submitted to the NRC’s Document
Control Desk, with a copy to the
appropriate Regional Office. If the
communication is on paper, the
submission to the Document Control
Desk must be the signed original.
Submissions should include the
following as appropriate:
(i) Physical security plan;
(ii) Safeguards contingency plan;
(iii) Cybersecurity plan;
(iv) Change to security plan, guard
training and qualification plan,
safeguards contingency plan, or
cybersecurity plan made without prior
Commission approval under § 53.1565;
and
(v) Application for amendment of
physical security plan, guard training
and qualification plan, safeguards
contingency plan, or cybersecurity plan
under § 53.1510.
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(5) Emergency plan and related
submissions. Written communications
as defined in paragraphs (b)(5)(i)
through (iii) of this section must be
submitted to the NRC’s Document
Control Desk, with a copy to the
appropriate Regional Office, and a copy
to the appropriate NRC Resident
Inspector if one has been assigned to the
site of the facility. If the communication
is on paper, the submission to the
Document Control Desk must be the
signed original. Submissions should
include the following as appropriate:
(i) Emergency plan;
(ii) Change to an emergency plan
under § 53.1565; and
(iii) Emergency implementing
procedures under § 53.855.
(6) Updated Final Safety Analysis
Report. An Updated Final Safety
Analysis Report or replacement pages
under § 53.1545 must be submitted to
the NRC’s Document Control Desk, with
a copy to the appropriate Regional
Office, and a copy to the appropriate
NRC Resident Inspector if one has been
assigned to the site of the facility or the
place of manufacture of a reactor
licensed under this part. Paper copy
submissions may be made using
replacement pages; however, if a
licensee chooses to use electronic
submission, all subsequent updates or
submissions must be performed
electronically on a total replacement
basis. If the communication is on paper,
the submission to the Document Control
Desk must be the signed original. If the
communications are submitted
electronically, see Guidance for
Electronic Submissions to the
Commission.
(7) Quality assurance related
submissions. (i) A change to the Safety
Analysis Report QA program
description under § 53.1565, or a change
to a licensee’s NRC-accepted QA topical
report under § 53.1565, must be
submitted to the NRC’s Document
Control Desk, with a copy to the
appropriate Regional Office, and a copy
to the appropriate NRC Resident
Inspector if one has been assigned to the
site of the facility or the place of
manufacture of a reactor licensed under
this part. If the communication is on
paper, the submission to the Document
Control Desk must be the signed
original.
(ii) A change to an NRC-accepted QA
topical report from non-licensees (i.e.,
architect/engineers, nuclear steam
supply system suppliers, fuel suppliers,
constructors, etc.) must be submitted to
the NRC’s Document Control Desk. If
the communication is on paper, the
signed original must be sent.
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(8) Certification of permanent
cessation of operations. The licensee’s
certification of permanent cessation of
operations, under subpart G of this part,
must state the date on which operations
have ceased or will cease, and must be
submitted to the NRC’s Document
Control Desk. This submission must be
under oath or affirmation.
(9) Certification of permanent fuel
removal. The licensee’s certification of
permanent fuel removal, under subpart
G of this part, must state the date on
which the fuel was removed from the
reactor vessel and the disposition of the
fuel, and must be submitted to the
NRC’s Document Control Desk. This
submission must be under oath or
affirmation.
(c) Form of communications. All
paper copies submitted to demonstrate
compliance with the requirements set
forth in paragraph (b) of this section
must be typewritten, printed, or
otherwise reproduced in permanent
form on unglazed paper. Exceptions to
these requirements imposed on paper
submissions may be granted for the
submission of micrographic,
photographic, or similar forms.
(d) Regulation governing submission.
Licensees, applicants, and holders of
standard design approvals submitting
correspondence, reports, and other
written communications under the
regulations of this part are requested but
not required to cite whenever practical,
in the upper right corner of the first
page of the submission, the specific
regulation or other basis requiring
submission.
§ 53.050
Deliberate misconduct.
(a) Any licensee or applicant for a
license; holder of or applicant for a
standard design approval; applicant for
a standard design certification;
employee of a licensee, holder of a
standard design approval, or applicant
for a license, standard design approval,
or standard design certification; or any
contractor (including a supplier or
consultant), subcontractor, employee of
a contractor or subcontractor of any
licensee or applicant for a license,
holder of or applicant for a standard
design approval, or applicant for a
standard design certification, who
knowingly provides to any licensee,
applicant, contractor, or subcontractor,
any components, equipment, materials,
or other goods or services that relate to
a licensee’s or applicant’s activities in
this part, may not—
(1) Engage in deliberate misconduct
that causes or would have caused, if not
detected, a licensee or applicant to be in
violation of any rule, regulation, or
order; or any term, condition, or
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limitation of any license issued by the
Commission; or
(2) Deliberately submit to the NRC, a
licensee, an applicant, or a licensee’s or
applicant’s contractor or subcontractor,
information that the person submitting
the information knows to be incomplete
or inaccurate in some respect material to
the NRC.
(b) A person who violates paragraph
(a)(1) or (2) of this section may be
subject to enforcement action in
accordance with the procedures in
subpart B of 10 CFR part 2.
(c) For the purposes of paragraph
(a)(1) of this section, deliberate
misconduct by a person means an
intentional act or omission that the
person knows—
(1) Would cause a licensee or
applicant to be in violation of any rule,
regulation, or order; or any term,
condition, or limitation, of any license
issued by the Commission; or
(2) Constitutes a violation of a
requirement, procedure, instruction,
contract, purchase order, or policy of a
licensee, applicant, contractor, or
subcontractor.
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§ 53.060
Employee protection.
(a) Discrimination by a Commission
licensee, holder of a standard design
approval, an applicant for a license,
standard design certification, or
standard design approval, a contractor
or subcontractor of a Commission
licensee, holder of a standard design
approval, applicant for a license,
standard design certification, or
standard design approval, against an
employee for engaging in certain
protected activities is prohibited.
Discrimination includes discharge and
other actions that relate to
compensation, terms, conditions, or
privileges of employment. The protected
activities are established in section 211
of the Energy Reorganization Act of
1974, as amended, and in general are
related to the administration or
enforcement of a requirement imposed
under the Act or the Energy
Reorganization Act of 1974, as
amended.
(1) The protected activities include
but are not limited to—
(i) Providing the Commission or his or
her employer information about alleged
violations of either of the statutes
named in paragraph (a) of this section
or possible violations of requirements
imposed under either of those statutes;
(ii) Refusing to engage in any practice
made unlawful under either of the
statutes named in paragraph (a) of this
section or under these requirements if
the employee has identified the alleged
illegality to the employer;
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(iii) Requesting the NRC to institute
action against his or her employer for
the administration or enforcement of
these requirements;
(iv) Testifying in any Commission
proceeding, or before Congress, or at any
Federal or State proceeding regarding
any provision (or proposed provision) of
either of the statutes named in
paragraph (a) of this section; and
(v) Assisting or participating in, or
being about to assist or participate in,
these activities.
(2) These activities are protected even
if no formal proceeding is actually
initiated as a result of the employee
assistance or participation.
(3) This section has no application to
any employee alleging discrimination
prohibited by this section who, acting
without direction from his or her
employer (or the employer’s agent),
deliberately causes a violation of any
requirement of the Energy
Reorganization Act of 1974, as
amended, or the Act.
(b) Any employee who believes that
they have been discharged or otherwise
discriminated against by any person for
engaging in protected activities
specified in paragraph (a)(1) of this
section may seek a remedy for the
discharge or discrimination through an
administrative proceeding in the
Department of Labor. The
administrative proceeding must be
initiated within 180 days after an
alleged violation occurs. The employee
may do this by filing a complaint
alleging the violation with the
Department of Labor, Wage and Hour
Division. The Department of Labor may
order reinstatement, back pay, and
compensatory damages.
(c) A violation of paragraph (a), (e), or
(f) of this section by a Commission
licensee, a holder of a standard design
approval, an applicant for a Commission
license, standard design certification, or
a standard design approval, or a
contractor or subcontractor of a
Commission licensee, holder of a
standard design approval, or any
applicant may be grounds for—
(1) Denial, revocation, or suspension
of the license or standard design
approval;
(2) Withdrawal or revocation of a
proposed or final standard design
certification;
(3) Imposition of a civil penalty on the
licensee, holder of a standard design
approval, or applicant (including an
applicant for a standard design
certification under this part following
Commission adoption of final design
certification rule) or a contractor or
subcontractor of the licensee, holder of
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a standard design approval, or
applicant; or
(4) Other enforcement action.
(d) Actions taken by an employer, or
others, which adversely affect an
employee may be predicated upon
nondiscriminatory grounds. The
prohibition applies when the adverse
action occurs because the employee has
engaged in protected activities. An
employee’s engagement in protected
activities does not automatically render
him or her immune from discharge or
discipline for legitimate reasons or from
adverse action dictated by
nonprohibited considerations.
(e)(1) Each holder or applicant for a
license or design approval, must
prominently post the revision of NRC
Form 3, ‘‘Notice to Employees,’’
referenced in § 19.11(e)(1) of this
chapter. This form must be posted at
locations sufficient to permit employees
protected by this section to observe a
copy on the way to or from their place
of work. Premises must be posted no
later than 30 days after an application
is docketed and remain posted while the
application is pending before the
Commission, during the term of the
license, and for 30 days following
license termination.
(2) Copies of NRC Form 3 may be
obtained by writing to the Regional
Administrator of the appropriate NRC
Regional Office listed in appendix D to
10 CFR part 20, via email to
Forms.Resource@nrc.gov, or by visiting
the NRC’s online library at https://
www.nrc.gov/reading-rm/doccollections/forms/.
(f) No agreement affecting the
compensation, terms, conditions, or
privileges of employment, including an
agreement to settle a complaint filed by
an employee with the Department of
Labor pursuant to section 211 of the
Energy Reorganization Act of 1974, as
amended, may contain any provision
which would prohibit, restrict, or
otherwise discourage an employee from
participating in protected activity as
defined in paragraph (a)(1) of this
section including, but not limited to,
providing information to the NRC or to
his or her employer on potential
violations or other matters within NRC’s
regulatory responsibilities.
(g) Part 19 of 10 CFR sets forth
requirements and regulatory provisions
applicable to licensees, holders of a
standard design approval, applicants for
a license, standard design certification,
or standard design approval, and
contractors or subcontractors of a
Commission licensee, or holder of a
standard design approval, and are in
addition to the requirements in this
section.
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§ 53.070 Completeness and accuracy of
information.
(a) Information provided to the
Commission by a holder of a license,
permit, design certification, or standard
design approval under this part or an
applicant for a license, permit, design
certification, or standard design
approval under this part, and
information required by statute or by the
Commission’s regulations, orders,
license conditions, or terms and
conditions of a standard design
approval to be maintained by the
applicant or the licensee must be
complete and accurate in all material
respects.
(b) Each applicant or licensee, each
holder of a standard design approval
under this part, and each applicant for
a standard design certification under
this part following Commission
adoption of a final design certification
regulation, must notify the Commission
of information identified by the
applicant or licensee as having for the
regulated activity a significant
implication for public health and safety
or common defense and security. An
applicant, licensee, or holder violates
this paragraph only if the applicant,
licensee, or holder fails to notify the
Commission of information that the
applicant, licensee, or holder has
identified as having a significant
implication for public health and safety
or common defense and security.
Notification must be provided to the
Administrator of the appropriate
Regional Office within 2 working days
of identifying the information. This
requirement is not applicable to
information which is already required to
be provided to the Commission by other
reporting or updating requirements.
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§ 53.080
Specific exemptions.
(a) The Commission may, upon
application by any interested person or
upon its own initiative, grant
exemptions from the requirements of
the regulations of this part, which are
authorized by law, will not present an
undue risk to the public health and
safety, and are consistent with the
common defense and security.
(b) The Commission will not consider
granting an exemption unless special
circumstances are present. Special
circumstances are present whenever—
(1) Application of the regulation in
the particular circumstances conflicts
with other rules or requirements of the
Commission;
(2) Application of the regulation in
the particular circumstances would not
serve the underlying purpose of the rule
or is not necessary to achieve the
underlying purpose of the rule;
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(3) Compliance would result in undue
hardship or other costs that are
significantly in excess of those
contemplated when the regulation was
adopted, or that are significantly in
excess of those incurred by others
similarly situated;
(4) The exemption would result in
benefit to the public health and safety
that compensates for any decrease in
safety that may result from the grant of
the exemption;
(5) The exemption would provide
only temporary relief from the
applicable regulation and the licensee or
applicant has made good faith efforts to
comply with the regulation; or
(6) There is present any other material
circumstance not considered when the
regulation was adopted for which it
would be in the public interest to grant
an exemption. If such condition is relied
on exclusively for demonstrating
compliance with paragraph (b) of this
section, the exemption may not be
granted until the Executive Director for
Operations has consulted with the
Commission.
(c) Any person may request an
exemption permitting the conduct of
construction activities prior to the
issuance of a CP. The Commission may
grant such an exemption upon
considering and balancing the following
factors:
(1) Whether conduct of the proposed
activities will give rise to a significant
adverse impact on the environment and
the nature and extent of such impact, if
any;
(2) Whether redress of any adverse
environment impact from conduct of the
proposed activities can reasonably be
effective should such redress be
necessary;
(3) Whether conduct of the proposed
activities would foreclose subsequent
adoption of alternatives; and
(4) The effect of delay in conducting
such activities on the public interest,
including whether the power needs to
be used by the proposed facility, the
availability of alternative sources, if any,
to meet those needs on a timely basis
and delay costs to the applicant and to
consumers.
(d) Issuance of such an exemption
must not be deemed to constitute a
commitment to issue a CP. During the
period of any exemption granted
pursuant to paragraph (c) of this section,
any activities conducted must be carried
out in such a manner as will minimize
or reduce their environmental impact.
(e) The Commission’s consideration of
requests for exemptions from
requirements of the regulations of other
parts in this chapter that are applicable
by virtue of this part must be governed
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87049
by the exemption requirements of those
parts.
§ 53.090
Standards for review.
(a) Common standards. In
determining that a CP, OL, early site
permit, COL, or ML under this part will
be issued to an applicant, the
Commission will be guided by the
following considerations:
(1) Except for an early site permit or
ML, the processes to be performed, the
operating procedures, the facility and
equipment, the use of the facility, and
other technical specifications, or the
proposals, in regard to any of the
foregoing, collectively provide
reasonable assurance that the applicant
will comply with the regulations in this
chapter, including the regulations in 10
CFR part 20, and that the health and
safety of the public will not be
endangered.
(2) The applicant for a CP, OL, COL,
or ML is technically and financially
qualified to engage in the proposed
activities in accordance with the
regulations in this chapter. However, no
consideration of financial qualification
is necessary for an electric utility
applicant for an OL for a utilization
facility of the type described in
paragraph (d) of this section or for an
applicant for an ML.
(3) The issuance of a CP, OL, early site
permit, COL, or ML to the applicant will
not, in the opinion of the Commission,
be inimical to the common defense and
security or to the health and safety of
the public.
(4) Any applicable requirements of
subpart A of 10 CFR part 51 have been
satisfied.
(b) Additional standards for licenses.
In determining whether a license will be
issued to an applicant, the Commission
will, in addition to applying the
standards set forth in paragraph (a) of
this section, consider whether the
proposed activities will serve a useful
purpose proportionate to the quantities
of SNM or source material to be utilized.
(c) Additional standards and
provisions affecting licenses for
commercial power. In addition to
applying the standards set forth in
paragraphs (a) and (b) of this section,
paragraphs (c)(1) through (c)(4) of this
section apply in the case of a license for
a facility for the generation of
commercial power.
(1) The NRC will—
(i) Give notice in writing of each
application to the regulatory agency or
State as may have jurisdiction over the
rates and services incident to the
proposed activity;
(ii) Publish notice of the application
in trade or news publications as it
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deems appropriate to give reasonable
notice to municipalities, private
utilities, public bodies, and cooperatives
which might have a potential interest in
the utilization or production facility;
and
(iii) Publish notice of the application
once each week for four consecutive
weeks in the Federal Register. No
license will be issued by the NRC prior
to the giving of these notices and until
four weeks after the last notice is
published in the Federal Register.
(2) If there are conflicting applications
for a limited opportunity for such
license, the Commission will give
preferred consideration in the following
order: first, to applications submitted by
public or cooperative bodies for
facilities to be located in high cost
power areas in the United States;
second, to applications submitted by
others for facilities to be located in such
areas; third, to applications submitted
by public or cooperative bodies for
facilities to be located in areas other
than high cost power areas; and, fourth,
to all other applicants.
(3) The licensee who transmits
electric energy in interstate commerce,
or sells it at wholesale in interstate
commerce, must be subject to the
regulatory provisions of the Federal
Power Act.
(4) Nothing shall preclude any
government agency, now or hereafter
authorized by law to engage in the
production, marketing, or distribution of
electric energy, if otherwise qualified,
from obtaining a CP, OL, or COL under
this part for a utilization facility for the
primary purpose of producing electric
energy for disposition for ultimate
public consumption.
(d) Licenses for commercial nuclear
plants. A license will be issued, to an
applicant who qualifies, for any one or
more of the following: to transfer or
receive in interstate commerce, or
manufacture, produce, transfer, acquire,
possess, or use a utilization facility for
industrial or commercial purposes.
§ 53.100
Jurisdictional limits.
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No permit, license, standard design
approval, or standard design
certification under this part shall be
deemed to have been issued for
activities that are not under or within
the jurisdiction of the United States.
§ 53.110
Attacks and destructive acts.
Licensees, applicants for licenses,
permits, certifications, and design
approvals, and applicants for an
amendment to any license, permit,
certification, or design approval under
this part are not required to provide for
design features or other measures for the
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specific purpose of protection against
the effects of—
(a) Attacks and destructive acts,
including sabotage, directed against the
facility by an enemy of the United
States, whether a foreign government or
other person; or
(b) Use or deployment of weapons
incident to U.S. defense activities.
§ 53.115 Rights related to special nuclear
material.
(a) No right to the SNM will be
conferred by a license issued under this
part except as may be defined by the
license.
(b) Neither a license issued under this
part, nor any right thereunder, nor any
right to utilize or produce SNM may be
transferred, assigned, or disposed of in
any manner, either voluntarily or
involuntarily, directly or indirectly,
through transfer of control of the license
to any person, unless the Commission,
after securing full information, finds
that the transfer is in accordance with
the provisions of the Act and gives its
consent in writing.
§ 53.117 License suspension and rights of
recapture.
Any license issued under this part
must be subject to suspension and to the
rights of recapture of the material or
control of the facility reserved to the
Commission under section 108 of the
Act in a state of war or national
emergency declared by Congress.
§ 53.120 Information collection
requirements: OMB approval.
(a) The NRC has submitted the
information collection requirements
contained in this part to the Office of
Management and Budget (OMB) for
approval as required by the Paperwork
Reduction Act (44 U.S.C. 3501 et seq.).
The NRC may not conduct or sponsor,
and a person is not required to respond
to, a collection of information unless it
displays a currently valid OMB control
number. OMB has approved the
information collection requirements
contained in this part under control
number 3150–XXXX.
(b) The approved information
collection requirements contained in
this part appear in §§ 53.070, 53.080,
53.240, 53.410, 53.420, 53.425, 53.430,
53.440, 53.450, 53.480, 53.500, 53.540,
53.605, 53.610, 53.620, 53.700, 53.710,
53.715, 53.720, 53.730, 53.780, 53.785,
53.805, 53.810, 53.815, 53.830, 53.850,
53.855, 53.865, 53.870, 53.875, 53.880,
53.910, 53.1010, 53.1020, 53.1030,
53.1045, 53.1060, 53.1070, 53.1075,
53.1080, 53.1100, 53.1109, 53.1115,
53.1130, 53.1140, 53.1144, 53.1146,
53.1173, 53. 1182, 53.1188, 53.1200,
53.1206, 53.1209, 53.1210, 53.1221,
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53.1230, 53.1236, 53.1239, 53.1241,
53.1254, 53.1257, 53,1263, 53.1270,
53.1276, 53.1279, 53.1282, 53.1288,
53.1295, 53.1300, 53.1306, 53.1309,
53.1312, 53.1327, 53.1330, 53.1333,
53.1336, 53.1348, 53.1360, 53.1366,
53.1369, 53.1372, 53.1384, 53.1410,
53.1413, 53.1416, 53.1419, 53.1437,
53.1449, 53.1452, 53.1458, 53.1470,
53.1505, 53.1510, 53.1515, 53.1525,
53.1530, 53.1535, 53.1540, 53.1545,
53.1550, 53.1560, 53.1565, 53.1570,
53.1575, 53.1580, 53.1620, 53.1630,
53.1645, 53.1680, 53.1690, 53.1720.
(c) This part contains information
collection requirements in addition to
those approved under the control
number specified in paragraph (a) of
this section. The information collection
requirement and the control numbers
under which it is approved are as
follows:
(1) In §§ 53.765, 53.770, 53.780, and
53.795, NRC Form 396 is approved
under control number 3150–0024.
(2) In §§ 53.775 and 53.795, NRC
Form 398 is approved under control
number 3150–0090.
(3) In § 53.1640, NRC Form 366 is
approved under control number 3150–
0104.
(4) In § 53.1630, NRC Form 361 is
approved under control number 3150–
0238.
(5) In § 53.1650, International Atomic
Energy Agency Design Information
Questionnaire forms are approved under
control number 3150–0056.
(6) In § 53.1650, DOC/NRC Form AP–
A and associated forms are approved
under control numbers 0694–0135.
Subpart B—Technology-Inclusive
Safety Requirements
§ 53.210 Safety criteria for design-basis
accidents.
Design features and programmatic
controls must be provided for each
commercial nuclear plant such that
identification and analyses of designbasis accidents (DBAs) in accordance
with § 53.240 demonstrate the
following:
(a) An individual located at any point
on the boundary of the exclusion area
for any 2-hour period following the
onset of the postulated fission product
release would not receive a radiation
dose in excess of 25 rem (250
millisieverts) total effective dose
equivalent (TEDE); and
(b) An individual located at any point
on the outer boundary of the lowpopulation zone who is exposed to the
radioactive cloud resulting from the
postulated fission product release
(during the entire period of its passage)
would not receive a radiation dose in
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excess of 25 rem (250 millisieverts)
TEDE.1
1 The use of 25 rem TEDE is not intended
to imply that this number constitutes an
acceptable limit for an emergency dose to the
public under accident conditions. Rather,
this dose value has been set forth in this
section as a reference value, which can be
used in the evaluation of plant design
features with respect to postulated reactor
accidents, to assure that these designs
provide assurance of low risk of public
exposure to radiation, in the event of an
accident.
§ 53.220 Safety criteria for licensing-basis
events other than design-basis accidents.
Design features and programmatic
controls must be provided for each
commercial nuclear plant such that
identification and analysis of licensingbasis events (LBEs) other than DBAs in
accordance with § 53.240 demonstrate
the following:
(a) Plant SSCs, personnel, and
programs provide the necessary
capabilities and maintain the necessary
reliability to address LBEs other than
DBAs in accordance with §§ 53.240 and
53.450(e), and provide measures for
defense in depth in accordance with
§ 53.250; and
(b) The analysis of risks to public
health and safety resulting from LBEs
other than DBAs under § 53.450(e)
includes comprehensive risk metrics
that satisfy associated risk performance
objectives that are acceptable to the NRC
and provide an appropriate level of
safety.
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§ 53.230
Safety functions.
(a) The primary safety function is
limiting the release of radioactive
materials from the facility and must be
maintained during normal operation
and for LBEs over the life of the plant.
(b) Additional safety functions needed
to support the retention of radioactive
materials during LBEs—such as
controlling reactivity, heat generation,
heat removal, and chemical
interactions—must be identified for
each commercial nuclear plant.
(c) The primary and additional safety
functions are required to satisfy the
safety criteria defined in §§ 53.210 and
53.220, or more restrictive alternative
criteria adopted under § 53.470, and
must be fulfilled by the design features,
human actions, and programmatic
controls specified throughout this part.
§ 53.240
Licensing-basis events.
(a) Licensing-basis events must be
identified for each commercial nuclear
plant and analyzed under § 53.450 to
demonstrate that the safety
requirements in this subpart have been
satisfied.
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(b) The identified LBEs, ranging from
anticipated event sequences to very
unlikely event sequences, must
collectively address combinations of
malfunctions of plant SSCs, human
errors, facility hazards, and the effects of
external hazards.
(c) The analysis of LBEs must—
(1) Include analysis of one or more
DBAs under § 53.450(f);
(2) Confirm the adequacy of design
features and programmatic controls
needed to satisfy the safety criteria
defined in §§ 53.210 and 53.220, or
more restrictive alternative criteria
adopted under § 53.470, and
(3) Establish related functional
requirements for plant SSCs, personnel,
and programs.
§ 53.250
Defense in depth.
(a) Measures must be taken for each
commercial nuclear plant to ensure
appropriate defense in depth is
provided to compensate for
uncertainties in the analysis of the
safety criteria such that there is
reasonable assurance that the safety
criteria in this subpart are met over the
life of the plant.
(b) The uncertainties that must be
addressed under paragraph (a) of this
section include those related to the state
of knowledge and modeling capabilities,
the ability of barriers to limit the release
of radioactive materials from the facility
during LBEs other than DBAs, the
reliability and performance of plant
SSCs and personnel, and the
effectiveness of programmatic controls.
(c) The safety analysis may not rely
upon a single engineered design feature,
human action, or programmatic control,
no matter how robust, to address the
range of LBEs other than DBAs.
§ 53.260
Normal operations.
Holders of licenses to operate
commercial nuclear plants under this
part must control public doses and dose
rates in unrestricted areas from normal
plant operations to meet the
requirements in 10 CFR part 20.
§ 53.270
Protection of plant workers.
Holders of licenses to operate
commercial nuclear plants under this
part must control occupational doses to
meet the requirements in 10 CFR part
20.
Subpart C—Design and Analysis
Requirements
§ 53.400 Design features for licensingbasis events.
(a) Design features must be provided
for each commercial nuclear plant such
that, when combined with
corresponding human actions and
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87051
programmatic controls, the plant will
satisfy the safety criteria defined in
§§ 53.210 and 53.220, or more restrictive
alternative criteria adopted under
§ 53.470.
(b) Design features must ensure that
the safety functions identified in
§ 53.230 are fulfilled during licensingbasis events (LBEs).
§ 53.410 Functional design criteria for
design-basis accidents.
(a) Functional design criteria must be
defined for each design feature required
by § 53.400 and relied upon to
demonstrate compliance with the safety
criteria defined in § 53.210.
(b) Corresponding human actions and
programmatic controls must be
identified and implemented in
accordance with this and other subparts
to achieve and maintain the reliability
and capability of structures, systems,
and components (SSCs) relied upon to
satisfy the defined functional design
criteria and the safety criteria required
in § 53.210, and to maintain consistency
with analyses required by § 53.450(f).
§ 53.415 Protection against external
hazards.
Safety-related (SR) SSCs must be
protected against or must be designed to
withstand the effects of natural
phenomena (e.g., earthquakes,
tornadoes, hurricanes, floods, tsunami,
and seiches) and constructed hazards
(e.g., dams, transportation routes,
military and industrial facilities)
considering an event severity up to the
design-basis external hazard levels as
determined under § 53.510 without
losing the capability to perform the
safety functions identified under
§ 53.230. Specific requirements for
earthquake engineering are included in
§ 53.480.
§ 53.420 Functional design criteria for
licensing-basis events other than designbasis accidents.
(a) Functional design criteria must be
defined for each design feature required
by § 53.400 and relied upon to—
(1) Demonstrate compliance with the
safety criteria in § 53.220 or more
restrictive alternative criteria adopted
under § 53.470; and
(2) Demonstrate compliance with the
evaluation criteria in § 53.450(e) or more
restrictive alternative criteria adopted
under § 53.470.
(b) Corresponding human actions and
programmatic controls must be
identified and implemented in
accordance with this and other subparts
to achieve and maintain the reliability
and capability of SSCs relied upon to—
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(1) Satisfy the safety criteria in
§ 53.220 or more restrictive alternative
criteria adopted under § 53.470; and
(2) Satisfy the evaluation criteria in
§ 53.450(e) or more restrictive alternate
criteria adopted under § 53.470.
§ 53.425 Design features and functional
design criteria for normal operations.
(a) Design features must be provided
for each commercial nuclear plant to
support the Radiation Protection
Program required in § 53.850.
(b) Functional design criteria must be
defined for each design feature relied
upon to demonstrate compliance with
§ 53.850.
(c) Functional design criteria,
including design objectives for dose to
the maximally exposed member of the
public, must be defined for design
features to show that plant design
features and corresponding
programmatic controls, including
monitoring programs, control liquid,
gaseous, and solid wastes, as required
under part 20 of this chapter.1
1 A guide for keeping doses to the public
as low as is reasonably achievable is that the
estimated annual dose to the maximally
exposed member of the public does not
exceed 10 mrem total effective dose
equivalent. A design objective of maintaining
doses below 10 mrem/year should not be
construed as a radiation protection standard.
§ 53.430 Design features and functional
design criteria for protection of plant
workers.
(a) Design features must be provided
for each commercial nuclear plant such
that, when combined with
corresponding programmatic controls,
the requirements in § 53.270 can be met.
(b) Functional design criteria must be
defined for each design feature relied
upon to demonstrate compliance with
§ 53.270.
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§ 53.440
Design requirements.
(a)(1) Analysis, appropriate test
programs, prototype testing, operating
experience, or a combination thereof
must demonstrate that each design
feature required by § 53.400 meets the
defined functional design criteria
required by §§ 53.410 and 53.420. This
demonstration must consider
interdependent effects throughout the
commercial nuclear plant and the range
of conditions under which the design
features required by § 53.400 must
function throughout the plant’s lifetime.
(2) The design processes for SR and
non-safety-related but safety-significant
(NSRSS) SSCs under this part must
include administrative procedures for
evaluating operating, design, and
construction experience and for
considering applicable important
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industry experiences in the design of
those SSCs.
(b) The design features required by
§ 53.400 must, wherever applicable, be
designed using generally accepted
consensus codes and standards that
have been endorsed or otherwise found
acceptable by the U.S. Nuclear
Regulatory Commission (NRC).
(c) The materials used for each SR and
NSRSS SSC must be qualified for their
service conditions over the design life of
the SSC.
(d) Possible degradation mechanisms
related to aging, fatigue, chemical
interactions, operating temperatures,
effects of irradiation, and other
environmental factors that may affect
the performance of SR and NSRSS SSCs
must be evaluated and used to inform
the design and the development of
integrity assessment programs under
§ 53.870.
(e)(1) Safety-related and NSRSS SSCs
must be designed and located to
minimize, consistent with other safety
requirements in this part, the
probability and effect of fires and
explosions.
(2) Noncombustible and fire-resistant
materials must be used wherever
practical throughout the facility,
particularly in locations with SR and
NSRSS SSCs.
(3) Fire detection and fire suppression
systems of appropriate capacity and
capability must be provided and
designed to minimize the adverse effects
of fires on SR and NSRSS SSCs.
(4) Fire suppression systems must be
designed to ensure that their rupture or
inadvertent operation does not
significantly impair the ability of SR
and NSRSS SSCs to perform their safety
functions to satisfy § 53.230.
(f) Safety and security must be
considered together in the design
process such that, where possible,
security issues are effectively resolved
through design and engineered security
features.
(g) The reactor system and waste
stores for each commercial nuclear plant
must be capable of achieving and
maintaining a subcritical condition
during normal operations and following
any LBE identified in accordance with
§ 53.240.
(h) Each commercial nuclear plant
must have a capability to provide longterm cooling of the reactor fuel and
waste stores during normal operations
and following any LBE identified in
accordance with § 53.240.
(i) The design, analysis, staffing, and
programmatic controls for each
commercial nuclear plant must consider
the number of reactors, waste stores,
and other significant inventories of
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radioactive materials and the associated
operating configurations, common
systems, system interfaces, and system
interactions.
(j)(1) Design features must be
provided and related functional design
criteria defined such that, with limited
use of operator actions, one or more
physical barriers are maintained to limit
the release of radionuclides from reactor
systems, waste stores, or other
significant inventories of radioactive
materials assuming the impact of a
large, commercial aircraft.
(2) The functional design criteria for
those design features provided to
address the requirements in paragraph
(j)(1) of this section must be based on an
assessment of the impact of a large,
commercial aircraft used for long
distance flights in the United States,
with aviation fuel loading typically used
in such flights, and an impact speed and
angle of impact considering the ability
of both experienced and inexperienced
pilots to control large, commercial
aircraft at low altitude representative of
a commercial nuclear plant’s low
profile.1
1 Changes to the detailed parameters on
aircraft impact characteristics set forth in
guidance must be approved by the
Commission.
(k) Design features and related
functional design criteria must be
defined such that analyses demonstrate
a low risk of permanent injury to the
public due to the health effects of the
chemical hazards of licensed material.
(l) Measures must be taken during the
design of commercial nuclear plants to
minimize, to the extent practicable,
contamination of the facility and the
environment, facilitate eventual
decommissioning, and minimize, to the
extent practicable, the generation of
radioactive waste in accordance with
§ 20.1406 of this chapter.
(m)(1) Each commercial nuclear plant
must include criticality monitoring
capabilities meeting the requirements of
either § 70.24 of this chapter or
paragraph (m)(2) of this section.
(2) In lieu of maintaining a monitoring
system capable of detecting criticality as
described in § 70.24 of this chapter,
criticality accident requirements may be
satisfied by—
(i) Demonstrating the sub-criticality of
special nuclear material, except when it
is inside the reactor and the reactor is
being operated, by maintaining keffective below 0.95 at a 95 percent
probability, 95 percent confidence level,
under conditions that maximize
reactivity for the applicable storage and
handling configurations, and
(ii) Providing radiation monitors for
fuel storage and associated handling
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areas when fuel is present to detect
excessive radiation levels and to
support initiating appropriate safety
actions.
(3) While a spent fuel transportation
package approved under 10 CFR part 71
of this chapter or spent fuel storage cask
approved under 10 CFR part 72 is in the
special nuclear material handing or
storage area, the requirements in 10 CFR
parts 71 or 72, as applicable, and the
requirements of the certificate of
compliance for that package or cask, are
the applicable requirements for the fuel
within that package or cask.
(n)(1) The design of each commercial
nuclear plant must reflect state-of-theart human factors principles for safe and
reliable performance in all locations that
human activities are expected for
performing or supporting the continued
availability of plant safety or emergency
response functions.
(2) The design must provide for the
capabilities described in § 53.730(b) to
ensure the plant staff are able to monitor
plant conditions and respond to events.
(3) The means by which the design
and human actions together will achieve
the safety requirements of subpart B of
this part must be evaluated and used to
inform the design and the development
of the concept of operations required by
§ 53.730(c).
(4) A functional requirements analysis
and function allocation must be used to
ensure that plant design features
address how safety functions and
functional safety criteria are satisfied,
and how the safety functions will be
assigned to appropriate combinations of
human action, automation, active safety
features, passive safety features, or
inherent safety characteristics.
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§ 53.450
Analysis requirements.
(a) Requirement to have a
probabilistic risk assessment (PRA). A
PRA of each commercial nuclear plant
must be performed to identify potential
failures, susceptibility to internal and
external hazards, and other contributing
factors to event sequences that might
challenge the safety functions identified
in § 53.230 and to support
demonstrating that each commercial
nuclear plant meets the safety criteria of
§ 53.220, or more restrictive alternative
criteria adopted under § 53.470.
(b) Specific uses of analyses. The PRA
in combination with other generally
accepted approaches for systematically
evaluating engineered systems must be
used—
(1) In informing the selection of the
LBEs, as described in § 53.240, which
must be considered in the design to
determine compliance with the safety
criteria in subpart B of this part.
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(2) For informing the classification of
SSCs according to their safety
significance in accordance with § 53.460
and for identifying the environmental
conditions under which the SSCs and
operating staff must perform their safety
functions.
(3) In evaluating the adequacy of
defense-in-depth measures required in
accordance with § 53.250.
(4) To identify and assess all plant
operating states where there is the
potential for the uncontrolled release of
radioactive material to the environment.
(5) To identify and assess events that
challenge plant control and safety
systems whose failure could lead to the
uncontrolled release of radioactive
material to the environment. These
include internal events, such as human
errors and equipment failures, and
external events identified in accordance
with subpart D of this part.
(c) Maintenance and upgrade of
analyses. The PRA must be maintained
at least every 5 years until the
permanent cessation of operations
under § 53.1070 and upgraded in
conformance with generally accepted
methods, standards, and practices that
have been endorsed or otherwise found
acceptable by the NRC.
(d) Qualification of analytical codes.
The analytical codes used in modeling
plant behavior in analyses of licensingbasis events (including but not limited
to thermodynamics, reactor physics,
fuel performance, and mechanistic
source term codes) must be qualified for
the range of conditions for which they
are to be used.
(e) Analyses of licensing-basis events
other than design-basis accidents.
(1) Analyses must be performed for
LBEs other than design-basis accidents
(DBAs). These LBEs must be identified
using insights from a PRA in
combination with other generally
accepted approaches for systematically
evaluating engineered systems to
identify and analyze equipment failures
and human errors.
(2) The analysis of LBEs other than
DBAs must include definition of
evaluation criteria for each event or
specific categories of LBEs to determine
the acceptability of the plant response to
the challenges posed by internal and
external hazards to provide an
appropriate level of safety.
(3) The analyses of LBEs other than
DBAs must address event sequences
from initiation to a defined end state
and be used in combination with other
engineering analyses to demonstrate
that the functional design criteria
required by § 53.420 provide sufficient
barriers to the unplanned release of
radionuclides to satisfy the evaluation
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criteria defined for each LBE other than
DBAs, to satisfy the safety criteria
specified in accordance with § 53.220
and provide defense in depth as
required by § 53.250.
(4) The methodology used to identify,
categorize, and analyze LBEs must
include a means to identify event
sequences deemed significant for
controlling the risks posed to public
health and safety.
(f) Analysis of design-basis accidents.
(1) The analysis of LBEs required by
§ 53.240 must include analysis of DBAs
that address possible challenges to the
safety functions identified under
§ 53.230. The events selected as DBAs
must be those that, if not terminated,
have the potential for exceeding the
safety criteria in § 53.210.
(2) The DBAs selected must be
analyzed using deterministic methods
that address event sequences from
initiation to a safe stable end state and
assume only the SR SSCs identified
under § 53.460 and human actions
addressed by the requirements of
subpart F of this part are available to
perform the safety functions identified
in accordance with § 53.230.
(3) The analysis must conservatively
demonstrate compliance with the safety
criteria in § 53.210.
(g) Other required analyses. Analyses
must be performed to assess—
(1) Fire protection. Fire protection
measures to demonstrate, through
inclusion of fires in the analysis of LBEs
or by separate analyses, that a fire or
explosion in any plant area would not—
(i) Prevent equipment from fulfilling
the safety functions identified in
accordance with § 53.230, or
(ii) Challenge the safety criteria in
§§ 53.210 and 53.220.
(2) Aircraft impact. Measures
provided to protect against aircraft
impacts under § 53.440(j).
(3) Dose to members of the public.
Measures taken under § 53.425,
including estimating—
(i) The quantity of each of the
principal radionuclides expected to be
released annually to unrestricted areas
in liquid effluents produced during
normal reactor operations and the dose
to the maximally exposed member of
the public in unrestricted areas.
(ii) The quantities of each of the
principal radionuclides of the gases,
halides, and particulates expected to be
released annually to unrestricted areas
in gaseous effluents produced during
normal reactor operations and the dose
to the maximally exposed member of
the public in unrestricted areas.
(iii) The annual external radiation
dose in unrestricted areas and the
maximally exposed member of the
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public in unrestricted areas due to
direct radiation from contained
radiation sources from the commercial
nuclear plant during normal reactor
operations.
§ 53.460 Safety categorization and special
treatments.
(a) Structures, systems, and
components must be classified
according to their safety significance.
The SSC categories must include
‘‘Safety-Related,’’ ‘‘Non-Safety-Related
but Safety-Significant,’’ and ‘‘NonSafety-Significant,’’ as defined in
subpart A of this part.
(b) For SR and NSRSS SSCs, the
conditions under which they must
perform their safety function in § 53.230
must be identified. Special treatments
must be established in accordance with
this and other subparts to provide
confidence that the SSCs will perform
under the service conditions and with
reliability consistent with the analysis
performed under § 53.450 to
demonstrate meeting the safety criteria
in §§ 53.210 and 53.220, or more
restrictive alternative criteria adopted
under § 53.470.
(1) The special treatments for SR SSCs
must include meeting the applicable
quality assurance requirements from
appendix B of part 50 of this chapter.
(2) The special treatments for NSRSS
SSCs and special treatments for SR SSCs
beyond those required under (b)(1) of
this section may include meeting
selected quality assurance requirements
from appendix B of part 50 of this
chapter when such treatment is needed
to address performance requirements,
equipment reliability, or uncertainties.
(c) Human actions needed to prevent
or mitigate LBEs must be identified, be
able to be performed reliably under the
postulated environmental conditions,
and be addressed by programs
established in accordance with subpart
F of this part to provide confidence that
those actions will be performed as
assumed in the analysis performed in
accordance with § 53.450 to
demonstrate meeting the criteria in
§§ 53.210, 53.220, and 53.450(e), or
more restrictive alternative criteria
adopted under § 53.470.
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§ 53.470 Maintaining analytical safety
margins used to justify operational
flexibilities.
Where an applicant or licensee so
chooses, alternative criteria more
restrictive than those defined in
§§ 53.220 and 53.450(e) may be adopted
to support operational flexibilities. In
such cases, applicants and licensees
must ensure that the functional design
criteria of § 53.420, the analysis
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requirements of § 53.450(e), and
identification of special treatment of
SSCs and human actions under § 53.460
reflect and support the use of alternative
criteria to justify operational
flexibilities. Licensees must ensure that
measures taken to provide the analytical
margins supporting operational
flexibilities are incorporated into design
features and programmatic controls and
are maintained within programs
required in other subparts.
§ 53.480
Earthquake engineering.
(a) Effects of earthquakes. Structures,
systems, and components classified as
SR or NSRSS must be able to withstand
the effects of earthquakes,
commensurate with the safety
significance of the SSC, without loss of
capability to perform their role in
fulfilling the safety functions required
by § 53.230.
(b) Definitions. For the purpose of this
section—
Design-Basis Ground Motions
(DBGMs) are the vibratory ground
motions for which certain SSCs must be
designed to remain functional.
Operating basis earthquake (OBE)
ground motion is the vibratory ground
motion for which those features of the
commercial nuclear plant necessary for
continued operation without undue risk
to the health and safety of the public are
designed to remain functional. The OBE
ground motion is used in § 53.720.
Response spectrum is a plot of the
maximum responses (acceleration,
velocity, or displacement) of idealized
single-degree-of-freedom oscillators as a
function of the natural frequencies of
the oscillators for a given damping
value. The response spectrum is
calculated for a specified vibratory
motion input at the oscillators’
supports.
Surface deformation is the distortion
of geologic strata on or near the ground
surface that occurs because of tectonic
forces that result from earthquakes.
(c) Design considerations—(1) DesignBasis Ground Motions. (i) The DBGMs
must be derived from the Site Ground
Motion Response Spectra developed in
accordance with § 53.510(c), by taking
into consideration the functional design
criteria of SSCs in accordance with
§§ 53.410 and 53.420. The horizontal
component of the DBGM(s) in the freefield at the foundation level of the
structures must be an appropriate
response spectrum that is determined
based on the risk-significance of SSCs
and their safety functions. In view of the
limited data available on vibratory
ground motion of strong earthquakes, it
is acceptable that the design response
spectra be smoothed spectra.
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(ii) The commercial nuclear plant
must be designed so that, if the DBGMs
occur, the following SSCs remain
functional and within applicable stress,
strain, and deformation limits:
(A) Structures, systems, and
components for which functional design
criteria are established in accordance
with § 53.410 or § 53.420; and
(B) Structures, systems, and
components classified as SR or NSRSS
commensurate with safety significance
in accordance with § 53.460.
(iii) In addition to seismic loads,
applicable concurrent normal operating,
functional, and accident-induced loads
must be taken into account in the design
of the SR SSCs and, commensurate with
safety significance, NSRSS SSCs.
(iv) The design of the commercial
nuclear plant must take into account the
possible effects of seismic-induced
ground disruption, such as fissuring,
lateral spreads, differential settlement,
liquefaction, and landsliding, on the
facility foundations.
(v) The SSCs fulfilling the safety
functions required by § 53.230 must be
demonstrated through design, testing, or
qualification methods to be able to
fulfill those safety functions during and
after the vibratory ground motion
associated with the DBGMs.
(vi) The evaluation of SSCs required
by this section to show they are able to
function during and after earthquake
ground motion must take into account
soil-structure interaction effects and the
expected duration of vibratory motion.
It is permissible to design for strain
limits in excess of yield strain in some
of these SSCs during the DBGMs and
under the postulated concurrent loads,
provided the necessary safety functions
are maintained.
(2) OBE Ground Motion. The OBE
Ground Motion must be characterized
by response spectra. The value of the
OBE Ground Motion must be set to onethird or less of the DBGMs response
spectra.
(3) [Reserved]
(4) Required seismic instrumentation.
Suitable instrumentation must be
provided so that the seismic response of
commercial nuclear plant SR SSCs or
NSRSS SSCs can be evaluated promptly
after an earthquake.
(d) Surface deformation. (1) The
potential for surface deformation must
be taken into account in the design of
the commercial nuclear plant by
providing reasonable assurance that in
the event of deformation, SSCs
classified as SR or NSRSS in accordance
with § 53.460 will remain functional.
(2) In addition to surface deformation
induced loads, the design of SSCs must
take into account, commensurate with
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safety significance, seismic loads and
applicable concurrent functional and
accident-induced loads.
(3) The design provisions for surface
deformation must be based on its
postulated occurrence in any direction
and azimuth and under any part of the
commercial nuclear plant, unless
evidence indicates this assumption is
not appropriate, and must take into
account the estimated rate at which the
surface deformation may occur.
(e) Seismically induced floods and
water waves and other design
conditions. Seismically induced floods
and water waves from either locally or
distantly generated seismic activity and
other design conditions determined
pursuant to subpart D of this part must
be taken into account in the design of
the commercial nuclear plant so as to
prevent undue risk to the health and
safety of the public.
(f) Analysis. The analyses required by
§ 53.450 must address seismic hazards
and related SSC responses in
determining that the safety criteria
defined in § 53.220 will be met.
(g) Design criteria, human actions,
and programmatic controls. Functional
design criteria, human actions, and
programmatic controls needed to
address seismic events must be
identified and implemented in
accordance with this and other subparts
to achieve and maintain the
performance of SSCs relied upon to
satisfy the safety criteria in § 53.220 and
to maintain consistency with analyses
required by § 53.450 when accounting
for the site-specific frequencies and
magnitudes of earthquakes for a
commercial nuclear plant.
Subpart D—Siting Requirements
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§ 53.500 General siting and siting
assessment.
(a) The siting of each commercial
nuclear plant must be supported by
assessments of proposed sites such that
the design, including design features
and programmatic controls
corresponding to the site characteristics,
satisfies the safety criteria defined in
§§ 53.210 and 53.220 or more restrictive
alternative criteria adopted under
§ 53.470. The siting assessment must
ensure that site characteristics that
might contribute to the initiation,
progression, or consequences of
licensing-basis events (LBEs) analyzed
under §§ 53.450 and 53.480 are
identified and mitigated by design
features or programmatic controls. The
siting assessment must take into
consideration the potential adverse
impacts that a commercial nuclear plant
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may have on nearby populations as a
result of normal operations or LBEs.
(b) Activities performed to identify
site characteristics or otherwise needed
to determine site-specific contributors to
functional design criteria or analysis
assumptions under subpart C of this
part must satisfy the applicable special
treatment requirements of § 53.460,
including, where applicable, the quality
assurance requirements from appendix
B of part 50 of this chapter.
§ 53.510
External hazards.
(a) General external hazard
requirements. The design-basis external
hazard level for the relevant external
hazards for a site must be identified and
characterized based on site-specific
assessments of natural and constructed
hazards with the potential to adversely
affect plant functions. The external
hazard frequencies and magnitudes
determined from the site-specific
assessments must take into account
uncertainties and variabilities in data,
models, and methods relied on to
characterize the external hazards.
(b) Definitions. For the purpose of this
section, the following terms mean:
Geological siting factors are geological
and seismic factors that may affect the
design and operation of the proposed
commercial nuclear plant.
Ground Motion Response Spectra
(GMRS) are the site-specific GMRS
resulting from the geologic
investigations and evaluations of the
site vicinity and region and used to
determine design-basis ground motions
for structures, systems, and components
under § 53.480.
Probabilistic seismic hazard analysis
is an analytical methodology that
incorporates uncertainty into estimates
of an annual frequency of exceedance
for a certain ground motion parameter
(e.g., peak ground acceleration, peak
ground velocity, response spectral
values) at a site.
(c) Geological investigations. The
GMRS for the site must be determined
based on the results of investigations of
the geological, seismological, and
engineering characteristics of the site
and its environs and must be
characterized by both horizontal and
vertical free-field GMRS at the free
ground surface. The size of the region to
be investigated and the type of data
pertinent to the investigations must be
determined based on the nature of the
region surrounding the site. Data on
vibratory ground motion, earthquake
recurrence rates, fault geometry and slip
rates, and site subsurface material
properties must be obtained by
reviewing pertinent literature and
carrying out field investigations.
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Uncertainties are inherent in the
parameters and models used to estimate
the GMRS for the site. The site
assessment must reflect these
uncertainties through an appropriate
analysis, such as a probabilistic seismic
hazard analysis.
(d) Geologic and seismic siting
factors. The geologic and seismic siting
factors considered for design under
§§ 53.415 and 53.480 must include, but
are not limited to, determination of the
potential for surface tectonic and
nontectonic deformations, the size and
character of seismically induced floods
and water waves that could affect a site
from either locally or distantly
generated seismic activity, soil and rock
stability, liquefaction potential, and
natural and artificial slope stability.
§ 53.520
Site characteristics.
Site characteristics that might
contribute to the initiation, progression,
or consequences of LBEs analyzed
under § 53.450 must be identified,
assessed, and considered in the design
and analyses required by subpart C of
this part.
§ 53.530 Population-related
considerations.
Every site must have an exclusion
area, a low-population zone, and a
population center distance as defined in
§ 53.020.
(a) The offsite radiological
consequences estimated by the analyses
required by § 53.450(f) must be used to
confirm that—
(1) An individual located at any point
on the boundary of the exclusion area
for any 2-hour period following onset of
the postulated fission product release
would not receive a radiation dose in
excess of 25 rem (250 millisieverts) total
effective dose equivalent.
(2) An individual located at any point
on the outer boundary of the lowpopulation zone who is exposed to the
radioactive cloud resulting from the
postulated fission product release
(during the entire period of its passage)
would not receive a radiation dose in
excess of 25 rem (250 millisieverts) total
effective dose equivalent.
(b) The population center distance
must be at least one and one-third times
the distance from the reactor to the
outer boundary of the low-population
zone. The boundary of the population
center must be determined upon
consideration of population
distribution. Political boundaries are not
controlling in the calculation of
population center distance.
(c) Reactor sites should be located
away from very densely populated
centers. Areas of low-population density
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are, generally, preferred. However, in
determining the acceptability of a
particular site located away from a very
densely populated center but not in an
area of low-population density,
consideration will be given to safety,
environmental, economic, or other
factors, which may result in the site
being found acceptable.
§ 53.540
Siting interfaces.
Site characteristics must be addressed
by the design features, programmatic
controls, and supporting analyses used
to demonstrate that the safety criteria in
§§ 53.210 and 53.220 are met for each
commercial nuclear plant. Site
characteristics must be such that
adequate emergency plans and security
plans can be developed and maintained.
Subpart E—Construction and
Manufacturing Requirements
§ 53.600 Construction and
manufacturing—scope and purpose.
This subpart applies to those
construction and manufacturing
activities authorized by a construction
permit (CP), combined license (COL),
manufacturing license (ML), or limited
work authorization (LWA) issued under
this part.
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§ 53.605 Reporting of defects and
noncompliance.
Each CP and ML issued under this
part is subject to the terms and
conditions in this section, and each COL
issued under this part is subject to the
terms and conditions in this section
until the date that the Commission
makes the finding under § 53.1452(g).
(a) Definitions. The definitions in
§ 21.3 of this chapter apply to this
section.
(b) Posting requirements. (1) Each
individual, partnership, corporation,
dedicating entity, or other entity subject
to the regulations in this section must
post current copies of this section and
the regulations in 10 CFR part 21;
section 206 of the Energy
Reorganization Act of 1974, as
amended; and procedures adopted
under these regulations. These
documents must be posted in a
conspicuous position on any premises
within the United States where the
activities subject to the license are
conducted.
(2) If posting of these regulations or
the procedures adopted under them is
not practical, the licensee may, in
addition to posting section 206 of the
Energy Reorganization Act of 1974, as
amended, post a notice that describes
the regulations/procedures, including
the name of the individual to whom
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reports may be made, and states where
they may be examined.
(c) Procedures. The holder of a CP,
COL, or ML subject to this section must
adopt appropriate procedures to—
(1) Evaluate deviations and failures to
comply to identify defects and failures
to comply associated with substantial
safety hazards as soon as practicable,
and, except as provided in paragraph
(c)(2) of this section, in all cases within
60 days of discovery, to identify a
reportable defect or failure to comply
that could create a substantial safety
hazard, were it to remain uncorrected.
(2) Ensure that if an evaluation of an
identified deviation or failure to comply
potentially associated with a substantial
safety hazard cannot be completed
within 60 days from the discovery of the
deviation or failure to comply, an
interim report is prepared and
submitted to the Commission through a
director or responsible officer, or
designated person as discussed in
paragraph (d)(5) of this section. The
interim report should describe the
deviation or failure to comply that is
being evaluated and should also state
when the evaluation will be completed.
This interim report must be submitted
in writing within 60 days of discovery
of the deviation or failure to comply.
(3) Ensure that a director or
responsible officer of the holder of a CP,
COL, or ML subject to this section is
informed as soon as practicable, and, in
all cases, within the 5 working days
after completion of the evaluation
described in paragraph (c)(1) or (c)(2) of
this section, if the construction or
manufacture of a facility or activity, or
a basic component supplied for such a
facility or activity—
(i) Fails to comply with the Atomic
Energy Act of 1954, as amended, or any
applicable regulation, order, or license
of the Commission relating to a
substantial safety hazard;
(ii) Contains a defect; or
(iii) Underwent any significant
breakdown in any portion of the quality
assurance program (QAP) conducted
under the requirements of appendix B to
part 50 of this chapter that could have
produced a defect in a basic component.
These breakdowns in the QAP are
reportable whether or not the
breakdown actually resulted in a defect
in a design approved and released for
construction, installation, or
manufacture.
(d) Reporting defects and
noncompliance. (1) The holder of a CP,
COL, or ML subject to this section that
obtains information reasonably
indicating that the facility or
manufactured reactors fails to comply
with the Atomic Energy Act of 1954, as
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amended, or any applicable regulation,
order, or license of the Commission
relating to a substantial safety hazard
must notify the Commission of the
failure to comply through a director,
responsible officer, or designated person
as discussed in paragraph (d)(5) of this
section.
(2) The holder of a CP, COL, or ML
subject to this section that obtains
information reasonably indicating the
existence of any defect found in the
construction or manufacture, or any
defect found in the final design of a
facility as approved and released for
construction or manufacture, must
notify the Commission of the defect
through a director, responsible officer,
or designated person as discussed in
paragraph (d)(5) of this section.
(3) The holder of a CP, COL, or ML
subject to this part, who obtains
information reasonably indicating that
the QAP has undergone any significant
breakdown discussed in paragraph
(c)(3)(iii) of this section must notify the
Commission of the breakdown in the
QAP through a director, responsible
officer, or designated person as
discussed in paragraph (d)(5) of this
section.
(4) When acting as a dedicating entity,
the holder of a CP, COL, or ML subject
to this section is responsible for
identifying and evaluating deviations;
reporting defects and failures to comply
associated with substantial safety
hazards for dedicated items; and
maintaining auditable records for the
dedication process.
(5) The notification requirements of
this paragraph apply to all defects and
failures to comply associated with a
substantial safety hazard regardless of
whether extensive evaluation, redesign,
or repair is required to conform to the
criteria and bases stated in the Safety
Analysis Report, CP, COL, or ML.
Evaluation of potential defects and
failures to comply and reporting of
defects and failures to comply under
this section satisfies the CP holder’s,
COL holder’s, and ML holder’s
evaluation and notification obligations
under 10 CFR part 21, and satisfies the
responsibility of individual directors or
responsible officers or holders of a CP,
COL, or ML subject to this section to
report defects, and failures to comply
associated with substantial safety
hazards under section 206 of the Energy
Reorganization Act of 1974, as
amended. The director or responsible
officer may authorize an individual to
provide the notification required by this
section. However, this does not relieve
the director or responsible officer of his
or her responsibility under this section.
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(e) Notification—timing and where
sent. The notification required by
paragraph (d) of this section must
consist of—
(1) Initial notification by telephone,
facsimile, or email identified in
appendix A to 10 CFR part 73 to the
U.S. Nuclear Regulatory Commission
(NRC) Operations Center within 2 days
following receipt of information by the
director or responsible corporate officer
under paragraph (c)(3) of this section,
on the identification of a defect or a
failure to comply. If the CP, COL, or ML
holder elects to use facsimile,
verification that the facsimile has been
received should be made by calling the
NRC Operations Center. This paragraph
does not apply to interim reports
described in paragraph (c)(2) of this
section.
(2) Written notification submitted to
the NRC Document Control Desk by an
appropriate method listed in § 53.040,
with a copy to the appropriate NRC
Regional Administrator at the address
specified in appendix D to 10 CFR part
20 and a copy to the appropriate NRC
resident inspector, if applicable, within
30 days following receipt of information
by the director or responsible corporate
officer under paragraph (c)(3) of this
section, on the identification of a defect
or failure to comply.
(f) Content of notification. The written
notification required by paragraph (e)(2)
of this section must clearly indicate that
the written notification is being
submitted under this section and
include the following information, to
the extent known.
(1) Name and address of the
individual or individuals informing the
Commission.
(2) Identification of the facility, the
activity, or the basic component
supplied for the facility or the activity
within the United States which contains
a defect or fails to comply.
(3) Identification of the firm
constructing or manufacturing the
facility or supplying the basic
component which fails to comply or
contains a defect.
(4) Nature of the defect or failure to
comply and the safety hazard which is
created or could be created by the defect
or failure to comply.
(5) The date on which the information
of a defect or failure to comply was
obtained.
(6) In the case of a basic component
that contains a defect or failure to
comply, the number and location of
these components in use at the facility
subject to the regulations in this part.
(7) In the case of a completed reactor
manufactured under this part, the
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entities to which the reactor was
supplied.
(8) The corrective action which has
been, is being, or will be taken; the
name of the individual or organization
responsible for the action; and the
length of time that has been or will be
taken to complete the action.
(9) Any advice related to the defect or
failure to comply about the facility,
activity, or basic component that has
been, is being, or will be given to other
entities.
(g) Procurement documents. Each
holder of a CP, COL, or ML subject to
this section must ensure that each
procurement document for a facility or
a basic component specifies the
provisions of 10 CFR part 21 or this
section that apply, as applicable.
(h) Coordination with 10 CFR part 21.
The requirements of this section are
satisfied when the defect or failure to
comply associated with a substantial
safety hazard has been previously
reported under 10 CFR part 21, under
§ 73.1205 of this chapter, under this
section, or under § 53.1640.
(i) Records retention. The holder of a
CP, COL, or ML subject to this section
must prepare and maintain records
necessary to accomplish the purposes of
this section, specifically—
(1) Retain procurement documents,
which define the requirements that
facilities or basic components must
satisfy in order to be considered
acceptable, for the lifetime of the facility
or basic component.
(2) Retain records of evaluations of all
deviations and failures to comply under
paragraph (c)(1) of this section for the
longest of—
(i) Ten years from the date of the
evaluation;
(ii) Five years from the date that an
early site permit is referenced in an
application for a COL; or
(iii) Five years from the date of
delivery of a manufactured reactor.
(3) Retain records of all interim
reports to the Commission made under
paragraph (c)(2) of this section, or
notifications to the Commission made
under paragraph (d) of this section for
the minimum time periods stated in
paragraph (i)(2) of this section;
(4) Suppliers of basic components
must retain records of—
(i) All notifications sent to affected
licensees or purchasers under paragraph
(d)(4) of this section for a minimum of
10 years following the date of the
notification;
(ii) The facilities or other purchasers
to whom the basic components or
associated services were supplied for a
minimum of 15 years from the delivery
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of the basic component or associated
services.
(5) Maintaining reports in accordance
with this section satisfies the
recordkeeping obligations under 10 CFR
part 21 of the entities, including
directors or responsible officers thereof,
subject to this section.
§ 53.610
Construction.
(a) Management and control.
Licensees must ensure that the
following plans, programs, and
organizational units are developed and
implemented to manage and control the
construction activities:
(1) Programs to ensure that the
construction of a commercial nuclear
plant supports the eventual compliance
with the design and analysis
requirements in subpart C of this part.
(2) An organization, headed by
qualified personnel, responsible for
managing, controlling, and evaluating
the adequacy of the construction
activities.
(3) Procedures describing the
qualifications for personnel in key
positions in the licensee’s management
and control organization and the
organizational responsibilities,
authority, and interfaces with other
parts of the licensee’s organization.
(4) Procedures to evaluate the
applicability of other national and
international construction experience to
the planned and ongoing construction
activities and to ensure the applicable
experience will be provided to those
constructing the plant.
(5) A fitness-for-duty program, under
10 CFR part 26.
(6)(i) A QAP meeting the
requirements of appendix B of part 50
of this chapter as required by
§ 53.460(b).
(ii) Appropriate programmatic
controls to provide special treatment for
non-safety-related but safety-significant
structures, systems, and components
(SSCs).
(7) A radiation protection program, in
accordance with 10 CFR part 20, that
includes measures for monitoring the
dose to individuals working with
radioactive materials brought onto the
site, as applicable.
(8) An information security program
in accordance with §§ 73.21, 73.22, and
73.23 of this chapter, as applicable.
(b) Construction activities. No person
may begin the construction of a
commercial nuclear plant on a site on
which the facility is to be operated
under this part until that person has
been issued either a CP or COL, an early
site permit authorizing activities under
§ 53.1130, or an LWA under this part.
(1) Licensees must satisfy the
following requirements:
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(i) As appropriate, considering the
types and quantities of radioactive
materials being brought onto the site—
(A) The licensee must maintain and
follow a special nuclear material (SNM)
material control and accounting
program, a measurement control
program, and other material control
procedures that include corresponding
record management requirements as
required by the provisions of § 70.32 of
this chapter. Prior to initial receipt of
SNM onsite, the licensee must
implement an SNM material control and
accounting program in accordance with
10 CFR part 74.
(B) Procedures must be in place to
receive, possess, use, and store source,
byproduct, and SNM in accordance with
applicable portions of 10 CFR parts 30,
40, and 70.
(C) A plant staff training program
associated with the receipt of
radioactive material must be approved
and implemented prior to initial receipt
of byproduct, source or SNM (excluding
exempt quantities as described in
§ 30.18 of this chapter).
(ii) For construction of a commercial
nuclear plant involving multiple reactor
units, plans and procedures must be in
place to prevent or mitigate potential
hazards to the SSCs of operating units
resulting from construction activities,
including the managerial and
administrative controls to be used to
provide assurance that the limiting
conditions for operation of the operating
units are not exceeded as a result of
construction activities.
(iii) Procedures must be in place prior
to the start of construction activities that
describe how construction will be
controlled so as not to impact other
features important to the design, such as
dewatering, slope stability, backfill,
compaction, and seepage.
(iv) For LWA holders, a plan must be
developed for redress of activities
performed under the LWA should one
of the following situations arise:
(A) LWA work activities are
terminated by the holder of the LWA;
(B) The LWA is revoked by the NRC;
or
(C) The Commission denies the
associated CP or COL application.
(2)(i) Onsite fresh fuel must be
protected and stored in compliance with
§ 73.67 of this chapter.
(ii) Before initial fuel load into the
reactor (or, for a fueled manufactured
reactor, before initiating the physical
removal of any one of the independent
physical mechanisms to prevent
criticality required under
§ 53.620(d)(1)), a cybersecurity program
that meets the requirements of §§ 73.54
or 73.110 of this chapter, a physical
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security program that meets the
requirements of §§ 73.55 or 73.100 of
this chapter, and an access
authorization program that meets the
requirements of §§ 73.56 or 73.120 of
this chapter must be established, as
applicable.
(iii) Fire protection measures must be
implemented for work and storage areas
(including adjacent fire areas that could
affect the work or storage area) before
initial receipt of byproduct, source, or
non-fuel SNM (excluding exempt
quantities as described in § 30.18 of this
chapter). The fire protection measures
for areas associated with new fuel
(including all fuel handling, fuel
storage, and adjacent fire areas that
could affect the new fuel) must be
implemented before receipt of fuel.
Prior to the receipt of fuel, a formal
letter of agreement must be in place
with the local fire department
specifying the nature of arrangements in
support of the fire protection program.
(c) Inspection and acceptance. (1) The
licensee must have a process for
accepting individual or groups of SSCs
upon completion of construction and
protecting them from damage or
tampering as other construction
activities continue.
(2) The post construction acceptance
process must address the inspections,
tests, analyses, and acceptance criteria
specified in the COL under § 53.1440 or
the equivalent verifications needed to
support the issuance of an operating
license under § 53.1387.
§ 53.620
Manufacturing.
(a) Management and control. Holders
of MLs must ensure that the following
plans, programs, and organizational
units are developed and implemented to
manage and control the manufacturing
activities within the scope of the ML:
(1) Programs to ensure that the
manufacturing of a manufactured
reactor or portions of a manufactured
reactor complies with the design and
analysis requirements in subpart C of
this part. The entity with design
authority for the manufactured reactor
covered by the ML must be identified in
the license.
(2) An organizational and
management structure responsible for
managing, controlling, and evaluating
the adequacy of the reactor design and
manufacturing activities.
(3) Procedures describing the
qualifications for personnel in key
positions in the licensee’s management
and control organization and the
organizational responsibilities,
authority, and interfaces with other
parts of the licensee’s organization.
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(4) A program to evaluate the
applicability of other national and
international design and manufacturing
experience to the planned and ongoing
manufacturing activities.
(5) A fitness-for-duty program, in
accordance with 10 CFR part 26.
(6)(i) A QAP meeting the
requirements of appendix B to part 50
of this chapter, to be applied to the
design, fabrication, construction, and
testing of the SSCs of the manufactured
reactor.
(ii) Appropriate programmatic
controls to provide special treatment
measures for non-safety-related but
safety-significant SSCs.
(7) A radiation protection program, in
accordance with 10 CFR part 20, that
includes measures for monitoring the
dose to individuals if the manufacturing
activities include working with
radioactive materials.
(8) An information security program
in accordance with §§ 73.21, 73.22 and
73.23 of this chapter, as applicable.
(b) Manufacturing activities. Holders
of MLs must satisfy the following
requirements:
(1) The manufacturing process must
be conducted within facilities for which
the ML holder has the authority to
establish controls on any activity that
might affect manufacturing. The
licensee must establish access controls
to the portions of each facility involved
in the manufacturing processes
governed by the ML.
(2) Manufacturing processes must be
performed in accordance with the ML
and the referenced codes and standards
that have been endorsed or otherwise
found acceptable by the NRC.
(3) A post-manufacturing inspection
and acceptance process must be
established and implemented before
transporting a manufactured reactor or
portions of a manufactured reactor for
installation at a commercial nuclear
plant. The process must consider the
results of inspections, tests, and
analyses that have been performed and
the acceptance criteria that are
necessary and sufficient to conclude
that manufacturing activities have been
completed in accordance with the ML.
(c) Control of radioactive materials.
As appropriate considering the types
and quantities of radioactive materials
being brought into the manufacturing
facility—
(1) Procedures must be in place to
receive, transfer, possess, and use
source, byproduct, and SNM in
accordance with the applicable portions
of 10 CFR parts 30, 40 and 70.
(2) A fire protection program must be
established and implemented before the
initial receipt of byproduct, source, or
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non-fuel SNM (excluding exempt
quantities as described in § 30.18 of this
chapter).
(3) An emergency plan appropriate for
responding to the facility-specific
hazards of an accidental release of
radioactive material and to limit the
health effects of the associated chemical
hazards of licensed material must be
approved and implemented prior to the
receipt of byproduct, source, or SNM
(excluding exempt quantities as
described in § 30.18 of this chapter).
(4) A plant staff training program
associated with the receipt of
radioactive material must be approved
and implemented before initial receipt
of byproduct, source, or SNM
(excluding exempt quantities as
described in § 30.18 of this chapter).
(5) Security requirements must be
implemented for the protection of SNM
based on the type, enrichment, and
quantity in accordance with 10 CFR part
73, as applicable, and for the protection
of Category 1 and Category 2 quantities
of radioactive material in accordance
with 10 CFR part 37, as applicable.
(d) Fuel loading. (1)(i) An ML may
authorize possession of a manufactured
reactor into which the licensee has
loaded fresh (unirradiated) fuel
pursuant to a license issued under part
70 of this chapter only if the
manufactured reactor is configured
during its loading, storage, and transport
with at least two independent physical
mechanisms in place, each of which is
sufficient to prevent criticality assuming
optimum neutron moderation and
neutron reflection conditions.
(ii) The ML applicant may file a
separate, subsequent application for the
10 CFR part 70 license or combine the
application for the 10 CFR part 70
license with the application for an ML.
(iii) The Commission has determined
that any such fueled manufactured
reactor in which the independent
physical mechanisms to prevent
criticality have been installed is not in
operation.
(iv) Upon installation of the fueled
manufactured reactor in its place of
operation and a Commission finding
that the acceptance criteria in the COL
that authorized reactor construction are
met under § 53.1452(g), the independent
physical mechanisms to prevent
criticality may be removed. Upon
initiating the physical removal of any
one of the independent physical
mechanisms to prevent criticality, the
fueled manufactured reactor has
commenced operation.
(2) Holders of 10 CFR part 70 licenses
authorizing the possession and loading
of fresh fuel into manufactured reactors
must comply with the requirements of
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10 CFR part 70 for the facilities and
activities related to the storage,
movement, and loading of fresh fuel in
the manufactured reactor. Holders of
these 10 CFR part 70 licenses must
comply with the requirements of
Subpart H to 10 CFR part 70, regardless
of whether their proposed activities
meet the applicability criteria found in
10 CFR 70.60. Procedures, equipment,
and personnel required by the 10 CFR
part 70 license, must be in place before
the receipt of SNM at the manufacturing
facility.
(i) Before the receipt of SNM, the
licensee must have security programs in
place that meet the performance
objectives of 10 CFR 73.67, with the
following additions and exceptions:
(A) A physical security plan
describing the physical security
program must be maintained and a
cybersecurity program must be
established for the possession and
loading of fresh fuel into a
manufactured reactor authorized by a 10
CFR part 70 license, regardless of fuel
type, enrichment, and quantity.
(B) The physical security program
must be designed to prevent unintended
and uncontrolled criticality events.
(C) The cybersecurity program must
provide reasonable assurance that a
cyberattack would not adversely impact
the functions performed by digital assets
used by the licensee for implementing
the physical security requirements of
this section, or the radiation monitoring
and criticality requirements in this
section or in 10 CFR part 70.
(D) All holders of a part 70 license
that authorizes loading of fresh fuel into
a manufactured reactor must perform
the screening required in § 73.67(d)(4) of
this chapter to confirm the identity,
trustworthiness, and reliability of
individuals prior to granting unescorted
access to special nuclear material; these
determinations must be documented.
(ii) [Reserved]
(3) The loading or unloading of fresh
fuel into or from a manufactured reactor
and any changes to the configuration of
reactivity control and prevention
systems for the fueled manufactured
reactor must be performed by a certified
fuel handler meeting the requirements
in subpart F of this part.
(e) Transportation. (1) A holder of an
ML may not transport or allow to be
removed from the places of manufacture
the manufactured reactor or portions
thereof as defined in the ML except for
transport to a site for which the
Commission has issued a COL that
references the subject ML.
(2) A holder of an ML must include
in any contract governing the transport
of a manufactured reactor or portions
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thereof as defined in the ML from the
places of manufacture to any other
location, a provision requiring that the
person transporting the manufactured
reactor comply with all shipping
requirements in applicable NRC
regulations, certificates of compliance,
and NRC-issued licenses.
(3) Procedures governing the
preparation of the manufactured reactor
or portions thereof as defined in the ML
for transport and the conduct of the
transport must be issued prior to
transport. The procedures must
implement the protective measures and
restrictions described in NRC
regulations and NRC-issued licenses to
protect the reactor from potential
conditions that would adversely affect
the safe operation of a commercial
nuclear plant.
(4) For a manufactured reactor that is
to be loaded with fresh fuel before
transport to the place of operation, the
ML must specify that transportation will
be in accordance with parts 71 and 73
of this chapter.
(f) Acceptance and installation at the
site for which the Commission has
issued a COL that references the subject
ML. (1) Installation at the site for which
the Commission has issued a COL that
references the subject ML must follow
the regulations in § 53.610.
(2) Upon arrival at the site, the
manufactured reactor or portions of a
manufactured reactor may not be
installed in its place of operation unless
the COL holder performs inspections
sufficient to verify the reactor is in
compliance with the ML and has not
been damaged in transit. The COL
holder must perform these inspections
in accordance with documented
procedures subject to quality assurance
measures commensurate with their
importance to safety. In addition,
inspections must confirm that the
interface requirements between the
manufactured reactor or portions of a
manufactured reactor and the remaining
portions of the commercial nuclear
plant are met.
Subpart F—Requirements for
Operation
§ 53.700
Operational objectives.
(a) Each holder of an operating license
(OL) or combined license (COL) under
this part must develop, implement, and
maintain controls for plant structures,
systems, and components (SSCs),
responsibilities of plant personnel, and
plant programs during the operating life
of each commercial nuclear plant such
that the requirements defined in subpart
B are satisfied. More specifically:
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(1) Each holder of an OL or COL
under this part must maintain the
capabilities, availability, and reliability
of plant SSCs to ensure that the safety
functions identified in § 53.230 will be
performed if called upon during
licensing-basis events (LBEs).
(2) Each holder of an OL or COL
under this part must ensure that plant
personnel have adequate knowledge and
skills to perform their assigned duties
that support the performance of the
safety functions identified in § 53.230.
(3) Each holder of an OL or COL
under this part must implement plant
programs sufficient to ensure that the
safety functions identified in § 53.230
will be performed if called upon during
normal operations and LBEs.
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§ 53.710 Maintaining capabilities and
availability of structures, systems, and
components.
Controls must be provided for each
commercial nuclear plant licensed
under this part such that the
capabilities, availability, and reliability
of plant SSCs, when combined with
corresponding programmatic controls
and human actions, provide that the
safety criteria defined in §§ 53.210 and
53.220 will be met.
(a) Technical specifications must be
developed, implemented, and
maintained that define conditions or
limitations on plant operations that are
necessary to ensure that safety-related
(SR) SSCs can fulfill the safety functions
identified under § 53.230 and support
meeting the safety criteria of § 53.210.
The technical specifications must
describe the following requirements:
(1) Limits on the inventory of
radioactive materials within the reactor
system and supporting systems with the
potential, individually or collectively, to
cause a release exceeding the safety
criteria in § 53.210 as a result of a
design-basis accident analyzed in
accordance with § 53.450(f).
(2) Operating limits for the facility
that if exceeded could lead to a failure
to perform a required safety function
necessary to demonstrate compliance
with the safety criteria in § 53.210.
(3) For each SSC classified as SR in
accordance with § 53.460, technical
specifications must define—
(i) Limiting conditions for operation.
Limiting conditions for operation are
the lowest functional capability or
performance levels of SR SSCs required
to ensure that the design-basis accidents
analyzed in accordance with § 53.450(f)
satisfy the safety criteria of § 53.210.
When a limiting condition for operation
is not met, the licensee must shut down
the plant or follow any remedial action
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permitted by the technical
specifications until the condition can be
met.
(ii) Surveillance requirements.
Surveillance requirements are
requirements relating to test, calibration,
or inspection to assure that the
necessary quality of systems and
components is maintained and that the
limiting conditions for operation will be
met.
(4) Design elements to be included are
those elements of the plant such as
materials of construction and geometric
arrangements, which, if altered or
modified, would have a significant
effect on safety and are not covered in
categories described in paragraphs (a)(1)
through (3) of this section.
(5) Administrative controls are the
provisions relating to organization and
management, procedures,
recordkeeping, review and audit, and
reporting necessary to assure operation
of the plant in a safe manner. Each
licensee must submit any reports to the
Commission pursuant to approved
technical specifications under § 53.040.
(b) Controls on plant operations,
including availability controls, must be
developed and implemented to ensure
that the configurations and special
treatments for SR SSCs and non-safetyrelated but safety-significant (NSRSS)
SSCs provide the capabilities,
availability, and reliability required to
demonstrate compliance with the
criteria of §§ 53.220 and 53.450(e).1 The
controls must—
1 The
comprehensive risk metrics and
related risk performance objectives
established under § 53.220 involve assessing
and averaging the risks over a defined period
(e.g., plant year) and do not constitute a realtime requirement that must be continuously
demonstrated by the licensee.
(1)(i) Identify who within the
commercial nuclear plant has authority
to make configuration changes;
(ii) Establish processes to make
configuration changes to NSRSS SSCs;
and
(iii) Establish processes to ensure that
all organizations of the commercial
nuclear plant affected by the
configuration changes are formally
notified and approve of the change.
(2) Describe how the special
treatments for each NSRSS SSC and
special treatments for SR SSCs beyond
those under paragraph (a) of this section
will be established and maintained over
the operating life of the commercial
nuclear plant.
§ 53.715 Maintenance, repair, and
inspection programs.
(a) A program to control maintenance
activities and monitor the performance
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or condition of SR and NSRSS SSCs
must be developed, implemented, and
maintained.
(b) Whenever a licensee determines
through activities related to
maintenance, repair, and inspection of
SSCs, the activities under § 53.710, or
otherwise that the performance or
condition of an SR or NSRSS SSC does
not demonstrate compliance with
established special treatments or
performance goals related to
capabilities, availability, or reliability,
the licensee must take appropriate
corrective action.
(c) Performance and condition
monitoring activities and associated
goals and preventive maintenance
activities must be evaluated at least
every 24 months. The evaluations must
take into account, where practical,
industry-wide operating experience.
Adjustments must be made where
necessary to ensure that the objective of
preventing failures of SSCs through
maintenance is appropriately balanced
against the objective of minimizing
unavailability of SSCs due to
monitoring or preventive maintenance.
(d) Before performing maintenance
activities (including but not limited to
surveillance, post-maintenance testing,
and corrective and preventive
maintenance), the licensee must assess
and manage the increase in risk that
may result from the proposed
maintenance activities.
§ 53.720
Response to seismic events.
If vibratory ground motion exceeding
that of the operating basis earthquake
ground motion or significant plant
damage due to vibratory ground motion
occurs, the licensee must shut down the
commercial nuclear plant. If structures,
systems, or components necessary for
the safe shutdown of the commercial
nuclear plant are not available after the
occurrence of this vibratory ground
motion, the licensee must consult with
the Commission and must propose a
plan for the timely, safe shutdown of the
commercial nuclear plant. Prior to
resuming operations, the licensee must
demonstrate to the Commission that
those features necessary for continued
operation without undue risk to the
health and safety of the public or
necessary to maintain the licensing
basis of the commercial nuclear plant
were either not functionally damaged or
have been repaired.
§ 53.725 General staffing, training,
personnel qualifications, and human factors
requirements.
(a) Two classes of commercial nuclear
plants. Commercial nuclear plants
licensed under this part are either of the
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class, based upon the similarity of
operating and technical characteristics
of the plants in the class, of self-reliantmitigation facilities or of interactiondependent-mitigation facilities. A
commercial nuclear plant is a selfreliant-mitigation facility if the U.S.
Nuclear Regulatory Commission (NRC)
determined as part of its approval of the
OL or COL for that plant that its design
demonstrates compliance with the
criteria of § 53.800(a)(1) through (a)(5).
Otherwise, the commercial nuclear
plant is an interaction-dependentmitigation facility.
(b) Purpose and applicability. The
regulations in §§ 53.725 through 53.830
address areas related to staffing,
training, personnel qualifications, and
human factors engineering for
applicants for or holders of OLs or COLs
under this part. These regulations are
organized as follows:
(1) Sections 53.725 through 53.745
address general requirements for
staffing, training, personnel
qualifications, and human factors
engineering. The regulations within
these sections are applicable to all
applicants for or holders of OLs or COLs
under this part, except where
specifically stated otherwise.
(2) Sections 53.760 through 53.795
address operator and senior operator
licensing requirements. The regulations
within these sections are applicable to
those applicants for or holders of OLs or
COLs under this part for interactiondependent-mitigation facilities that have
not yet certified the permanent
cessation of operations and permanent
removal of fuel from the reactor vessel
as described under § 53.1070.
(3) Sections 53.800 through 53.820
address generally licensed reactor
operator requirements. The regulations
within these sections are in lieu of
§§ 53.760 through 53.795 for those
applicants for or holders of OLs or COLs
under this part for self-reliant-mitigation
facilities that have not yet certified the
permanent cessation of operations and
permanent removal of fuel from the
reactor vessel as described under
§ 53.1070.
(4) Section 53.830 provides general
personnel training requirements. The
regulations within this section are
applicable to all applicants for or
holders of OLs or COLs under this part.
(c) Definitions. When used in
§§ 53.725 through 53.830:
Applicant refers to an applicant for an
operator or senior operator license;
licensee refers to the holder of an
operator, senior operator, or generally
licensed reactor operator license; and
facility licensee refers to the licensee for
the commercial nuclear plant where the
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applicant would be licensed or the
licensee is licensed.
Automation means a device or system
that accomplishes (partially or fully) a
function or task.
Auxiliary operator means any
individual who operates components of
a commercial nuclear plant but does not
manipulate controls or direct the
manipulation of controls of the plant
and is not required to be licensed under
the provisions of this part.
Controls when used with respect to a
nuclear reactor means apparatus and
mechanisms, the manipulation of which
directly affects the reactivity or power
level of the reactor.
Generally licensed reactor operator
means any individual licensed under
the provisions of § 53.810 to manipulate
controls of a self-reliant-mitigation
facility and to direct the licensed
activities of generally licensed reactor
operators.
Interaction-dependent-mitigation
facility means a commercial nuclear
plant design other than one that
demonstrates compliance with the
operating and technical characteristics
defined under § 53.800.
Load following means a commercial
nuclear plant automatically changing its
output to match expected demand in
response to externally originated
instructions or signals.
Operator means any individual
licensed under the provisions of
§§ 53.760 through 53.795 to manipulate
controls of an interaction-dependentmitigation facility.
Performance testing means testing
conducted to verify a simulation
facility’s performance as compared to
actual or predicted reference plant
performance.
Reference plant means the specific
commercial nuclear plant on which a
simulation facility’s configuration,
system control arrangement, and design
data are based. The reference plant may
or may not be constructed.
Self-reliant-mitigation facility means a
commercial nuclear plant design that
demonstrates compliance with the
operating and technical characteristics
defined under § 53.800.
Senior operator means any individual
licensed under the provisions of
§§ 53.760 through 53.795 to manipulate
controls of an interaction-dependentmitigation facility and to direct the
licensed activities of operators.
Simulation facility means an interface
designed to provide a realistic imitation
of the operation of a commercial nuclear
plant used for the administration of
examinations, for training, and/or to
demonstrate compliance with
experience requirements for applicants
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or licensees. A simulation facility may
rely, in whole or part, upon the physical
utilization of the reference plant itself.
Systems approach to training means a
training program that includes the
following five elements:
(1) Systematic analysis of the jobs to
be performed.
(2) Learning objectives derived from
the analysis which describe desired
performance after training.
(3) Training design and
implementation based on the learning
objectives.
(4) Evaluation of trainee mastery of
the objectives during training.
(5) Evaluation and revision of the
training based on the performance of
trained personnel in the job setting.
§ 53.726
Communications.
(a) An applicant or licensee or facility
licensee must submit any
communication or report required by
the regulations contained within
§§ 53.725 through 53.830 and must
submit any application filed under these
regulations to the Commission.
(b) Each licensee that is required to
comply with the requirements of
§§ 53.760 through 53.795 (i.e.,
interaction-dependent-mitigation
facilities) must notify the appropriate
NRC contact within 30 days of the
following in regard to a licensed
operator or senior operator:
(1) Permanent reassignment from the
position for which the licensee has
certified the need for a licensed operator
or senior operator under § 53.775(a)(1);
(2) Termination of any operator or
senior operator; or
(3) Permanent disability or illness as
required under § 55.770 of this chapter.
§ 53.728 Completeness and accuracy of
information.
Information provided to the
Commission by an applicant for an
operator or senior operator license or by
a licensee or information required by
statute or by the Commission’s
regulations, orders, or license
conditions to be maintained by the
applicant or the licensee must be
complete and accurate in all material
respects.
§ 53.730 Defining, fulfilling, and
maintaining the role of personnel in
ensuring safe operations.
Each applicant for or holder of an OL
or COL for a commercial nuclear plant
under this part must comply with the
following:
(a) Human factors engineering design
requirements. The plant design must
reflect state-of-the-art human factors
engineering principles for safe and
reliable performance in all locations that
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human activities are expected for
performing or supporting the continued
availability of plant safety or emergency
response functions.
(b) Human system interface design
requirements. The plant design must
provide for the following to support
operating personnel in monitoring plant
conditions and responding to plant
events:
(1) Features for displaying to
operating personnel a minimum set of
parameters that define the safety status
of the plant and are capable of
displaying both the full range of
important plant parameters and data
trends on demand, as well as indicating
when process limits are being
approached or exceeded;
(2) Automatic indication of the
bypassed and operable status of safety
systems;
(3) Direct indication of SSC status that
relates to the ability of the SSC to
perform its safety function, such as
relief and safety valve position (i.e.,
open or closed) for barriers important to
fulfilling safety functions of with such
devices, and ultimate heat sink and
cooling system status and availability;
(4) Instrumentation to measure,
record, and display key plant
parameters related to the performance of
SSCs and the integrity of barriers
important to fulfilling safety functions
to support operators in monitoring plant
conditions and responding to plant
events. Examples include temperatures
and pressures within important systems
or structures, core or fuel system
conditions (including possible damage
states), temperatures and levels
associated with cooling functions,
combustible gas concentrations,
radiation levels in systems and within
structures, and radioactive effluent
releases;
(5) Leakage control and detection in
the design of systems that pass through
barriers important to fulfilling safety
functions for the release of
radionuclides. An example is an SSC
that penetrates a containment structure
that might contain radioactive materials
that could contribute to the source term
during an accident;
(6) Monitoring of in-plant radiation
and airborne radioactivity as
appropriate for a broad range of normal
operating and accident conditions; and
(7) For self-reliant-mitigation
facilities, the plant design must also
provide the generally licensed reactor
operators with the capability to do the
following:
(i) Receive plant operating data,
including reactor parameters and
information needed for the evaluation of
emergency conditions.
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(ii) Immediately initiate a reactor
shutdown from their location.
(iii) Promptly dispatch operations and
maintenance personnel.
(iv) Immediately implement
responsibilities under the facility
emergency plan, as applicable.
(c) Concept of operations. A concept
of operations that is of sufficient scope
and detail to address the following must
be provided:
(1) Plant goals;
(2) The roles and responsibilities of
operating personnel and automation (or
any combination thereof) that are
responsible for completing plant
functions;
(3) Staffing, qualifications, and
training;
(4) The management of normal
operations;
(5) The management of off-normal
conditions and emergencies;
(6) The management of maintenance
and modifications; and
(7) The management of tests,
inspections, and surveillances.
(d) Functional requirements analysis
and function allocation. A functional
requirements analysis and a function
allocation must be provided that are
sufficient to demonstrate compliance
with the following:
(1) The functional requirements
analysis must address how safety
functions and functional safety criteria
are satisfied, and
(2) The function allocation must
describe how the safety functions will
be assigned to human action,
automation, active safety features,
passive safety features, and/or inherent
safety characteristics.
(e) Operating experience. A program,
during construction and during
operation, as applicable, for evaluating
and applying operating experience must
be developed, implemented, and
maintained.
(f) Staffing plan. A staffing plan must
be developed and comply with the
following:
(1) The staffing plan must include a
description of how engineering
expertise will be available to the onshift operating personnel during all
plant conditions, to assist if they
encounter a situation not covered by
procedures or training. Engineering
expertise includes familiarity with the
operation of the plant for which the
expertise is provided and one of the
following:
(i) A bachelor’s degree in engineering,
engineering technology, or physical
science from an institution accredited
by a U.S. government recognized
accrediting body or equivalent; or
(ii) A Professional Engineer’s license
from a U.S. State or territory.
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(2) Applicants for or holders of OLs or
COLs for interaction-dependentmitigation facilities must include within
their staffing plans a description of how
the proposed numbers, positions, and
qualifications of operators and senior
operators across all modes of plant
operations will be sufficient to ensure
that plant safety functions will be
maintained. This description must be
supported by human factors engineering
analyses and assessments.
(3) Applicants for or holders of OLs or
COLs for self-reliant-mitigation facilities
must include within their staffing plans
a description of how generally licensed
reactor operator staffing that is both
sufficient to continually monitor the
operations of fueled reactors and to
provide for a continuity of
responsibility for facility operations at
all times during the operating phase will
be maintained.
(4) Applicants for or holders of OLs or
COLs under this part must include
within their staffing plans a description
of how the numbers, positions, and
responsibilities of personnel contained
within those plans will adequately
support all necessary functions within
areas such as plant operations,
equipment surveillance and
maintenance, radiological protection,
chemistry control, fire brigades,
engineering, security, and emergency
response.
(5) The staffing plan must be
approved by the NRC as part of its
approval of the OL or COL for the plant.
The approved staffing plan is subject to
the requirements of § 53.1565.
(g) Training, examination, and
proficiency programs. Develop,
implement, and maintain programs that
comply with the following
requirements. These programs must be
approved by the NRC as part of its
approval of the OL or COL for the plant:
(1) For those applicants for or holders
of OLs or COLs for interactiondependent-mitigation facilities:
(i) The operator licensing initial
training program required under
§ 53.780(a);
(ii) The operator licensing initial
examination program required under
§ 53.780(b);
(iii) The operator licensing
requalification program required under
§ 53.780(c); and
(iv) The operator proficiency program
required under § 53.780(g).
(2) For those applicants for or holders
of OLs or COLs for self-reliantmitigation facilities, the generally
licensed reactor operator training,
examination, and proficiency programs
required under § 53.815.
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(3) The operator licensing
requalification programs required under
§ 53.780(c) or § 53.815(b) must be
implemented upon commencing the
administration of initial examinations
under the operator licensing
examination program required under
§ 53.780(b) or § 53.815(b), respectively.
§ 53.735
General exemptions.
The regulations in §§ 53.725 through
53.830 do not require a license for an
individual who—
(a) Under the direction and in the
presence of an operator or senior
operator or a generally licensed reactor
operator, as appropriate, manipulates
the controls of a commercial nuclear
plant as a part of the individual’s
training in a facility licensee’s training
program as approved by the
Commission to qualify for an operator or
senior operator license or a generally
licensed reactor operator license there,
as appropriate, under these regulations;
or
(b) Under the direction and in the
presence of a senior operator or
generally licensed reactor operator, as
appropriate, manipulates the controls of
a commercial nuclear plant to load or
unload the fuel into, out of, or within
the reactor vessel while the reactor is
not operating.
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§ 53.740 Facility licensee requirements—
General.
(a) Facility licensees must
demonstrate compliance with the
requirements of either §§ 53.760 through
53.795 for interaction-dependentmitigation facilities or §§ 53.800 through
53.820 for self-reliant-mitigation
facilities.
(b) The facility licensee must
maintain the staffing complement
described under its approved facility
staffing plan until such time as the
permanent cessation of operations and
permanent removal of fuel from the
reactor vessel has been certified as
described under § 53.1070. The
approved staffing plan is subject to the
requirements of § 53.1565.
(c) Except as provided under § 53.735,
the facility licensee may not permit the
manipulation of the controls of a
commercial nuclear plant by anyone
who is not an operator or senior
operator or generally licensed reactor
operator, as appropriate.
(d) Facility licensees for interactiondependent-mitigation facilities that have
not yet certified the permanent
cessation of operations and permanent
removal of fuel from the reactor vessel
as described under § 53.1070 must
designate senior operators to be
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responsible for supervising the licensed
activities of operators.
(e) Apparatus and mechanisms other
than controls, the operation of which
may affect the reactivity or power level
of a reactor, must be manipulated only
while plant conditions are being
monitored by an individual who is an
operator or senior operator or a
generally licensed reactor operator, as
appropriate.
(f)(1) Load following is permitted if at
least one of the following is
immediately capable of refusing
demands when they could challenge the
safe operation of the plant or when
precluded by the plant equipment
conditions:
(i) The actuation of an automatic
protection system that utilizes setpoints
more conservative than those otherwise
credited for the purposes of reactor
protection; or
(ii) An automated control system; or
(iii) An operator or senior operator or
a generally licensed reactor operator, as
appropriate.
(2) The provisions of paragraph (e) of
this section do not apply during load
following operations.
(g)(1) Facility licensees for
interaction-dependent-mitigation
facilities must have present during
alteration of the core (including fuel
loading or transfer) an individual
holding a senior operator license, or a
senior operator license limited to fuel
handling to directly supervise the
activity and, during this time, the
facility licensee must not assign other
duties to this person.
(2) Facility licensees for self-reliantmitigation facilities must have present
during alteration of the core (including
fuel loading or transfer) an individual
holding a generally licensed reactor
operator license to directly supervise
the activity and, during this time, the
facility licensee must not assign other
duties to this person.
(3) The provisions of paragraphs (g)(1)
and (2) of this section do not apply to
core alterations performed as part of
refueling operations while a facility that
is capable of online refueling is
operating at power.
(h) Facility licensees may take
reasonable action that departs from a
license condition or a technical
specification (contained in a license
issued under this part) in an emergency
when this action is immediately needed
to protect the public health and safety
and no action consistent with license
conditions and technical specifications
that can provide adequate or equivalent
protection is immediately apparent.
Such facility licensee action must be
approved, as a minimum, by a senior
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operator or a generally licensed reactor
operator, as applicable, or, after
certifying the permanent cessation of
operations and permanent removal of
fuel from the reactor vessel as described
under § 53.1070 by a certified fuel
handler, senior operator, or generally
licensed reactor operator, as applicable,
prior to taking the action.
§ 53.745
Operator license requirements.
A person must be authorized by a
license issued by the Commission to
perform the function of an operator,
senior operator, or generally licensed
reactor operator as defined in this part.
§ 53.760
Operator licensing.
(a) Applicability. Sections 53.760
through 53.795 address operator and
senior operator licensing requirements.
The regulations within these sections
are applicable to those applicants for or
holders of OLs or COLs under this part
for interaction-dependent-mitigation
facilities that have not yet certified the
permanent cessation of operations and
permanent removal of fuel from the
reactor vessel as described under
§ 53.1070.
(b) Reserved.
§ 53.765
Medical requirements.
(a) An applicant for an operator or
senior operator license must have a
medical examination by a physician. An
operator or senior operator must have a
medical examination by a physician
every 2 years.
(b) To certify the medical fitness of an
applicant for an operator or senior
operator license, an authorized
representative of the facility licensee
must complete and sign NRC Form 396,
‘‘Certification of Medical Examination
by Facility Licensee,’’ which can be
obtained by writing the Office of the
Chief Information Officer, U.S. Nuclear
Regulatory Commission, Washington,
DC 20555–0001, by calling 301–415–
7232, or by visiting the NRC’s website
at https://www.nrc.gov and selecting
forms from the index found on the home
page, or by other means provided by the
NRC.
(1) Form NRC 396 must certify that a
physician has conducted the medical
examination of the applicant as required
in paragraph (a) of this section.
(2) When the medical certification
requests a conditional license based on
medical evidence, the medical evidence
must be submitted on NRC Form 396 to
the Commission to enable the
Commission to make a determination in
accordance with § 53.775(b).
(c) The facility licensee must
document and maintain the results of
medical qualifications data, test results,
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and each operator’s or senior operator’s
medical history for the current license
period and provide the documentation
to the Commission upon request. The
facility licensee must retain this
documentation while an individual
performs the functions of an operator or
senior operator.
§ 53.770 Incapacitation because of
disability or illness.
If, during the term of the operator or
senior operator license, the licensee
develops a permanent physical or
mental condition that causes the
licensee to fail to demonstrate
compliance with the requirements of
§ 53.775(b)(1)(i), the facility licensee
must notify the Commission within 30
days of learning of the diagnosis. For
conditions for which a conditional
license (as described in § 53.775(b)) is
requested, the facility licensee must
provide medical certification on Form
NRC 396 to the Commission (as
described in § 53.765(b)).
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§ 53.775 Applications for operators and
senior operators.
(a) How to apply. (1) The applicant for
an operator or senior operator license
must—
(i) Complete NRC Form 398,
‘‘Personal Qualification Statement—
Licensee,’’ which can be obtained by
writing the Office of the Chief
Information Officer, U.S. Nuclear
Regulatory Commission, Washington,
DC 20555–0001, by calling 301–415–
5877, or by visiting the NRC’s website
at https://www.nrc.gov and selecting
forms from the index found on the home
page, or by other means provided by the
NRC;
(ii) File an original of NRC Form 398,
or an equivalent electronic submittal,
together with the information required
in paragraphs (a)(1)(iii) and (a)(1)(iv) of
this section, with the appropriate
Regional Administrator.
(iii) Provide evidence that the
applicant, as a trainee, has successfully
demonstrated competence in
manipulating the controls of either the
facility for which a license is sought or
a simulation facility that demonstrates
compliance with the requirements of
§ 53.780(e). For operators applying for a
senior operator license, certification that
the operator has successfully operated
the controls of the facility as an operator
will be accepted; and
(iv) Provide certification by the
facility licensee of medical condition
and general health on Form NRC 396, to
comply with § 53.765.
(2) The Commission may at any time
after the application has been filed, and
before the license has expired, require
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further information under oath or
affirmation to enable it to determine
whether to grant or deny the application
or whether to revoke, modify, or
suspend the license.
(3) An applicant whose application
has been denied because of a medical
condition or their general health may
submit a further medical report at any
time as a supplement to the application.
(4) Each application and statement
must contain complete and accurate
disclosure as to all matters required to
be disclosed. The applicant must sign
statements required by paragraphs
(a)(1)(i) and (a)(1)(ii) of this section.
(b) Disposition of an initial
application. (1) License approval. The
Commission will approve an initial
application if it finds that the following
criteria are met:
(i) Health. The applicant’s medical
condition and general health will not
adversely affect the performance of
assigned operator or senior operator job
duties or cause operational errors
endangering public health and safety.
The Commission will base its finding
upon the certification by the facility
licensee as detailed in § 53.765(b).
(ii) Examination. The applicant has
passed the requisite examination in
accordance with § 53.780(b). The
examination determines whether the
applicant for an operator’s or senior
operator’s license has learned to operate
a facility competently and safely, and
additionally, in the case of a senior
operator, whether the applicant has
learned to supervise the licensed
activities of operators competently and
safely.
(2) Conditional license. If an
applicant’s general medical condition
does not demonstrate compliance with
the minimum standards under
§ 53.775(b)(1)(i) of this section, the
Commission may approve the
application and include conditions in
the license to accommodate the medical
condition. The Commission will
consider the recommendations and
supporting evidence of the facility
licensee and of the examining physician
(provided on Form NRC 396) in arriving
at its decision.
(c) Re-applications. (1) An applicant
whose application for a license has been
denied because of failure to pass the
examination may file a new application.
The application must be submitted on
Form NRC 398 and include a statement
signed by an authorized representative
of the facility licensee by whom the
applicant will be employed that states
in detail the extent of the applicant’s
additional training and remediation
since the denial and certifies that the
applicant is ready for re-examination.
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(2) An applicant who has passed a
portion of the examination and failed
another may request in a new
application on Form NRC 398 to be
excused from re-examination on the
portions of the examination that the
applicant has passed. The Commission
may in its discretion grant the request
if it determines that sufficient
justification is presented.
§ 53.780 Training, examination, and
proficiency program.
(a) Operator licensing initial training
program. (1) A program that is based
upon a systems approach to training, as
defined by § 53.725(b), must be utilized
for the training of applicants for
operator and senior operator licenses.
The program must ensure that
applicants at the facility will possess the
knowledge, skills, and abilities
necessary to protect the public health
and maintain those plant safety
functions specific to the facility design.
The program must be approved by the
Commission prior to its use for training
applicants, as described under
§ 53.730(g). The approved operator
licensing initial training program is
subject to the requirements of § 53.1565.
(2) The facility licensee must
maintain operator licensing initial
training program records documenting
the initial operator licensing training
administered and completed by each
applicant. The facility licensee must
retain these records during the period in
which any trainees subsequently remain
licensed as operators or senior operators
at the facility.
(b) Operator licensing initial
examination program. (1) The facility
licensee must establish and implement
an examination program for testing a
representative sample of the knowledge,
skills, and abilities needed to safely
perform operator and senior operator
duties, to include both the examination
methods and criteria to be used to assess
passing performance. The program must
provide for valid and reliable
examinations and be approved by the
Commission prior to its use for
examining applicants, as described
under § 53.730(g). The approved
operator licensing initial examination
program is subject to the requirements
of § 53.1565.
(2) The facility licensee must submit
prepared examinations to the
Commission for review and approval in
advance of their administration.
(3) The Commission will either
administer an approved examination or
allow the facility licensee to administer
the examination. The facility licensee
must ensure that sufficient advance
notification is provided to the
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Commission to either administer the
examination or allow for a
representative of the Commission to be
afforded the opportunity to be present
when the facility licensee administers
the examination.
(4) Graded examination
documentation for each applicant must
be promptly provided to the
Commission for review in making
operator licensing decisions.
(5) The facility licensee must
maintain operator licensing initial
examination program records
documenting the participation of each
operator and senior operator applicant
in the initial examination. The records
must contain copies of examinations
administered, the answers given by the
applicant, documentation of the grading
of examinations, and documentation of
any additional training administered in
areas in which an applicant exhibited
deficiencies. The facility licensee must
retain these records during the period in
which the associated operators or senior
operators remain licensed at the facility.
(c) Operator licensing requalification
program. (1) A program based upon a
systems approach to training, as defined
by § 53.725(b), must be utilized for the
continuing training of operators and
senior operators.
(i) The program must ensure that
operators and senior operators at the
facility maintain the knowledge, skills,
and abilities necessary to protect the
public health and maintain those plant
safety functions specific to the facility
design. The program must be conducted
for a continuous period not to exceed 24
months in duration.
(ii) The program must be approved by
the Commission prior to its use for
continuing training, as described under
§ 53.730(g). The approved operator
licensing requalification program is
subject to the requirements of § 53.1565.
(2) The following requirements apply
to operator licensing requalification
programs:
(i) The facility licensee must propose
a requalification examination program
for testing, for each requalification
period, a sample of the topics included
under the systems approach to training,
to include both the examination
methods and criteria to be used to assess
passing performance. The program must
provide for valid and reliable
examinations and be approved by the
Commission prior to its use for
examining operators and senior
operators, as described under
§ 53.730(g). The approved
requalification examination program is
subject to the requirements of § 53.1565.
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(ii) The following requirements apply
to the requalification examination
program:
(A) The facility licensee must make
prepared requalification examinations
available to the Commission for review.
(B) The facility licensee must ensure
that a representative of the Commission
is afforded the opportunity to be present
during requalification examination
administration.
(C) The facility licensee must ensure
that each operator and senior operator is
administered a complete requalification
examination on a periodicity not to
exceed 24 months. Additionally, the
facility licensee must ensure that any
licensed operator or senior licensed
operator who either demonstrates
unsatisfactory performance on, or fails
to complete, the biennial requalification
examination is removed from the
performance of licensed operator and
senior licensed operator duties until
such time that any necessary remedial
training has been completed and a
retake examination has been passed.
(D) The facility licensee must
promptly provide a summary of
examination results for each operator
and senior operator following the
completion of the requalification
examination.
(3) The facility licensee must
maintain operator licensing
requalification program records
documenting the participation of each
operator and senior operator in the
requalification program. The records
must contain copies of examinations
administered, the answers given by the
operator or senior operator,
documentation of the grading of
examinations, and documentation of
any additional training administered in
areas in which an operator or senior
operator exhibited deficiencies. The
facility licensee must retain these
records until the operator’s or senior
operator’s license is renewed.
(d) Examination integrity. Applicants,
operators and senior operators, and
facility licensees must not engage in any
activity that compromises the integrity
of any application or examination
required by §§ 53.760 through 53.795.
The integrity of an examination is
considered compromised if any activity,
regardless of intent, affected, or, but for
detection, could have affected the
equitable and consistent administration
of the examination. This includes
activities related to the preparation and
certification of applications and all
activities related to the preparation,
administration, and grading of
examinations required by §§ 53.760
through 53.795.
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(e) Simulation facilities. (1) This
section addresses the use of a
simulation facility for the
administration of examinations, for
training, or to demonstrate compliance
with experience requirements for
applicants for operator and senior
operator licenses.
(2) Simulation facilities used for
training purposes, for demonstrating
compliance with experience
requirements, or for the conduct of
examinations under § 53.780(b) and (c)
must demonstrate compliance with the
following criteria as they relate to the
facility licensee’s reference plant:
(i) The simulation facility must be of
sufficient scope and fidelity for
individuals to acquire and demonstrate
the necessary knowledge, skills, and
abilities to safely perform operator and
senior operator duties.
(ii) The simulation facility must
utilize models relating to nuclear,
thermal-hydraulic, and other applicable
design-specific characteristics that
either replicate the most recent fuel load
in the reference commercial nuclear
plant or, prior to initial fuel load (or, for
a fueled manufactured reactor, prior to
initiating the physical removal of any
one of the independent physical
mechanisms to prevent criticality
required under § 53.620(d)(1)), replicate
the intended initial fuel load for the
reference commercial nuclear plant,
with the exception of those portions of
the simulation facility that utilize the
reference plant itself.
(iii) Simulation facility fidelity must
be demonstrated so that significant
control manipulations are completed
without procedural exceptions,
simulator performance exceptions, or
deviation from the approved training
scenario sequence.
(3) Facility licensees that maintain a
simulation facility that has been
approved by the Commission for
training purposes, demonstrating
compliance with experience
requirements, or the conduct of
examinations under § 53.780(b) and (c)
for the facility licensee’s reference plant
must:
(i) Conduct performance testing
throughout the life of the simulation
facility in a manner sufficient to ensure
that paragraph (e)(2) of this section is
met;
(ii) Retain the results of performance
testing for 4 years after the completion
of each performance test or until
superseded by updated test results;
(iii) Promptly correct modeling and
hardware discrepancies and
discrepancies identified from scenario
validation and from performance testing
or provide justification as to why the
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presence of such discrepancies will not
adversely affect simulator performance
with respect to the criteria of paragraph
(e)(2) of this section;
(iv) Make the results of any
uncorrected performance test failures
that may exist at the time of the initial
license examination or requalification
examination available for NRC review,
prior to or concurrent with preparations
for each initial license examination or
requalification examination; and
(v) Maintain the provisions for license
application and examination integrity
consistent with § 53.780(d).
(4) A simulation facility must
demonstrate compliance with the
requirements of paragraphs (e)(2) and
(e)(3) of this section for the Commission
to accept the simulation facility for
conducting initial examinations as
described in § 53.780(b), requalification
training as described in § 53.780(c), or
performing control manipulations that
affect reactivity to establish eligibility
for an operator or senior operator
license as described in § 53.775(a).
(f) Waiver of examination
requirement. On application, the
Commission may waive any or all of the
requirements for an examination if it
finds that the applicant has
demonstrated the required knowledge,
skills, and abilities to safely operate the
plant, and is capable of continuing to do
so. The Commission may make such a
finding based on demonstration of the
following:
(1) Operating experience at a
comparable facility;
(2) Proof of the applicant’s past
competent and safe performance; and
(3) Proof of the applicant’s current
qualifications.
(g) Proficiency. The facility licensee
must develop, implement, and maintain
a proficiency program to ensure that
operators and senior operators will
actively perform the functions of an
operator or senior operator, respectively,
as needed to maintain proficiency with
on-shift duties and familiarity with
plant status. This program must include
those steps that will be taken to reestablish proficiency when it cannot be
maintained. This program must be
approved by the Commission as part of
its approval of the OL or COL for the
plant. The approved proficiency
program is subject to the requirements
of § 53.1565.
(h) Records. Each record required by
this section must be legible throughout
the retention period specified by each
Commission regulation. The record may
be the original, a reproduced copy, or an
electronic copy provided that the copy
is authenticated by authorized
personnel.
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§ 53.785 Conditions of operator and senior
operator licenses.
Each operator and senior operator
license contains and is subject to the
following conditions whether stated in
the license or not:
(a) Neither the license nor any right
under the license may be assigned or
otherwise transferred.
(b) The license is limited to the
facility for which it is issued.
(c) The license is limited to those
controls of the facility or facilities
specified in the license.
(d) The license is subject to, and the
licensee must observe, all applicable
rules, regulations, and orders of the
Commission.
(e) The licensee must maintain or reestablish proficiency in accordance with
the facility licensee’s Commissionapproved proficiency program required
under § 53.780(g).
(f) The licensee must be subject to the
facility’s Commission-approved
operator licensing requalification and
requalification examination programs
required under § 53.780(c).
(g) The licensee must have a biennial
medical examination as described by
§ 53.765.
(h) The licensee must notify the
Commission within 30 days about a
conviction for a felony.
(i) The licensee must not consume or
ingest alcoholic beverages within the
protected area of commercial nuclear
plants. The licensee must not use,
possess, or sell any illegal drugs. The
licensee must not perform activities
authorized by a license issued under
this part while under the influence of
alcohol or any prescription, over-thecounter, or illegal substance that could
adversely affect his or her ability to
safely and competently perform his or
her licensed duties. For the purpose of
this paragraph, with respect to alcoholic
beverages and drugs, the term ‘‘under
the influence’’ means the licensee
exceeded, as evidenced by a confirmed
test result, the lower of the cutoff levels
for drugs or alcohol contained in 10 CFR
part 26, or as established by the facility
licensee. The term ‘‘under the
influence’’ also means the licensee
could be mentally or physically
impaired as a result of substance use
including prescription and over-thecounter drugs, as determined under the
provisions, policies, and procedures
established by the facility licensee for
its fitness-for-duty program, in such a
manner as to adversely affect his or her
ability to safely and competently
perform licensed duties.
(j) Each licensee must participate in
the drug and alcohol testing programs as
required under 10 CFR part 26.
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(k) The licensee must comply with
any other conditions that the
Commission may impose to protect
health or to minimize danger to life or
property.
§ 53.790 Issuance, modification, and
revocation of operator and senior operator
licenses.
(a) Issuance of operator and senior
operator licenses. If the Commission
determines that an applicant for an
operator license or a senior operator
license demonstrates compliance with
the requirements of the Atomic Energy
Act of 1954, as amended, (the Act) and
its regulations, it will issue a license in
the form and containing any conditions
and limitations it considers appropriate
and necessary.
(b) Modification and revocation of
operator and senior operator licenses.
(1) The terms and conditions of all
operator and senior operator licenses are
subject to amendment, revision, or
modification by reason of rules,
regulations, or orders issued in
accordance with the Act or any
amendments thereto.
(2) Any license may be revoked,
suspended, or modified, in whole or in
part—
(i) For any material false statement in
the application or in any statement of
fact required under section 182 of the
Act;
(ii) Because of conditions revealed by
the application or statement of fact or
any report, record, inspection, or other
means that would warrant the
Commission to refuse to grant a license
on an original application;
(iii) For willful violation of, or failure
to observe, any of the terms and
conditions of the Act or the license, or
of any rule, regulation, or order of the
Commission;
(iv) For any conduct determined by
the Commission to be a hazard to safe
operation of the facility; or
(v) For the sale, use, or possession of
illegal drugs, or refusal to participate in
the facility drug and alcohol testing
program, or a confirmed positive test for
drugs, drug metabolites, or alcohol in
violation of the conditions and cutoff
levels established by § 53.785(i) or the
consumption of alcoholic beverages
within the protected area of commercial
nuclear plants, or a determination of
unfitness for scheduled work as a result
of the consumption of alcoholic
beverages.
§ 53.795 Expiration and renewal of
operator and senior operator licenses.
(a) Expiration. (1) Each operator
license and senior operator license
expires 6 years after the date of
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issuance, upon termination of
employment with the facility licensee,
or upon determination by the facility
licensee that the licensed individual no
longer needs to maintain a license.
(2) If a licensee files an application for
renewal or an upgrade of an existing
license on Form NRC 398 at least 30
days before the expiration of the
existing license, it does not expire until
disposition of the application for
renewal or for an upgraded license has
been finally determined by the
Commission. Filing by mail will be
deemed to be complete at the time the
application is postmarked
(b) Renewal. (1) The applicant for
renewal of an operator license or senior
operator license must—
(i) Complete and sign Form NRC 398
and include the number of the license
for which renewal is sought.
(ii) File an original of NRC Form 398
as specified in § 53.775.
(iii) Provide written evidence of the
applicant’s experience under the
existing license and the approximate
number of hours that the licensee has
operated the facility.
(iv) Provide a statement by an
authorized representative of the facility
licensee that during the effective term of
the current license the applicant has
satisfactorily completed the
requalification program for the facility
for which operator or senior operator
license renewal is sought.
(v) Provide evidence that the
applicant has discharged the license
responsibilities competently and safely.
The Commission may accept as
evidence of the applicant’s having met
this requirement a certificate of an
authorized representative of the facility
licensee or holder of an authorization by
which the licensee has been employed.
(vi) Provide certification by the
facility licensee of medical condition
and general health on Form NRC 396, to
comply with § 53.765.
(2) The license will be renewed if the
Commission finds that—
(i) The medical condition and the
general health of the licensee continue
to be such as not to cause operational
errors that endanger public health and
safety. The Commission will base this
finding upon the certification by the
facility licensee as described in
§ 53.765(b).
(ii) The licensee—
(A) Is capable of continuing to
competently and safely assume licensed
duties;
(B) Has successfully completed a
requalification program that has been
approved by the Commission as
required by § 53.780(c); and
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(C) Has passed the requalification
examinations as required by § 53.780(c).
(iii) There is a continued need for an
operator to operate or for a senior
operator to supervise operators at the
facility designated in the application.
(iv) The past performance of the
licensee has been satisfactory to the
Commission. In making its finding, the
Commission will include in its
evaluation information such as notices
of violations or letters of reprimand in
the licensee’s docket.
§ 53.800 Facility licensees for self-reliantmitigation facilities.
(a) A commercial nuclear plant is a
self-reliant-mitigation facility if the NRC
determined as part of its approval of the
OL or COL for that plant that its design
demonstrates compliance with criteria
(a)(1) though (a)(5) of this section. A
self-reliant-mitigation facility is of a
class, based upon the similarity of
operating and technical characteristics
of the plants in the class, such that its
licensee must comply with the
requirements of §§ 53.800 through
53.820 in lieu of those in §§ 53.760
through 53.795.
(1) The safety performance criteria of
§§ 53.210 and 53.220 and, if applicable,
any alternative criteria used in
accordance with § 53.470, must be met
without reliance upon human action for
credited event mitigation.
(2) The results of a probabilistic risk
analysis must demonstrate that the
evaluation criteria for the events
analyzed in accordance with § 53.450
will be met without reliance on human
actions to achieve acceptable event
mitigation.
(3) The functional requirements
analysis and function allocation
performed under § 53.730(d) must
demonstrate that functions required for
safety are not reliant upon credited
human action.
(4) The plant response to events
analyzed under § 53.450 must rely
exclusively on safety features and
characteristics that will neither be
rendered unavailable by credible human
errors of commission or omission nor
credibly require manual human
operation in response to equipment
failures. Compliance with this
paragraph may be achieved through the
use of SSCs that function through
inherent characteristics or that have
engineered protections against human
failures.
(5) The plant design must provide for
a layered defense-in-depth approach
that is not dependent upon any single
barrier or credited human action.
(b) [Reserved]
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§ 53.805 Facility licensee requirements
related to generally licensed reactor
operators.
(a) Licensees for self-reliantmitigation facilities that have not yet
certified the permanent cessation of
operations and permanent removal of
fuel from the reactor vessel as described
under § 53.1070 must demonstrate
compliance with the following
requirements:
(1) Ensure that, in addition to being
qualified to perform those items
identified by the facility-specific
systems approach to training conducted
under § 53.815, generally licensed
reactor operators are qualified to safely
and competently—
(i) Perform administrative tasks,
including compliance with technical
specifications, and perform operability
determinations;
(ii) Implement maintenance and
configuration controls;
(iii) Comply with radioactive release
limitations;
(iv) Understand plant operating data,
including reactor parameters, and
evaluate emergency conditions;
(v) Initiate a reactor shutdown from
necessary locations;
(vi) Dispatch and direct operations
and maintenance personnel;
(vii) Implement any applicable
responsibilities under the facility
emergency plan; and
(viii) Make required notifications to
local, State, participating Tribal and
Federal authorities.
(2) Develop, implement, and maintain
facility technical specifications that
provide the necessary administrative
controls to ensure the implementation
of these requirements.
(3) Develop, implement, and maintain
the generally licensed reactor operator
training, examination, and proficiency
programs required under § 53.815.
(4) Ensure that generally licensed
reactor operators are subject to the
facility’s generally licensed reactor
operator training, examination, and
proficiency programs required under
§ 53.815. Ensure that generally licensed
reactor operators are subject to and
comply with the applicable
programmatic requirements for plant
personnel required under 10 CFR parts
26 and 73. An individual that is not in
compliance with any of these programs
is not qualified to be in a position that
may involve the manipulation of the
controls of the commercial nuclear
plant.
(5) Report annually to the NRC the
identity of all generally licensed reactor
operators at the commercial nuclear
plant, including all additions and
deletions since the previous report.
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(6) Ensure that the facility design
continues to meet the criteria of
§ 53.800.
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§ 53.810 Generally licensed reactor
operators.
(a) A general license to manipulate
the controls of a self-reliant-mitigation
facility and to direct the licensed
activities of generally licensed reactor
operators is hereby issued to any
individual employed in a position that
may involve the manipulation of the
controls of that self-reliant-mitigation
facility and who observes the
restrictions of this section.
(b) A generally licensed reactor
operator must comply with the
operating procedures and other
conditions specified in the license
authorizing operation of the facility.
(c) The general license is limited to
the facility or facilities at which the
operator is employed.
(d) The Commission will suspend the
general license on an individual basis
for violations of any provision of the Act
or any rule or regulation issued
thereunder whenever the Commission
deems such suspension desirable,
including—
(1) For willful violation of, or failure
to observe, any of the terms and
conditions of the Act or the general
license, or of any rule, regulation, or
order of the Commission;
(2) For any conduct determined by the
Commission to be a hazard to safe
operation of the facility; or
(3) For the sale, use, or possession of
illegal drugs, or refusal to participate in
the facility drug and alcohol testing
program, or a confirmed positive test for
drugs, drug metabolites, or alcohol in
violation of the conditions and cutoff
levels established by § 53.810(f) or the
consumption of alcoholic beverages
within the protected area of commercial
nuclear plants, or a determination of
unfitness for scheduled work as a result
of the consumption of alcoholic
beverages.
(e) The Commission may require
information from a generally licensed
reactor operator to determine whether a
general license should be revoked or
suspended with respect to that operator.
(f) The generally licensed reactor
operator must not consume or ingest
alcoholic beverages within the protected
area of commercial nuclear plants. The
generally licensed reactor operator must
not use, possess, or sell any illegal
drugs. The generally licensed reactor
operator must not perform activities
requiring a general license while under
the influence of alcohol or any
prescription, over-the-counter, or illegal
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substance that could adversely affect his
or her ability to safely and competently
perform these activities. For the purpose
of this paragraph, with respect to
alcoholic beverages and drugs, the term
‘‘under the influence’’ means the
generally licensed reactor operator
exceeded, as evidenced by a confirmed
test result, the lower of the cutoff levels
for drugs or alcohol contained in 10 CFR
part 26, or as established by the facility
licensee. The term ‘‘under the
influence’’ also means the generally
licensed reactor operator could be
mentally or physically impaired as a
result of substance use including
prescription and over-the-counter drugs,
as determined under the provisions,
policies, and procedures established by
the facility licensee for its fitness-forduty program, in such a manner as to
adversely affect his or her ability to
safely and competently perform
generally licensed reactor operator
duties.
(g) The generally licensed reactor
operator must notify the Commission
within 30 days about a conviction for a
felony.
§ 53.815 Generally licensed reactor
operator training, examination, and
proficiency programs.
(a) Applicability. The requirements of
this section apply to each licensee of a
self-reliant-mitigation facility that has
not yet certified the permanent
cessation of operations and permanent
removal of fuel from the reactor vessel
as described under § 53.1070.
(b) Requirements. (1) The licensee
must develop, implement, and maintain
training and examination programs that
demonstrate compliance with the
requirements of paragraphs (b)(2) and
(3) of this section.
(2) The training program must provide
for both the initial and continuing
training of generally licensed reactor
operators and be derived from a systems
approach to training as defined in this
part.
(3)(i) The training program must
incorporate the instructional
requirements necessary to provide
qualified generally licensed reactor
operators to operate and maintain the
facility in a safe manner in all modes of
operation. The training program must
comply with the facility license,
including all technical specifications
and applicable regulations. The facility
licensee must periodically evaluate and
revise the training program as
appropriate to reflect industry
experience and relevant changes,
including changes to the facility,
procedures, regulations, and quality
assurance (QA) requirements. Facility
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licensee management must periodically
review the training program for
effectiveness.
(ii) The training program must ensure
that generally licensed reactor operators
have and maintain the necessary
knowledge, skills, and abilities.
(iii) The training program must
include the generally licensed reactor
operator manipulating the controls of
either the facility or a simulation facility
that demonstrates compliance with the
requirements of § 53.815(e).
(iv) The training program must
include an initial examination program
for testing a representative sample of the
knowledge, skills, and abilities needed
to safely perform generally licensed
reactor operator duties, to include both
the examination methods and criteria to
be used to assess passing performance.
The facility licensee must provide the
opportunity for a representative of the
Commission to be present during initial
examination administration.
(v) The training program must include
a requalification examination program
for testing a sample of the topics
included under the systems approach to
training, to include the examination
methods and criteria to be used to assess
passing performance. The
requalification examination program
must specify an appropriate periodicity
for administering a complete
requalification examination to each
generally licensed reactor operator, and
the facility licensee must provide the
opportunity for a representative of the
Commission to be present during
requalification examination
administration.
(A) The facility licensee must ensure
that any generally licensed reactor
operator who either demonstrates
unsatisfactory performance on, or fails
to complete, the requalification
examination is removed from the
performance of generally licensed
reactor operator duties until such time
that any necessary remedial training has
been completed and a retake
examination has been passed.
(B) [Reserved]
(vi) The training program must be
approved by the Commission prior to its
use, as described under § 53.730(g). The
examination program must provide for
valid and reliable examinations and
must be approved by the Commission
prior to their use, as described under
§ 53.730(g). The approved programs are
subject to the requirements of § 53.1565.
(c) Records. The following is required
regarding the documentation of the
generally licensed reactor operator
training and examination programs:
(1) Sufficient records must be
maintained by the facility licensee to
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maintain the integrity of the programs
and kept available for NRC inspection to
verify the adequacy of the programs.
(2) The facility licensee must
maintain records documenting the
participation of each generally licensed
reactor operator in the training and
examination programs. The records
must contain copies of examinations
administered, the answers given by the
generally licensed reactor operator,
documentation of the grading of
examinations, and documentation of
any additional training administered in
areas in which a generally licensed
reactor operator exhibited deficiencies.
The facility licensee must retain these
records while the associated generally
licensed reactor operators remain
employed at the facility.
(3) Each record required by this part
must be legible throughout the retention
period. The record may be the original,
a reproduced copy, or an electronic
copy provided that the copy is
authenticated by authorized personnel.
(d) Examination integrity. Generally
licensed reactor operators and facility
licensees must not engage in any
activity that compromises the integrity
of any examination conducted under the
generally licensed reactor operator
training and examination programs. The
integrity of an examination is
considered compromised if any activity,
regardless of intent, affected, or, but for
detection, could have affected the
equitable and consistent administration
of the examination. This includes all
activities related to the preparation,
administration, and grading of
examinations.
(e) Simulation facilities. (1)
Simulation facilities used for training
purposes, for maintaining proficiency,
or for the conduct of examinations must
demonstrate compliance with the
following criteria as they relate to the
facility licensee’s reference plant:
(i) The simulation facility must be of
sufficient scope and fidelity for
individuals to acquire and demonstrate
the necessary knowledge, skills, and
abilities to safely perform generally
licensed reactor operator duties.
(ii) The simulation facility must
utilize models relating to nuclear,
thermal-hydraulic, and other applicable
design-specific characteristics that
either replicate the most recent fuel load
in the reference commercial nuclear
plant or, prior to initial fuel load (or, for
a fueled manufactured reactor, prior to
initiating the physical removal of any
one of the independent physical
mechanisms to prevent criticality
required under § 53.620(d)(1)), replicate
the intended initial fuel load for the
reference commercial nuclear plant,
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with the exception of those portions of
the simulation facility that utilize the
reference plant itself.
(iii) Simulator fidelity must be
demonstrated so that significant control
manipulations are completed without
procedural exceptions, simulator
performance exceptions, or deviation
from the approved training scenario
sequence.
(2) Facility licensees that maintain a
simulation facility for training purposes,
for maintaining proficiency, or for the
conduct of examinations must—
(i) Conduct performance testing
throughout the life of the simulation
facility in a manner sufficient to ensure
that paragraph (e)(1) of this section is
met;
(ii) Retain the results of performance
testing for 4 years after the completion
of each performance test or until
superseded by updated test results;
(iii) Promptly correct modeling and
hardware discrepancies and
discrepancies identified from scenario
validation and from performance testing
or provide justification for why the
presence of such discrepancies will not
adversely affect the criteria of paragraph
(e)(1) of this section;
(iv) Make the results of any
uncorrected performance test failures
that may exist at the time of an
inspection available for NRC review;
and
(v) Maintain the provisions for
examination integrity consistent with
§ 53.815(d).
(f) Waiver of examination
requirement. The facility licensee may
waive any or all of the requirements for
an examination in accordance with the
facility licensee’s Commission-approved
generally licensed reactor operator
training and examination programs.
(g) Proficiency. The facility licensee
must develop, implement, and maintain
a proficiency program to allow generally
licensed reactor operators to maintain
proficiency regarding position functions
and familiarity with plant status. This
program must include those steps that
will be taken in order to re-establish
proficiency when it cannot be
maintained.
§ 53.820 Cessation of individual
applicability.
The general license ceases to be
applicable on an individual basis once
a generally licensed reactor operator is
no longer being employed in a position
that may involve the manipulation of
the controls of the self-reliant mitigation
facility.
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§ 53.830 Training and qualification of
commercial nuclear plant personnel.
(a) This section addresses personnel
training requirements. The regulations
within this section are applicable to all
applicants for or holders of OLs or COLs
under this part.
(b) Prior to initial fuel load (or, for a
fueled manufactured reactor, prior to
initiating the physical removal of any
one of the independent physical
mechanisms to prevent criticality
required under § 53.620(d)(1)), each
holder of an operating or COL under
this part must, with sufficient time to
provide trained and qualified personnel
to operate the facility, establish,
implement, and maintain a training
program that demonstrates compliance
with the requirements of paragraphs (c)
and (d) of this section.
(c) The training program must be
derived from a systems approach to
training as defined in this part and must
provide, at a minimum, for the training
and qualification of the following
categories of commercial nuclear plant
personnel:
(1) Supervisors (e.g., shift
supervisors);
(2) Technicians (e.g., maintenance,
chemistry, and radiological); and
(3) Other appropriate operating
personnel (e.g., auxiliary operators,
certified fuel handlers, and individuals
who provide engineering expertise to
on-shift operating personnel).
(d) The training program must
incorporate the instructional
requirements necessary to provide
qualified personnel to operate
components of a commercial nuclear
plant and maintain the facility in a safe
manner in all modes of operation. The
training program must be developed to
be in compliance with the facility
license, including all technical
specifications and applicable
regulations.
(1) The training program must be
periodically evaluated and revised as
appropriate to reflect industry
experience and relevant changes,
including changes to the facility,
procedures, regulations, and QA
requirements. The training program
must be periodically reviewed by
facility licensee management for
effectiveness.
(2) Sufficient records must be
maintained by the facility licensee to
maintain program integrity and kept
available for NRC inspection to verify
the adequacy of the training program.
§ 53.845
Programs.
(a) The required plant programs under
this part must include but are not
necessarily limited to the programs
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described in the following sections of
this subpart. Licensees may combine,
separate, and otherwise organize
programs and related documents as
appropriate for the technologies and
organizations associated with the
commercial nuclear plant.
(b) In addition to the programs
described in the following sections,
programs must be provided for each
commercial nuclear plant, if necessary,
to ensure that the performance of design
features and human actions are
consistent with the analyses performed
under §§ 53.450 and 53.730 and that the
plant will demonstrate compliance with
the safety criteria defined in §§ 53.210
and 53.220.
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§ 53.850
Radiation protection.
(a) Each holder of an OL or COL
under this part must develop,
implement, and maintain a Radiation
Protection Program for operations that is
commensurate with the scope and
extent of licensed activities under this
part and includes measures for limiting
and monitoring radioactive plant
effluents and limiting and monitoring
the dose to individuals working with
radioactive materials in accordance with
10 CFR part 20.
(b) Each holder of an OL or COL
under this part must develop,
implement, and maintain a program for
the control of radioactive effluents and
for keeping the doses to members of the
public from radioactive effluents as low
as is reasonably achievable and for
environmental monitoring. The program
must be contained in an Offsite Dose
Calculations Manual, must be
implemented by procedures, and must
include remedial actions to be taken
whenever the program limits are
exceeded. The Offsite Dose Calculations
Manual must—
(1) Contain the methodology and
parameters used in the calculation of
offsite doses resulting from radioactive
gaseous and liquid effluents, in the
calculation of gaseous and liquid
effluent monitoring alarm and trip
setpoints, and in the conduct of the
radiological environmental monitoring
program; and
(2) Contain the radioactive effluent
controls and radiological environmental
monitoring activities, and descriptions
of the information that should be
included in the Annual Radiological
Environmental Operating and
Radioactive Effluent Release Reports
required by § 53.1645.
(c) Each holder of an OL or COL
under this part must develop,
implement, and maintain a Process
Control Program that identifies the
administrative and operational controls
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for solid radioactive waste processing,
process parameters, and surveillance
requirements sufficient to ensure
compliance with the requirements of 10
CFR part 20, 10 CFR part 61, and 10
CFR part 71.
§ 53.855
Emergency preparedness.
(a) Each holder of an OL or COL
under this part must have an emergency
response plan that must contain
information needed to demonstrate
compliance with either the
requirements in § 50.160 of this chapter
or the requirements in appendix E to
part 50 and the planning standards of
§ 50.47(b) of this chapter.
(b) No initial OL, initial COL, or early
site permit that includes complete and
integrated emergency plans will be
issued under this part unless a finding
is made by the NRC, in accordance with
§ 50.47 of this chapter, that there is
reasonable assurance that adequate
protective measures can and will be
taken in the event of a radiological
emergency.
§ 53.860
Security programs.
(a) Physical Protection Program. Each
holder of an OL or COL under this part
must develop, implement, and maintain
a physical protection program under the
following requirements:
(1) The licensee must implement
security requirements for the protection
of special nuclear material based on the
type, enrichment, and quantity in
accordance with 10 CFR part 73, as
applicable, and implement security
requirements for the protection of
Category 1 and Category 2 quantities of
radioactive material in accordance with
10 CFR part 37, as applicable; and
(2) The licensee must demonstrate
compliance with the provisions set forth
in either §§ 73.55 or 73.100 of this
chapter, unless the licensee
demonstrates compliance with the
following criterion:
(i) The radiological consequences
from a design-basis threat-initiated
event involving the loss of engineered
systems for decay heat removal and
possible breaches in physical structures
surrounding the reactor, spent fuel, and
other inventories of radioactive
materials result in offsite doses below
the values in § 53.210.
(ii) The applicant must perform a sitespecific analysis, including
identification of target sets, to
demonstrate that the criterion in
§ 53.860(a)(2)(i) is satisfied. The analysis
must assume that licensee mitigation
and recovery actions, including any
operator actions, are unavailable or
ineffective. The licensee must maintain
the analysis until the permanent
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cessation of operations and permanent
removal of fuel from the reactor vessel
as described under § 53.1070.
(b) Fitness for Duty. Each holder of an
OL or COL under this part must
develop, implement, and maintain a
fitness for duty program under 10 CFR
part 26.
(c) Access Authorization. Each holder
of an OL or COL under this part must
develop, implement, and maintain an
access authorization program under
§ 73.120 of this chapter if the criterion
in § 53.860(a)(2)(i) is satisfied, or the
requirements in § 73.56 of this chapter
if the criterion is not satisfied.
(d) Cybersecurity. Each holder of an
OL or COL under this part must
develop, implement, and maintain a
cybersecurity program under §§ 73.54 or
73.110 of this chapter.
(e) Information Security. Each holder
of an OL or COL under this part must
develop, implement, and maintain an
information protection system under
§§ 73.21, 73.22, and 73.23 of this
chapter, as applicable.
§ 53.865
Quality assurance.
Each holder of an OL or COL under
this part must develop, implement, and
maintain a quality assurance program in
accordance with appendix B of part 50
of this chapter. A written quality
assurance program manual must be
developed and used to guide the
conduct of the program in accordance
with generally accepted consensus
codes and standards that have been
endorsed or otherwise found acceptable
by the NRC.
§ 53.870
Integrity assessment programs.
Each holder of an OL or COL under
this part must develop, implement, and
maintain an integrity assessment
program to monitor, evaluate, and
manage—
(a) The effects of plant aging on SR
and NSRSS SSCs. The program may
refer to surveillances, tests, and
inspections conducted for specific SSCs
in accordance with other requirements
in this part or conducted in accordance
with applicable consensus codes and
standards endorsed or otherwise found
acceptable by the NRC;
(b) Cyclic or transient load limits to
ensure that SR and NSRSS SSCs are
maintained within the applicable design
limits; and
(c) Degradation mechanisms related to
chemical interactions, operating
temperatures, effects of irradiation, and
other environmental factors to ensure
that the capabilities, availability, and
reliability of SR and NSRSS SSCs
demonstrate compliance with the
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§ 53.880 Inservice inspection and
inservice testing.
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§ 53.875
Fire protection.
(a)(1) Each holder of an OL or COL
under this part must have a fire
protection plan that describes the
overall fire protection program for the
facility; identifies the various positions
within the licensee’s organization that
are responsible for the program; states
the authorities that are delegated to each
of these positions to implement those
responsibilities; and outlines the plans
for fire protection, fire detection and
suppression capability; and limitation of
fire damage.
(2) The fire protection plan must also
describe specific features necessary to
implement the program described in
paragraph (a)(1) of this section such as
the following: administrative controls
and personnel requirements for fire
prevention and manual fire suppression
activities; automatic and manually
operated fire detection and suppression
systems; and the means to limit fire
damage to SSCs so that the capability to
demonstrate compliance with the
requirements of § 53.210 is ensured.
(b)(1) Each holder of an OL or COL
under this part must develop a
performance-based or deterministic fire
protection program that demonstrates
compliance with the safety criteria
outlined in §§ 53.210 and 53.220,
related safety functions outlined in
§ 53.230, and defense in depth as
outlined in § 53.250 with specific fire
protection measures related to fire
prevention, fire detection, and fire
suppression.
(2) The fire protection program must
comply with the following:
(i) Safety-related and NSRSS SSCs
must be designed, located, and
maintained to minimize, consistent with
other safety requirements, the
probability and effect of fires and
explosions.
(ii) Noncombustible and fire-resistant
materials must be used wherever
practical throughout the facility,
particularly in locations with SR and
NSRSS SSCs.
(iii) Fire detection and fire
suppression systems of appropriate
capacity and capability must be
provided and designed and maintained
to minimize the adverse effects of fires
on SR and NSRSS SSCs.
(iv) Fire suppression systems must be
designed and maintained to ensure that
their rupture or inadvertent operation
does not significantly impair the ability
of SR and NSRSS SSCs to perform their
safety functions to satisfy § 53.230.
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(a) Each holder of an OL or COL
under this part must develop,
implement, and maintain a program for
inservice inspection (ISI) and inservice
testing (IST) prior to receiving an OL or
COL. The ISI/IST programs must,
wherever applicable, be in accordance
with generally accepted consensus
codes and standards that have been
endorsed or otherwise found acceptable
by the NRC. The ISI/IST program must
include all inspections and tests
required by the codes and standards
used in the design and be supplemented
by risk insights that identify the most
important SSCs to plant safety. The
types of testing and inspections and
their frequency should be informed by
risk insights to maintain the reliability
and performance of SSCs consistent
with the associated design and analyses
activities involving those SSCs. Risk
insights must also be used to determine
when to conduct the inspections and
tests (e.g., full power, shutdown,
refueling) to minimize risk to the plant
workers and the public. The ISI/IST
program must be documented in a
written manual and managed by
qualified personnel reporting to the
Plant Manager.
(b) Prior to plant operation, baseline
inspections and testing must be
performed using the same techniques as
will be used for future inspections and
testing. The results of these inspections
and testing must be used as benchmarks
for evaluating the results of future
inspections and testing. Sufficient room
and support must be provided to
accommodate the personnel, ISI/IST
equipment, and shielding necessary to
perform the inspections and testing.
Acceptance criteria for determining
whether corrective action is needed
must be developed (or taken from the
codes and standards used in the design)
for evaluating the results of the
inspections and testing. The results of
the inspections and testing must be
provided to the Plant Manager who is
responsible for determining what, if
any, corrective action is needed and
when it should be taken. The ISI/IST
results and corrective actions must be
documented and the documentation
retained for the life of the plant.
§ 53.910
Procedures and guidelines.
(a) Each holder of an OL or COL
under this part must have a program for
developing, implementing, and
maintaining an integrated set of
procedures, guidelines, and related
supporting activities to support normal
operations and respond to possible
unplanned events.
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(b) The program required by
paragraph (a) of this section must
include but is not limited to
development, implementation,
maintenance, and supporting activities
of procedures and guidelines for the
following:
(1) Plant operations;
(2) Maintenance activities under
§ 53.715;
(3) Program requirements under this
subpart F of this part;
(4) Emergency operating procedures,
if developed to address the role of
human actions in responding to LBEs;
(5) Accident management guidelines,
if developed to address the role of
human actions in responding to LBEs;
(6) Procedures for each area in which
licensed special nuclear material is
handled, used, or stored to protect
personnel upon the sounding of a
criticality alarm required by
§ 53.440(m); and
(7) Procedures that describe how the
licensee will address the following areas
if the licensee is notified of a potential
aircraft threat:
(i) Verification of the authenticity of
threat notifications;
(ii) Maintenance of continuous
communication with threat notification
sources;
(iii) Contacting all onsite personnel
and applicable offsite response
organizations;
(iv) Onsite actions necessary to
enhance the capability of the facility to
mitigate the consequences of an aircraft
impact;
(v) Measures to reduce visual
discrimination of the site relative to its
surroundings or individual buildings
within the protected area;
(vi) Dispersal of equipment and
personnel, as well as rapid entry into
site protected areas for essential onsite
personnel and offsite responders who
are necessary to mitigate the event; and
(vii) Recall of site personnel.
Subpart G—Decommissioning
Requirements
§ 53.1000
Scope and purpose.
This subpart defines the requirements
related to decommissioning for
applicants for, or holders of, an
operating license (OL) or combined
license (COL). The requirements related
to maintaining financial assurance for
decommissioning are in §§ 53.1010
through 53.1060. The requirements for
transitioning from operations to
decommissioning and for the release of
property and termination of the license
are in §§ 53.1070 through 53.1080.
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§ 53.1010 Financial assurance for
decommissioning.
(a) This section establishes
requirements for indicating to the U.S.
Nuclear Regulatory Commission (NRC)
how an applicant for or holder of an OL
or COL under this part will provide
reasonable assurance that funds will be
available for the decommissioning
process. Reasonable assurance consists
of a series of steps as provided in
paragraph (b) of this section and
§§ 53.1020, 53.1030 and 53.1040.
Funding for the decommissioning of
commercial nuclear plants may also be
subject to the regulation of Federal or
State government agencies (e.g., Federal
Energy Regulatory Commission (FERC)
and State Public Utility Commissions)
that have jurisdiction over rate
regulation. The requirements of this
subpart, in particular § 53.1020, are in
addition to, and not a substitution for,
other requirements, and are not
intended to be used by themselves or by
other agencies to establish rates.
(b) Each applicant for an OL or COL
under this part must prepare a plan and
an associated decommissioning report
that ensures and documents that
adequate funding will be available to
decommission the facility. Each holder
of an OL or COL must implement and
maintain the plan.
(1)(i) Before the Commission issues an
OL under this part, the applicant must
update the decommissioning report to
certify that it has provided financial
assurance for decommissioning in the
amount proposed in the application and
approved by the NRC under § 53.1020.
(ii) No later than 30 days after the
Commission issues the notice of
intended operation under § 53.1452 for
a COL under this part, the licensee must
update the decommissioning report to
certify that it has provided financial
assurance for decommissioning in the
amount proposed in the application and
approved by the NRC under § 53.1020.
(2) The amount of financial assurance
for decommissioning to be provided
must be based on a site-specific cost
estimate for decommissioning the
facility under § 53.1020.
(3) The amount of financial assurance
for decommissioning to be provided
must be adjusted annually using a rate
at least equal to that stated in § 53.1030.
(4) The amount of financial assurance
for decommissioning to be provided
must be covered by one or more of the
methods described in § 53.1040 as
acceptable to the NRC. A copy of the
financial instrument obtained to satisfy
the requirements of § 53.1040 must be
submitted to the NRC as part of the
application for an OL under this part;
however, an applicant for or holder of
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a COL need not obtain such financial
instrument or submit a copy to the
Commission except as provided in
§ 53.1060(b).
§ 53.1040 Methods for providing financial
assurance for decommissioning.
Financial assurance for
decommissioning is to be provided by
the following methods.
§ 53.1020 Cost estimates for
(a) Prepayment. Prepayment is the
decommissioning.
deposit made preceding the start of
operation or the transfer of a license
Cost estimates for decommissioning
under § 53.1570 into an account
must be site-specific. Site-specific
segregated from licensee assets and
decommissioning cost estimates (DCEs)
must account for the engineering, labor, outside the administrative control of the
equipment, transportation, disposal, and licensee and its subsidiaries or affiliates
of cash or liquid assets such that the
related charges needed to support
amount of funds would be sufficient to
termination of the license. They must
pay decommissioning costs. Prepayment
include the costs for decontaminating
may be in the form of a trust, escrow
structures, systems, and components
account, or Government fund with
and the site environs; removal of
contaminated components and materials payment by certificate of deposit,
deposit of government or other
from the plant and the site environs;
securities, or other method acceptable to
disposal of removed components and
the NRC. This trust, escrow account,
materials in appropriate facilities; and
Government fund, or other type of
any other activities supporting the
agreement must be established in
release of the property and termination
writing and maintained at all times in
of the license. They must also address
the United States with an entity that is
the approach to annual adjustments
an appropriate State or Federal
required by § 53.1030. Finally, siteGovernment agency, or an entity whose
specific DCEs must include plans for
operations in which the prepayment
adjusting levels of funds assured for
deposit is managed are regulated and
decommissioning to demonstrate that a
examined by a Federal or State agency.
reasonable level of assurance will be
A licensee that has prepaid funds based
provided that funds will be available
on a site-specific cost estimate under
when needed to cover the cost of
§ 53.1020 may take credit for projected
decommissioning.
earnings on the prepaid
§ 53.1030 Annual adjustments to cost
decommissioning trust funds, using up
estimates for decommissioning.
to a 2 percent annual real rate of return
through the time of termination of the
Each holder of an OL or COL under
license. A licensee may use a credit of
this part must annually adjust the cost
greater than 2 percent if the licensee’s
estimate for decommissioning to
rate-setting authority has specifically
account for escalation in labor, energy,
authorized a higher rate. Actual
and waste burial costs. Licensees may
earnings on existing funds may be used
elect to use either a site-specific
to calculate future fund needs.
adjustment factor, approved as part of
(b) External sinking fund. An external
the plan and associated
sinking fund is a fund established and
decommissioning report required by
maintained by setting funds aside
§ 53.1010, in paragraph (a) of this
periodically in an account segregated
section or the generic adjustment factor
from licensee assets and outside the
in paragraph (b) of this section.
administrative control of the licensee
(a) A site-specific adjustment factor
and its subsidiaries or affiliates in
must address the estimated
which the total amount of funds would
contributions and escalation of costs for
be sufficient to pay decommissioning
the following aspects of
costs. An external sinking fund may be
decommissioning:
in the form of a trust, escrow account,
(1) Labor, materials, and services;
or Government fund, with payment by
(2) Energy and waste transportation;
certificate of deposit, deposit of
and
Government or other securities, or other
(3) Radioactive waste burial or other
method acceptable to the NRC. This
disposition.
trust, escrow account, Government
(b) A generic adjustment factor must
fund, or other type of agreement must be
be at least equal to 0.65 L + 0.13 E + 0.22 established in writing and maintained at
B, where L and E are escalation factors
all times in the United States with an
for labor and energy, respectively, and
entity that is an appropriate State or
are to be taken from regional data of
Federal Government agency, or an entity
U.S. Department of Labor Bureau of
whose operations in which the external
Labor Statistics and B is an escalation
sinking fund is managed are regulated
factor for waste burial and is to be taken and examined by a Federal or State
from NRC report NUREG–1307, ‘‘Report agency. A licensee that has collected
on Waste Burial Charges.’’
funds based on a site-specific cost
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estimate under § 53.1020 may take
credit for projected earnings on the
external sinking funds using up to a 2
percent annual real rate of return from
the time of future funds’ collection
through the time of termination of the
license. A licensee may use a credit of
greater than 2 percent if the licensee’s
rate-setting authority has specifically
authorized a higher rate. Actual
earnings on existing funds may be used
to calculate future fund needs. A
licensee whose rates for
decommissioning costs cover only a
portion of these costs may make use of
this method only for the portion of these
costs that are collected in one of the
manners described in this paragraph.
This method may be used as the
exclusive mechanism relied upon for
providing financial assurance for
decommissioning in the following
circumstances:
(1) By a licensee that recovers, either
directly or indirectly, the estimated total
cost of decommissioning through rates
established by ‘‘cost of service’’ or
similar ratemaking regulation. Public
utility districts, municipalities, rural
electric cooperatives, and State and
Federal agencies, including associations
of any of the foregoing, that establish
their own rates and are able to recover
their cost of service allocable to
decommissioning, are deemed to satisfy
this condition.
(2) By a licensee whose source of
revenues for its external sinking fund is
a ‘‘non-bypassable charge,’’ the total
amount of which will provide funds
estimated to be needed for
decommissioning pursuant to
§§ 53.1020, 53.1060, or 53.1575.
(c) A surety method, insurance, or
other guarantee method. (1) These
methods guarantee that
decommissioning costs will be paid. A
surety method may be in the form of a
surety bond, or letter of credit. Any
surety method or insurance used to
provide financial assurance for
decommissioning must contain the
following conditions:
(i) The surety method or insurance
must be open-ended, or, if written for a
specified term, such as 5 years, must be
renewed automatically, unless 90 days
or more prior to the renewal day the
issuer notifies the NRC, the beneficiary,
and the licensee of its intention not to
renew. The surety or insurance must
also provide that the full-face amount be
paid to the beneficiary automatically
prior to the expiration without proof of
forfeiture if the licensee fails to provide
a replacement acceptable to the NRC
within 30 days after receipt of
notification of cancellation.
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(ii) The surety or insurance must be
payable to a trust established for
decommissioning costs. The trustee and
trust must be acceptable to the NRC. An
acceptable trustee includes an
appropriate State or Federal
Government agency or an entity that has
the authority to act as a trustee and
whose trust operations are regulated and
examined by a Federal or State agency.
(2) A parent company guarantee of
funds for decommissioning costs based
on a financial test may be used if the
guarantee and test are as contained in
appendix A to 10 CFR part 30.
(3) For commercial companies that
issue bonds, a guarantee of funds by the
applicant or licensee for
decommissioning costs based on a
financial test may be used if the
guarantee and test are as contained in
appendix C to 10 CFR part 30. For
commercial companies that do not issue
bonds, a guarantee of funds by the
applicant or licensee for
decommissioning costs may be used if
the guarantee and test are as contained
in appendix D to 10 CFR part 30. A
guarantee by the applicant or licensee
may not be used in any situation in
which the applicant or licensee has a
parent company holding majority
control of voting stock of the company.
(d) Funding method for Federal
licensees. For a Federal licensee, a
statement of intent containing a cost
estimate for decommissioning and
indicating that funds for
decommissioning will be obtained when
necessary.
(e) Contractual funding method.
Contractual obligation(s) on the part of
a licensee’s customer(s), the total
amount of which over the duration of
the contract(s) will provide the
licensee’s total share of uncollected
funds estimated to be needed for
decommissioning pursuant to
§§ 53.1020, 53.1060, or 53.1575. To be
acceptable to the NRC as a method of
decommissioning funding assurance,
the terms of the contract(s) must include
provisions that the buyer(s) of electricity
or other products will pay for the
decommissioning obligations specified
in the contract(s), notwithstanding the
operational status either of the licensed
plant to which the contract(s) pertains
or force majeure provisions. All
proceeds from the contract(s) for
decommissioning funding will be
deposited to the external sinking fund.
The NRC reserves the right to evaluate
the terms of any contract(s) and the
financial qualifications of the
contracting entity or entities offered as
assurance for decommissioning funding.
(f) Other funding mechanisms. Any
other mechanism, or combination of
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mechanisms, that provides, as
determined by the NRC upon its
evaluation of the specific circumstances
of each licensee submittal, assurance of
decommissioning funding equivalent to
that provided by the mechanisms
specified in paragraphs (a) through (e) of
this section. Licensees who do not have
sources of funding described in
paragraph (b) of this section may use an
external sinking fund in combination
with a guarantee mechanism, as
specified in paragraph (c) of this
section, provided that the total amount
of funds estimated to be necessary for
decommissioning is assured.
§ 53.1045 Limitations on the use of
decommissioning trust funds.
(a)(1) Decommissioning trust funds
may be used by licensees if—
(i) The withdrawals are for expenses
for decommissioning activities
consistent with the definition of
decommission or decommissioning in
§ 53.020;
(ii) The expenditure would not reduce
the value of the decommissioning trust
below an amount necessary to place and
maintain the reactor in a safe storage
condition if unforeseen conditions or
expenses arise; and
(iii) The withdrawals would not
inhibit the ability of the licensee to
complete funding of any shortfalls in
the decommissioning trust needed to
ensure the availability of funds to
ultimately release the site and terminate
the license.
(2) Initially, 3 percent of the amount
determined in accordance with
§ 53.1020 may be used for
decommissioning planning. For
licensees that have submitted the
certifications required under § 53.1070
and commencing 90 days after the NRC
has received the post-shutdown
decommissioning activities report
(PSDAR) required by § 53.1060, an
additional 20 percent may be used. An
updated site-specific DCE must be
submitted to the NRC prior to the
licensee using any funding in excess of
these amounts.
(b) Licensees that are not ‘‘electric
utilities’’ as defined in § 53.020 that use
prepayment or an external sinking fund
to provide financial assurance must
provide in the terms of the arrangements
governing the trust, escrow account, or
Government fund, used to segregate and
manage the funds that—
(1) The trustee, manager, investment
advisor, or other person directing
investment of the funds—
(i) Is prohibited from investing the
funds in securities or other obligations
of the licensee or any other owner or
operator of any commercial nuclear
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plant or their affiliates, subsidiaries,
successors or assigns, or in a mutual
fund in which at least 50 percent of the
fund is invested in the securities of a
licensee or parent company whose
subsidiary is an owner or operator of a
foreign or domestic commercial nuclear
plant. However, the funds may be
invested in securities tied to market
indices or other non-nuclear sector
collective, commingled, or mutual
funds, provided that no more than 10
percent of trust assets may be indirectly
invested in securities of any entity
owning or operating one or more
commercial nuclear plants.
(ii) Is obligated at all times to adhere
to a standard of care set forth in the
trust, which either shall be the standard
of care, whether in investing or
otherwise, required by State or Federal
law or one or more State or Federal
regulatory agencies with jurisdiction
over the trust funds, or, in the absence
of any such standard of care, whether in
investing or otherwise, that a prudent
investor would use in the same
circumstances. The term ‘‘prudent
investor,’’ shall have the same meaning
as set forth in FERC’s ‘‘Regulations
Governing Nuclear Plant
Decommissioning Trust Funds’’ at 18
CFR 35.32(a)(3), or any successor
regulation.
(2) The licensee, its affiliates, and its
subsidiaries are prohibited from being
engaged as investment manager for the
funds or from giving day-to-day
management direction of the funds’
investments or direction on individual
investments by the funds, except in the
case of passive fund management of
trust funds where management is
limited to investments tracking market
indices.
(3) The trust, escrow account,
Government fund, or other account used
to segregate and manage the funds may
not be amended in any material respect
without written notification to the
Director, Office of Nuclear Reactor
Regulation, or Director, Office of
Nuclear Material Safety and Safeguards,
as applicable, at least 30 working days
before the proposed effective date of the
amendment. The licensee must provide
the text of the proposed amendment and
a statement of the reason for the
proposed amendment. The trust, escrow
account, Government fund, or other
account may not be amended if the
person responsible for managing the
trust, escrow account, Government
fund, or other account receives written
notice of objection from the Director,
Office of Nuclear Reactor Regulation, or
Director, Office of Nuclear Material
Safety and Safeguards, as applicable,
within the notice period.
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(4) Except for withdrawals being
made under paragraph (a) of this section
or for payments of ordinary
administrative costs (including taxes)
and other incidental expenses of the
fund (including legal, accounting,
actuarial, and trustee expenses) in
connection with the operation of the
fund, no disbursement or payment may
be made from the trust, escrow account,
Government fund, or other account used
to segregate and manage the funds until
written notice of the intention to make
a disbursement or payment has been
given to the Director, Office of Nuclear
Reactor Regulation, or Director, Office of
Nuclear Material Safety and Safeguards,
as applicable, at least 30 working days
before the date of the intended
disbursement or payment. The
disbursement or payment from the trust,
escrow account, Government fund or
other account may be made following
the 30 working day notice period if the
person responsible for managing the
trust, escrow account, Government
fund, or other account does not receive
written notice of objection from the
Director, Office of Nuclear Reactor
Regulation, or Director, Office of
Nuclear Material Safety and Safeguards,
as applicable, within the notice period.
Disbursements or payments from the
trust, escrow account, Government
fund, or other account used to segregate
and manage the funds, other than for
payment of ordinary administrative
costs (including taxes) and other
incidental expenses of the fund
(including legal, accounting, actuarial,
and trustee expenses) in connection
with the operation of the fund, are
restricted to decommissioning expenses
or transfer to another financial
assurance method acceptable under
§ 53.1040 until final decommissioning
has been completed. After
decommissioning has begun and
withdrawals from the decommissioning
fund are made under paragraph (a) of
this section, no further notification need
be made to the NRC.
(c) Licensees that are ‘‘electric
utilities’’ under § 53.020 that use
prepayment or an external sinking fund
to provide financial assurance must
include a provision in the terms of the
trust, escrow account, Government
fund, or other account used to segregate
and manage funds that except for
withdrawals being made under
paragraph (a) of this section or for
payments of ordinary administrative
costs (including taxes) and other
incidental expenses of the fund
(including legal, accounting, actuarial,
and trustee expenses) in connection
with the operation of the fund, no
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disbursement or payment may be made
from the trust, escrow account,
Government fund, or other account used
to segregate and manage the funds until
written notice of the intention to make
a disbursement or payment has been
given the Director, Office of Nuclear
Reactor Regulation, or Director, Office of
Nuclear Material Safety and Safeguards,
as applicable, at least 30 working days
before the date of the intended
disbursement or payment. The
disbursement or payment from the trust,
escrow account, Government fund or
other account may be made following
the 30 working day notice period if the
person responsible for managing the
trust, escrow account, Government
fund, or other account does not receive
written notice of objection from the
Director, Office of Nuclear Reactor
Regulation, or Director, Office of
Nuclear Material Safety and Safeguards,
as applicable, within the notice period.
Disbursements or payments from the
trust, escrow account, Government
fund, or other account used to segregate
and manage the funds, other than for
payment of ordinary administrative
costs (including taxes) and other
incidental expenses of the fund
(including legal, accounting, actuarial,
and trustee expenses) in connection
with the operation of the fund, are
restricted to decommissioning expenses
or transfer to another financial
assurance method acceptable under
§ 53.1040 until final decommissioning
has been completed. After
decommissioning has begun and
withdrawals from the decommissioning
fund are made under paragraph (a) of
this section, no further notification need
be made to the NRC.
(d) A licensee that is not an ‘‘electric
utility’’ under § 53.020 and using a
surety method, insurance, or other
guarantee method to provide financial
assurance must provide that the trust
established for decommissioning costs
to which the surety or insurance is
payable contains in its terms the
requirements in § 53.1045(b)(1) through
(4).
§ 53.1050
NRC oversight.
The NRC reserves the right to take the
following steps in order to ensure a
licensee’s adequate accumulation of
decommissioning funds: review, as
needed, the rate of accumulation of
decommissioning funds and, either
independently or in cooperation with
FERC and the licensee’s State Public
Utility Commission, take additional
actions as appropriate on a case-by-case
basis, including modification of a
licensee’s schedule for the accumulation
of decommissioning funds.
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§ 53.1060 Reporting and recordkeeping
requirements.
(a) Each holder of an OL under this
part or holder of a COL under this part
after the date that the Commission has
made the finding under § 53.1452(g)
must report, at least once every 2 years,
by March 31, on the status of its
certification of decommissioning
funding for each commercial nuclear
reactor or part of a commercial nuclear
reactor that it owns. The information in
this report must include, at a minimum,
the amount of decommissioning funds
estimated to be required under
§§ 53.1020 and 53.1030; the amount of
decommissioning funds accumulated to
the end of the calendar year preceding
the date of the report; a schedule of the
annual amounts remaining to be
collected; the assumptions used
regarding rates of escalation in
decommissioning costs, rates of
earnings on decommissioning funds,
and rates of other factors used in
funding projections; any contracts upon
which the licensee is relying under
§ 53.1040(e); any modifications
occurring to a licensee’s method of
providing financial assurance since the
last submitted report; and any material
changes to trust agreements. If any of
the preceding items is not applicable,
the licensee should so state in its report.
Any licensee for a plant that is within
5 years of the projected end of its
operation, or where conditions have
changed such that it will close within 5
years (before the end of its licensed life),
or that has already closed (before the
end of its licensed life), or that is
involved in a merger or an acquisition
must submit this report annually.
(b) Each holder of a COL under this
part must, 2 years before and 1 year
before the scheduled date for initial
loading of fuel (or, for a fueled
manufactured reactor, 2 years before
and 1 year before the scheduled date for
initiating the physical removal of any
one of the independent physical
mechanisms to prevent criticality
required under § 53.620(d)(1)), submit a
report to the NRC containing a
certification updating the DCEs and a
copy of the financial instrument to be
used to satisfy § 53.1040. No later than
30 days after the Commission publishes
notice in the Federal Register under
§ 53.1452(a), the licensee must submit
an updated decommissioning report
required under § 53.1010(b)(1)(ii),
including a copy of the financial
instrument obtained to satisfy § 53.1040.
(c) Each licensee must keep records of
information important to the safe and
effective decommissioning of the facility
in an identified location until the
license is terminated by the
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Commission. If records of relevant
information are kept for other purposes,
reference to these records and their
locations may be used. Information the
Commission considers important to
decommissioning consists of—
(1) Records of spills or other unusual
occurrences involving the spread of
contamination in and around the
facility, equipment, or site. These
records may be limited to instances
when significant contamination remains
after any cleanup procedures or when
there is reasonable likelihood that
contaminants may have spread to
inaccessible areas as in the case of
possible seepage into porous materials
such as concrete. These records must
include any known information on
identification of involved nuclides,
quantities, forms, and concentrations.
(2) As-built drawings and
modifications of structures and
equipment in restricted areas where
radioactive materials are used and/or
stored and of locations of possible
inaccessible contamination such as
buried pipes that may be subject to
contamination. If required drawings are
referenced, each relevant document
need not be indexed individually. If
drawings are not available, the licensee
must substitute appropriate records of
available information concerning these
areas and locations.
(3) Records of the cost estimate
performed for the decommissioning
funding plan or of the amount certified
for decommissioning, and records of the
funding method used for assuring funds
if either a funding plan or certification
is used.
(4) Records of—
(i) The licensed site area, as originally
licensed and any revisions, which must
include a site map and any acquisition
or use of property outside the originally
licensed site area for the purpose of
receiving, possessing, or using licensed
materials;
(ii) The licensed activities carried out
on the acquired or used property; and
(iii) The release and final disposition
of any property recorded in paragraph
(c)(4)(i) of this section, the historical site
assessment performed for the release,
radiation surveys performed to support
release of the property, submittals to the
NRC made under § 53.1070, and the
methods employed to ensure that the
property met the radiological criteria of
subpart E of 10 CFR part 20 at the time
the property was released.
(d) Each holder of an OL or COL
under this part must at or about 5 years
prior to the projected end of operations
submit a preliminary DCE which
includes an up-to-date assessment of the
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major factors that could affect the cost
to decommission.
(e) Prior to or within 2 years following
permanent cessation of operations, the
licensee must submit a PSDAR to the
NRC, and a copy to the affected State(s).
The PSDAR must contain a description
of the planned decommissioning
activities along with a schedule for their
accomplishment, a discussion that
provides the reasons for concluding that
the environmental impacts associated
with site-specific decommissioning
activities will be bounded by
appropriate previously issued
environmental impact statements, and a
site-specific DCE, including the
projected cost of managing irradiated
fuel.
(f) For decommissioning activities
that delay completion of
decommissioning by including a period
of storage or surveillance, the licensee
must provide a means of adjusting cost
estimates and associated funding levels
over the storage or surveillance period.
(g) After submitting its site-specific
DCE required by paragraph (e) of this
section, and until the licensee has
completed its final radiation survey and
demonstrated that residual radioactivity
has been reduced to a level that permits
termination of its license, the licensee
must annually submit to the NRC, by
March 31, a financial assurance status
report. The report must include the
following information, current through
the end of the previous calendar year:
(1) The amount spent on
decommissioning, both cumulative and
over the previous calendar year, the
remaining balance of any
decommissioning funds, and the
amount provided by other financial
assurance methods being relied upon;
(2) An estimate of the costs to
complete decommissioning, reflecting
any difference between actual and
estimated costs for work performed
during the year, and the
decommissioning criteria upon which
the estimate is based;
(3) Any modifications occurring to a
licensee’s current method of providing
financial assurance since the last
submitted report; and
(4) Any material changes to trust
agreements or financial assurance
contracts.
(5) If the sum of the balance of any
remaining decommissioning funds, plus
earnings on such funds calculated at not
greater than a 2 percent real rate of
return, together with the amount
provided by other financial assurance
methods being relied upon, does not
cover the estimated cost to complete the
decommissioning, the financial
assurance status report must include
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additional financial assurance to cover
the estimated cost of completion.
(h) After submitting its site-specific
DCE required by paragraph (e) of this
section, the licensee must annually
submit to the NRC, by March 31, a
report on the status of its funding for
managing irradiated fuel. The report
must include the following information,
current through the end of the previous
calendar year:
(1) The amount of funds accumulated
to cover the cost of managing the
irradiated fuel;
(2) The projected cost of managing
irradiated fuel until title to the fuel and
possession of the fuel is transferred to
the Secretary of Energy; and
(3) If the funds accumulated do not
cover the projected cost, a plan to obtain
additional funds to cover the cost.
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§ 53.1070
Termination of license.
For each holder of an OL or COL
under this part—
(a)(1) When the licensee has
determined to permanently cease
operations the licensee must, within 30
days, submit a written certification to
the NRC, consistent with the
requirements of § 53.040(b)(8);
(2) When appropriate to support
decommissioning activities and the
eventual permanent removal of fuel
from the reactor vessel, the licensee
must develop defueled technical
specifications by reviewing the
operational technical specifications and
determining which specifications no
longer apply during decommissioning
and which ones should remain
applicable. The licensee must make the
appropriate submittals to the NRC in
accordance with § 53.1510 to request
changes to the technical specifications;
and
(3)(i) Once fuel has been permanently
removed from the reactor vessel, the
licensee must submit a written
certification to the NRC that meets the
requirements of § 53.040(b)(9); and
(ii) The licensee must establish and
maintain staffing consisting of certified
fuel handlers, as defined under § 53.020,
and other non-licensed personnel with
appropriate qualifications, and in
sufficient numbers, to ensure support
for facility operations and radiological
control activities, as required by the
facility defueled technical
specifications. These personnel must be
subject to the training requirements of
§ 53.830.
(b) Upon docketing of the
certifications for permanent cessation of
operations and permanent removal of
fuel from the reactor vessel, or when a
final legally effective order to
permanently cease operations has come
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into effect, the license issued under this
part no longer authorizes operation of
the reactor or emplacement or retention
of fuel into the reactor vessel.
(c) Decommissioning will be
completed within 60 years of permanent
cessation of operations. Completion of
decommissioning beyond 60 years will
be approved by the Commission only
when necessary to protect public health
and safety. Factors that will be
considered by the Commission in
evaluating an alternative that provides
for completion of decommissioning
beyond 60 years of permanent cessation
of operations include unavailability of
waste disposal capacity and other sitespecific factors affecting the licensee’s
capability to carry out
decommissioning, including presence of
other nuclear facilities at the site.
(d)(1) Prior to or within 2 years
following permanent cessation of
operations, the licensee must submit a
PSDAR and site-specific DCE in
accordance with § 53.1060(e).
(2) The NRC must notice receipt of the
PSDAR and make the PSDAR publicly
available and publish notice of its
availability for public comment in the
Federal Register. The NRC must also
schedule a public meeting readily
accessible to individuals in the vicinity
of the licensee’s facility. The NRC must
publish a notice in the Federal Register
and in a forum, such as local
newspapers, that is readily accessible to
individuals in the vicinity of the site,
announcing the date, time, and location
of the meeting, along with a brief
description of the purpose of the
meeting.
(e) Licensees must not perform any
major decommissioning activities, as
defined in § 53.020, until 90 days after
the NRC has received the licensee’s
PSDAR submittal and until
certifications of permanent cessation of
operations and permanent removal of
fuel from the reactor vessel, as required
under paragraph (a) of this section, have
been submitted.
(f) Licensees must not perform any
decommissioning activities, as defined
in § 53.020, that—
(1) Foreclose release of the site for
possible unrestricted use;
(2) Result in significant
environmental impacts not previously
reviewed; or
(3) Result in there no longer being
reasonable assurance that adequate
funds will be available for
decommissioning.
(g) In taking actions permitted under
§ 53.1540 following submittal of the
PSDAR, the licensee must notify the
NRC in writing, and send a copy to the
affected State(s), before performing any
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decommissioning activity inconsistent
with, or making any significant
schedule change from, those actions and
schedules described in the PSDAR,
including changes that increase the
decommissioning cost by more than 20
percent from the previously provided
DCE.
(h) Licensees may use
decommissioning trust funds consistent
with the limitations of § 53.1045(a).
Licensees must report on the status of
decommissioning trust funds consistent
with the requirements of § 53.1060.
(i) Licensees must submit an
application for termination of license in
accordance with § 53.1070. The
application for termination of license
must be accompanied or preceded by a
license termination plan to be submitted
for NRC approval.
(1) The license termination plan must
be a supplement to the Final Safety
Analysis Report or equivalent and must
be submitted at least 2 years before
termination of the license date.
(2) The license termination plan must
include—
(i) A site characterization;
(ii) Identification of remaining
dismantlement activities;
(iii) Plans for site remediation;
(iv) Detailed plans for the final
radiation survey;
(v) A description of the end use of the
site, if restricted;
(vi) An updated site-specific estimate
of remaining decommissioning costs;
(vii) A supplement to the
environmental report, pursuant to
§ 51.53 of this chapter, describing any
new information or significant
environmental change associated with
the licensee’s proposed termination
activities; and
(viii) Identification of parts, if any, of
the facility or site that were released for
use before approval of the license
termination plan.
(3) Following receipt of the license
termination plan, the NRC must make
the license termination plan publicly
available and publish notice of its
availability for public comment in the
Federal Register. The NRC must also
schedule a public meeting readily
accessible to individuals in the vicinity
of the licensee’s facility upon receipt of
the license termination plan. The NRC
must publish a notice in the Federal
Register and in a forum, such as local
newspapers, that is readily accessible to
individuals in the vicinity of the site,
announcing the date, time, and location
of the meeting, along with a brief
description of the purpose of the
meeting.
(j) If the license termination plan
demonstrates that the remainder of
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decommissioning activities will be
performed in accordance with the
regulations in this chapter, will not be
inimical to the common defense and
security or to the health and safety of
the public, and will not have a
significant effect on the quality of the
environment and after notice to
interested persons, the Commission will
approve the plan, by license
amendment, subject to such conditions
and limitations as it deems appropriate
and necessary and authorize
implementation of the license
termination plan.
(k) The Commission will terminate
the license if it determines that—
(1) The remaining dismantlement has
been performed in accordance with the
approved license termination plan, and
(2) The final radiation survey and
associated documentation, including an
assessment of dose contributions
associated with parts released for use
before approval of the license
termination plan, demonstrate that the
facility and site have met the criteria for
decommissioning in subpart E of 10
CFR part 20.
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§ 53.1075 Program requirements during
decommissioning.
(a) Licensees that have submitted the
certifications required under § 53.1070
must maintain a decommissioning fire
protection program to address the
potential for fires that could cause the
release or spread of radioactive
materials.
(1) The objectives of the
decommissioning fire protection
program are to
(i) Reasonably prevent these fires from
occurring;
(ii) Rapidly detect, control, and
extinguish those fires that do occur and
that could result in a radiological
hazard; and
(iii) Ensure that the risk of fireinduced radiological hazards to the
public, environment, and plant
personnel is minimized.
(2) The licensee must assess the
decommissioning fire protection
program on a regular basis. The licensee
must revise the decommissioning fire
protection program documentation as
appropriate throughout the various
stages of facility decommissioning.
(3) The licensee may make changes to
the decommissioning fire protection
program without NRC approval if these
changes do not reduce the effectiveness
of fire protection for structures, systems,
and components that could result in a
radiological hazard, taking into account
the decommissioning plant conditions
and activities.
(b) [Reserved]
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§ 53.1080 Release of part of a commercial
nuclear plant or site for unrestricted use.
(a) Prior written NRC approval is
required to release part of a commercial
nuclear plant or site for unrestricted use
at any time before receiving approval of
a license termination plan. Section
53.1060 specifies recordkeeping
requirements associated with partial
release. Holders of an OL or COL under
this part seeking NRC review and
approval must—
(1) Evaluate the effect of releasing the
property to ensure that—
(i) The dose to individual members of
the public does not exceed the limits
and standards of subpart D of 10 CFR
part 20;
(ii) There is no reduction in the
effectiveness of emergency planning or
physical security;
(iii) Effluent releases remain within
license conditions;
(iv) The environmental monitoring
program and offsite dose calculation
manual are revised to account for the
changes;
(v) The siting criteria of subpart D of
this part continue to be met; and
(vi) All other applicable statutory and
regulatory requirements continue to be
met.
(2) Perform a historical site
assessment of the part of the commercial
nuclear plant or site to be released; and
(3) Perform surveys adequate to
demonstrate compliance with the
radiological criteria for unrestricted use
specified in § 20.1402 of this chapter for
impacted areas.
(b) For release of non-impacted areas,
the licensee may submit a written
request for NRC review and approval of
the release if a license amendment is not
otherwise required. The request
submittal must include—
(1) The results of the evaluations
performed in accordance with
paragraphs (a)(1) and (a)(2) of this
section;
(2) A description of the part of the
commercial nuclear plant or site to be
released;
(3) The schedule for release of the
property;
(4) The results of the evaluations
performed in accordance with
§ 53.1540; and
(5) A discussion that provides the
reasons for concluding that the
environmental impacts associated with
the licensee’s proposed release of the
property will be bounded by
appropriate previously issued
environmental impact statements.
(c) After receiving a request from the
licensee for NRC approval of the release
of a non-impacted area, the NRC must—
(1) Determine whether the licensee
has adequately evaluated the effect of
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87077
releasing the property as required by
paragraph (a)(1) of this section;
(2) Determine whether the licensee’s
classification of any release areas as
non- impacted is adequately justified;
and
(3) If determining that the licensee’s
submittal is adequate, inform the
licensee in writing that the release is
approved.
(d) For release of impacted areas, the
licensee must submit an application for
amendment of its license for the release
of the property. The application must
include—
(1) The information specified in
paragraphs (b)(1) through (b)(3) of this
section;
(2) The methods used for and results
obtained from the radiation surveys
required to demonstrate compliance
with the radiological criteria for
unrestricted use specified in § 20.1402;
and
(3) A supplement to the
environmental report, under § 51.53 of
this chapter, describing any new
information or significant
environmental change associated with
the licensee’s proposed release of the
property.
(e) After receiving a license
amendment application from the
licensee for the release of an impacted
area, the NRC must—
(1) Determine whether the licensee
has adequately evaluated the effect of
releasing the property as required by
paragraph (a)(1) of this section;
(2) Determine whether the licensee’s
classification of any release areas as
non-impacted is adequately justified;
(3) Determine whether the licensee’s
radiation survey for an impacted area is
adequate; and
(4) If determining that the licensee’s
submittal is adequate, approve the
licensee’s amendment application.
(f) The NRC must publish notice
receipt of the release approval request or
license amendment application in the
Federal Register and make the approval
request or license amendment
application available for public
comment. Before acting on an approval
request or license amendment
application submitted in accordance
with this section, the NRC must conduct
a public meeting readily accessible to
individuals in the vicinity of the
licensee’s facility for the purpose of
obtaining public comments on the
proposed release of part of the
commercial nuclear plant or site. The
NRC must publish a document in the
Federal Register and in a forum, such
as local newspapers, which is readily
accessible to individuals in the vicinity
of the site, announcing the date, time,
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and location of the meeting, along with
a brief description of the purpose of the
meeting.
Subpart H—Licenses, Certifications,
and Approvals
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§ 53.1100 Filing of application for licenses,
certifications, or approvals; oath or
affirmation.
(a) Serving of applications. (1) Each
filing of an application for a standard
design approval, standard design
certification, or license under this part,
and any amendments to the
applications, must be submitted to the
U.S. Nuclear Regulatory Commission
(NRC) under § 53.040, as applicable.
(2) Each applicant for a construction
permit (CP), early site permit, combined
license (COL), or manufacturing license
(ML) under this part must, upon
notification by the presiding officer
designated to conduct the public
hearing required by the Atomic Energy
Act of 1954, as amended, (the Act)
update the application and serve the
updated copies of the application or
parts of it, eliminating all superseded
information, together with an index of
the updated application, as directed by
presiding officer. Any subsequent
amendment to the application must be
served on those served copies of the
application and must be submitted to
the NRC as specified in § 53.040, as
applicable.
(3) The applicant must make a copy
of the updated application available at
the public hearing for the use of any
other parties to the proceeding and must
certify that the updated copies of the
application contain the current contents
of the application submitted in
accordance with the requirements under
this part.
(4) At the time of filing an
application, the Commission will make
available at the NRC website, https://
www.nrc.gov, a copy of the application,
subsequent amendments, and other
records pertinent to the matter that is
the subject of the application for public
inspection and copying.
(5) The serving of copies required by
this section must not occur until the
application has been docketed under
§ 2.101(a) of this chapter. Copies must
be submitted to the Commission, as
specified in § 53.040, as applicable, to
enable the Director, Office of Nuclear
Reactor Regulation to determine
whether the application is sufficiently
complete to permit docketing.
(b) Oath or affirmation. Each
application for a standard design
approval, standard design certification,
or license, including, whenever
appropriate, a CP or early site permit, or
amendment of it, and each amendment
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of each application must be executed in
a signed original by the applicant or
duly authorized officer thereof under
oath or affirmation.
(c) [Reserved]
(d) [Reserved]
(e) Filing fees. Each application for a
standard design approval, standard
design certification, or commercial
nuclear plant license under this part,
including, whenever appropriate, a CP,
COL, operating license (OL), ML, or
early site permit, other than a license
exempted from 10 CFR part 170, must
be accompanied by the fee prescribed in
10 CFR part 170. No fee will be required
to accompany an application for
renewal, amendment, or termination of
a CP, OL, COL, or ML, except as
provided in § 170.21 of this chapter.
(f) Environmental report. An
application for a CP, OL, early site
permit, design certification, COL, or ML
for a commercial nuclear plant must be
accompanied by an environmental
report required under subpart A of 10
CFR part 51.
§ 53.1101
Requirement for license.
Except as provided in § 53.1120, no
person within the United States may
transfer or receive in interstate
commerce, manufacture, produce,
transfer, acquire, possess, or use any
utilization facility except as authorized
by a license issued by the Commission.
§ 53.1103
licenses.
Combining applications and
(a) An applicant may combine several
applications in one application for
different kinds of licenses under the
regulations in this chapter.
(b) The Commission may combine in
a single license the activities of an
applicant which would otherwise be
licensed separately.
§ 53.1106
Elimination of repetition.
An applicant may incorporate by
reference in its application information
contained in previous applications,
statements, or reports filed with the
Commission, provided, however, that
such references are clear and specific.
§ 53.1109 Contents of applications;
general information.
Each application must include, unless
otherwise indicated in this subpart—
(a) Name of applicant;
(b) Address of applicant;
(c) Description of business or
occupation of applicant;
(d)(1) If applicant is an individual, the
citizenship of applicant;
(2) If applicant is a partnership, the
name, citizenship and address of each
partner and the principal location where
the partnership does business;
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(3) If applicant is a corporation or an
unincorporated association, the
following information:
(i) The State where it is incorporated
or organized and the principal location
where it does business;
(ii) The names, addresses and
citizenship of its directors and of its
principal officers; and
(iii) Whether it is owned, controlled,
or dominated by an alien, a foreign
corporation, or foreign government, and
if so, give details; or
(4) If the applicant is acting as agent
or representative of another person in
filing the application, identify the
principal and furnish information
required under this paragraph with
respect to such principal;
(e) The class and type of license
applied for, the use to which the facility
will be put, the period of time for which
the license is sought, and a list of other
licenses, except operator’s licenses,
issued or applied for in connection with
the proposed facility;
(f) [Reserved]
(g)(1) Except as provided in paragraph
(g)(2) of this section, if the application
is for an OL or COL for a commercial
nuclear plant, or if the application is for
an early site permit for a commercial
nuclear plant and contains plans for
coping with emergencies under
§ 53.1146(b)(2)(ii), the applicant must
submit the radiological emergency
response plans of State, local, and
participating Tribal governmental
entities in the United States that are
wholly or partially within the plume
exposure pathway emergency planning
zone (EPZ),1 and the plans of State
governments wholly or partially within
the ingestion pathway EPZ.2 If the
application is for an early site permit
that, under § 53.1146(b)(2)(i), proposes
major features of the emergency plans
describing the EPZs, then the
descriptions of the EPZs must meet the
requirements of this paragraph.
Generally, the plume exposure pathway
EPZ for a commercial nuclear plant
must consist of an area about 10 miles
(16 km) in radius and the ingestion
pathway EPZ must consist of an area
about 50 miles (80 km) in radius. The
exact size and configuration of the EPZs
surrounding a particular commercial
nuclear plant must be determined in
relation to the local emergency response
needs and capabilities as they are
affected by such conditions as
demography, topography, land
characteristics, access routes, and
jurisdictional boundaries. The size of
the EPZs also may be determined on a
case-by-case basis for gas-cooled
reactors and for reactors with an
authorized power level less than 250
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megawatt thermal. The plans for the
ingestion pathway must focus on such
actions as are appropriate to protect the
food ingestion pathway.
(2) Applicants for commercial nuclear
plants consisting of either small
modular reactors or non-light-water
reactors complying with § 50.160 of this
chapter who apply for a CP, an OL, a
COL, or an early site permit under this
part must submit as part of the
application the analysis used to
determine whether the criteria in
§ 53.1109(g)(2)(i)(A) and (B) are met
and, if they are met, the size of the
plume exposure pathway EPZ.
(i) The plume exposure pathway EPZ
is the area within which:
(A) Public dose, as defined in
§ 20.1003 of this chapter, is projected to
exceed 10 millisieverts (1 rem) total
effective dose equivalent over 96 hours
from the release of radioactive materials
from the facility considering accident
likelihood and source term, timing of
the accident sequence, and meteorology;
and
(B) Pre-determined, prompt protective
measures are necessary.
(ii) If the application is for an OL or
COL or if the application is for an early
site permit and contains plans for
coping with emergencies under
§ 53.1146(b)(2)(ii), and if the plume
exposure pathway EPZ extends beyond
the site boundary:
(A) The applicant must submit
radiological emergency response plans
of State, local, and participating Tribal
governmental entities in the United
States that are wholly or partially within
the plume exposure pathway EPZ.
(B) The exact configuration of the
plume exposure pathway EPZ
surrounding the facility shall be
determined in relation to the local
emergency response needs and
capabilities as they are affected by such
conditions as demography, topography,
land characteristics, access routes, and
jurisdictional boundaries.
(iii) If the application is for an early
site permit that, under § 53.1146(b)(2)(i),
proposes major features of the
emergency plans and describes the EPZ,
and if the EPZ extends beyond the site
boundary, then the exact configuration
of the plume exposure pathway EPZ
surrounding the facility must be
determined in relation to the local
emergency response needs and
capabilities as they are affected by such
conditions as demography, topography,
land characteristics, access routes, and
jurisdictional boundaries.
(h) [Reserved]
(i) A list of the names and addresses
of such regulatory agencies as may have
jurisdiction over the rates and services
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incident to the proposed activity, and a
list of trade and news publications
which circulate in the area where the
proposed activity will be conducted and
which are considered appropriate to
give reasonable notice of the application
to those municipalities, private utilities,
public bodies, and cooperatives, which
might have a potential interest in the
facility; and
(j) If the application contains
Restricted Data or classified National
Security information, confirmation that
all Restricted Data and classified
National Security information are
separated from the unclassified
information.
1 EPZs are discussed in NUREG–0396, U.S.
Environmental Protection Agency 520/1–78–
016, ‘‘Planning Basis for the Development of
State and Local Government Radiological
Emergency Response Plans in Support of
Light-Water Nuclear Power Plants,’’
December 1978.
2 If the State, local, and participating Tribal
emergency response plans have been
previously provided to the NRC for inclusion
in the facility docket, the applicant need only
provide the appropriate reference to meet
this requirement.
§ 53.1112
Environmental conditions.
(a) Each CP, early site permit, and
COL under this part may include
conditions to address environmental
issues during construction. These
conditions are to be set out in an
attachment to the license, which is
incorporated in and made a part of the
license. These conditions will be
derived from information contained in
the environmental report submitted
pursuant to § 51.50 of this chapter, as
analyzed and evaluated in the NRC
record of decision and will identify the
obligations of the licensee in the
environmental area, including, as
appropriate, requirements for reporting
and keeping records of environmental
data, and any conditions and
monitoring requirement for the
protection of the nonaquatic
environment.
(b) Each license authorizing operation
of a commercial nuclear plant under
this part, and each license for a
commercial nuclear plant for which the
certification of permanent cessation of
operations required under § 53.1070 has
been submitted may include conditions
to address environmental issues during
operation and decommissioning. These
conditions are to be set out in an
attachment to the license, which is
incorporated in and made a part of the
license. These conditions will be
derived from information contained in
the environmental report or the
supplement to the environmental report
submitted under §§ 51.50 and 51.53 of
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87079
this chapter as analyzed and evaluated
in the NRC record of decision, and will
identify the obligations of the licensee
in the environmental area, including, as
appropriate, requirements for reporting
and keeping records of environmental
data and any conditions and monitoring
requirement for the protection of the
nonaquatic environment.
§ 53.1115 Agreement limiting access to
classified information.
As part of its application and in any
event before the receipt of Restricted
Data or classified National Security
Information or the issuance of a license
or standard design approval under this
part, or before the Commission has
adopted a final standard design
certification rule under this part, the
applicant must agree in writing that it
will not permit any individual to have
access to or any facility to possess
Restricted Data or classified National
Security Information until the
individual and/or facility has been
approved for access under the
provisions of 10 CFR parts 25 and/or 95.
The agreement of the applicant becomes
part of the license or standard design
approval.
§ 53.1118
Ineligibility of certain applicants.
Any person who is a citizen, national,
or agent of a foreign country, or any
corporation, or other entity which the
Commission knows or has reason to
believe is owned, controlled, or
dominated by an alien, a foreign
corporation, or a foreign government,
will be ineligible to apply for and obtain
a license.
§ 53.1120 Exceptions and exemptions
from licensing requirements.
Nothing in this part must be deemed
to require a license for—
(a) The manufacture, production, or
acquisition by the Department of
Defense of any utilization facility
authorized pursuant to section 91 of the
Act or the use of such facility by the
Department of Defense or by a person
under contract with and for the account
of the Department of Defense;
(b) Except to the extent that the
Department of Energy facilities of the
types subject to licensing pursuant to
section 202 of the Energy
Reorganization Act of 1974, as
amended, are involved—
(1)(i) The processing, fabrication or
refining of special nuclear material
(SNM) or the separation of SNM, or the
separation of SNM from other
substances by a prime contractor of the
Department of Energy under a prime
contract for—
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(A) The performance of work for the
Department of Energy at a United States
government-owned or controlled site;
(B) Research in, or development,
manufacture, storage, testing or
transportation of, atomic weapons or
components thereof; or
(C) The use or operation of a
utilization facility in a United States
owned vehicle or vessel; or
(ii) The processing, fabrication or
refining of SNM of the separation of
SNM, or the separation of SNM from
other substances by a prime contractor
or subcontractor of the Commission or
the Department of Energy under a prime
contract or subcontract when the
Commission determines that the
exemption of the prime contractor or
subcontractor is authorized by law; and
that, under the terms of the contract or
subcontract, there is adequate assurance
that the work thereunder can be
accomplished without undue risk to the
public health and safety; or
(2)(i) The construction or operation of
a utilization facility for the Department
of Energy at a United States
government-owned or controlled site,
including the transportation of the
utilization facility to or from such site
and the performance of contract services
during temporary interruptions of such
transportation; or the construction or
operation of a utilization facility for the
Department of Energy in the
performance of research in, or
development, manufacture, storage,
testing, or transportation of, atomic
weapons or components thereof; or the
use or operation of a utilization facility
for the Department of Energy in a
United States government-owned
vehicle or vessel; provided that such
activities are conducted by a prime
contractor of the Department of Energy
under a prime contract with the
Department of Energy; or
(ii) The construction or operation of a
utilization facility by a prime contractor
or subcontractor of the Commission or
the Department of Energy under his or
her prime contract or subcontract when
the Commission determines that the
exemption of the prime contractor or
subcontractor is authorized by law; and
that, under the terms of the contract or
subcontract, there is adequate assurance
that the work thereunder can be
accomplished without undue risk to the
public health and safety; or
(c) The transportation or possession of
any utilization facility by a common or
contract carrier or warehouse employee
in the regular course of carriage for
another or storage incident thereto.
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§ 53.1121 Public inspection of
applications.
Applications and documents
submitted to the Commission in
connection with applications may be
made available for public inspection
under the provisions of part 2 of this
chapter.
§ 53.1124
Relationship between sections.
(a) Limited work authorization. An
application for a limited work
authorization (LWA) under this part
may be submitted as part of an
application for an early site permit, CP,
or COL under this part as required in
§ 53.1130(a)(2).
(b) Early site permit. (1) A holder of
an early site permit may request an
LWA.
(2) An application for a CP or COL
under this part may, but need not,
reference an early site permit.
(c) Standard design approval. An
application for a standard design
approval under this part may, but need
not, reference an OL or custom COL
under this part that is essentially the
same as the information supporting the
standard design for which approval is
being requested.
(d) Standard design certification. An
application for a standard design
certification under this part may, but
need not, reference an OL or custom
COL under this part that is essentially
the same as the standard design for
which certification is being requested.
(e) Manufacturing license. (1) A
manufactured reactor manufactured
under an ML issued under this part may
only be transported to and installed at
a site for which a COL under this part
has been issued.
(2) An ML applicant under this part
may reference a standard design
certification or a standard design
approval under this part in its
application.
(f) Construction permit. An
application for a CP may, but need not,
reference a standard design certification
or standard design approval issued
under this part, respectively, and may
also reference an early site permit
issued under this part. In the absence of
a demonstration that an entity other
than the one originally sponsoring a
standard design certification is qualified
to supply a design, the Commission will
entertain an application for a CP that
references a standard design
certification issued under this part only
if the entity that sponsored the
certification supplies the design for the
applicant’s use.
(g) Operating license. (1) An
application for an OL under this part
may, but need not, reference an early
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site permit, standard design
certification, or standard design
approval issued under this part. In the
absence of a demonstration that an
entity other than the one originally
sponsoring a standard design
certification is qualified to supply a
design, the Commission will entertain
an application for an OL that references
a standard design certification issued
under this part only if the entity that
sponsored the certification supplies the
design for the applicant’s use.
(2) The holder of a CP must, at the
time of submission of the Final Safety
Analysis Report (FSAR), file an
application for an OL.
(h) Combined licenses. An application
for a COL under this part may, but need
not, reference an early site permit,
standard design certification, standard
design approval, or ML issued under
this part. In the absence of a
demonstration that an entity other than
the one originally sponsoring and
obtaining a standard design certification
is qualified to supply a design, the
Commission will entertain an
application for a COL that references a
standard design certification issued
under this part only if the entity that
sponsored the certification supplies the
design for the applicant’s use.
§ 53.1130
Limited work authorizations.
(a) Request for limited work
authorization. (1) Any person to whom
the Commission may otherwise issue
either a license or permit related to a
commercial nuclear plant may request
an LWA allowing that person to perform
the driving of piles, subsurface
preparation, placement of backfill,
concrete, or permanent retaining walls
within an excavation, and installation of
the foundation, including placement of
concrete, any of which are for a
structure, system, or component (SSC)
of the facility for which either a CP or
COL is otherwise required under
§ 53.610.
(2) An application for an LWA may be
submitted as part of a complete
application for a CP or COL in
accordance with § 2.101(a)(1) through
(a)(5) of this chapter, or as a partial
application in accordance with
§ 2.101(a)(9) of this chapter. An
application for an LWA by the holder of
an early site permit must be submitted
as a complete application in accordance
with § 2.101(a)(1) through (a)(4) of this
chapter.
(3) The application must include—
(i) A Safety Analysis Report required
by §§ 53.1146, 53.1309 or 53.1416, as
applicable, a description of the activities
requested to be performed, and the
design and construction information
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otherwise required by the Commission’s
rules and regulations to be submitted for
a CP or COL under this part but limited
to those portions of the facility that are
within the scope of the LWA. The Safety
Analysis Report must demonstrate that
activities conducted under the LWA
will be conducted in compliance with
the technically relevant Commission
requirements in 10 CFR chapter I
applicable to the design of those
portions of the facility within the scope
of the LWA;
(ii) An environmental report in
accordance with § 51.49 of this chapter;
and
(iii) A plan for redress of activities
performed under the LWA, should
limited work activities be terminated by
the holder, or the LWA be revoked by
the NRC or upon effectiveness of the
Commission’s final decision denying
the associated CP or COL application, as
applicable.
(b) Issuance of limited work
authorization. (1) The Director, Office of
Nuclear Reactor Regulation may issue
an LWA only after—
(i) The NRC staff issues the final
environmental impact statement for the
LWA under subpart A of part 51 of this
chapter;
(ii) The presiding officer makes the
finding in §§ 51.105(c) or 51.107(d) of
this chapter, as applicable;
(iii) The Director determines that the
applicable standards and requirements
of the Act, and the Commission’s
regulations applicable to the activities to
be conducted under the LWA, have
been met, the applicant is technically
qualified to engage in the activities
authorized, and that issuance of the
LWA will provide reasonable assurance
of adequate protection to public health
and safety and will not be inimical to
the common defense and security; and
(iv) The presiding officer finds that
there are no unresolved safety issues
relating to the activities to be conducted
under the LWA that would constitute
good cause for withholding the
authorization.
(2) Each LWA will specify the
activities that the holder is authorized to
perform.
(c) Effect of limited work
authorization. Any activities
undertaken under an LWA are entirely
at the risk of the applicant and, except
as to the matters determined under
paragraph (b)(1) of this section, the
issuance of the LWA has no bearing on
the issuance of a CP or COL with respect
to the requirements of the Act and rules,
regulations, or orders issued under the
Act. The environmental impact
statement for a CP or COL application
for which an LWA was previously
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issued will not address, and the
presiding officer will not consider, the
sunk costs of the holder of the LWA in
determining the proposed action (i.e.,
issuance of the CP or COL).
(d) Implementation of redress plan. If
construction is terminated by the
holder, the underlying application is
withdrawn by the applicant or denied
by the NRC, or the LWA is revoked by
the NRC, then the holder must begin
implementation of the redress plan in a
reasonable time. The holder must also
complete the redress of the site no later
than 18 months after termination of
construction, revocation of the LWA, or
upon effectiveness of the Commission’s
final decision denying the associated CP
application or the associated COL
application, as applicable.
§ 53.1140
Early site permits.
Sections 53.1140 through 53.1188 set
out the requirements and procedures
applicable to Commission issuance of
an early site permit under this part for
approval of a site for a commercial
nuclear plant separate from the filing of
an application for a CP or COL for the
facility.
§ 53.1143
Filing of applications.
Any person who may apply for a CP
or for a COL under this part, may file
an application for an early site permit
with the Director, Office of Nuclear
Reactor Regulation. An application for
an early site permit may be filed
notwithstanding the fact that an
application for a CP or a COL has not
been filed in connection with the site
for which a permit is sought.
§ 53.1144 Contents of applications for
early site permits; general information.
The application must contain all of
the information required by § 53.1109(a)
through (d) and (j).
§ 53.1146 Contents of applications for
early site permits; technical information.
(a) The application must contain—
(1) A Site Safety Analysis Report that
must include the following:
(i) The specific number, type, and
thermal power level of the facilities, or
range of possible facilities, for which the
site may be used;
(ii) The anticipated maximum levels
of radiological and thermal effluents
each facility will produce;
(iii) The type of cooling systems,
including intakes and outflows, where
appropriate, that may be associated with
each facility;
(iv) The boundaries of the site;
(v) The proposed general location of
each facility on the site;
(vi) The external hazards and site
characteristics required by this part;
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(vii) The location and description of
any nearby industrial, military, or
transportation facilities and routes;
(viii) The existing and projected
future population profile of the area
surrounding the site;
(ix) A description and assessment of
the site on which a facility is to be
located. The assessment must address
the requirements of subpart D of this
part;
(x) Information demonstrating that
site characteristics are such that
adequate security plans and measures
can be developed; and
(xi) A description of the quality
assurance program (QAP) required by
appendix B to part 50 of this chapter
applied to site-related activities for the
future design, fabrication, construction,
and testing of the SSCs of a facility or
facilities that may be constructed on the
site.
(2) A complete environmental report
as required by § 51.50(b) of this chapter.
(b)(1) The Site Safety Analysis Report
must identify physical characteristics of
the proposed site, such as egress
limitations from the area surrounding
the site, that could pose a significant
impediment to the development of
emergency plans. If physical
characteristics are identified that could
pose a significant impediment to the
development of emergency plans, the
application must identify measures that
would, when implemented, mitigate or
eliminate the significant impediment.
(2) The Site Safety Analysis Report
may also—
(i) Propose major features of the
emergency plans, under either § 50.160
or the requirements in appendix E to
part 50 and § 50.47(b) of this chapter, as
applicable, such as the exact size and
configuration of the EPZs, for review
and approval by the NRC, in
consultation with the Federal
Emergency Management Agency
(FEMA), as applicable, in the absence of
complete and integrated emergency
plans; or
(ii) Propose complete and integrated
emergency plans for review and
approval by the NRC, in consultation
with FEMA, as applicable, in
accordance with either § 50.160 or the
requirements in appendix E to part 50
and § 50.47(b) of this chapter. To the
extent approval of emergency plans is
sought, the application must contain the
information required by § 53.1109(g).
(3) Emergency plans submitted under
paragraph (b)(2)(ii) of this section must
include the proposed inspections, tests,
and analyses that the holder of a COL
referencing the early site permit must
perform, and the acceptance criteria that
are necessary and sufficient to provide
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reasonable assurance that, if the
inspections, tests, and analyses are
performed and the acceptance criteria
met, the facility has been constructed
and will be operated in conformity with
the emergency plans, the provisions of
the Act, and the Commission’s rules and
regulations. Major features of an
emergency plan submitted under
paragraph (b)(2)(i) of this section may
include proposed inspections, tests,
analyses, and acceptance criteria
(ITAAC).
(4) Under paragraphs (b)(1) and
(b)(2)(i) of this section, the Site Safety
Analysis Report must include, where
appropriate, a description of contacts
and arrangements made with Federal,
State, participating Tribal, and local
governmental agencies with emergency
planning responsibilities. The Site
Safety Analysis Report must contain any
certifications that have been obtained. If
these certifications, where appropriate,
cannot be obtained, the Site Safety
Analysis Report must contain
information, including a utility plan,
sufficient to show that the proposed
plans provide reasonable assurance that
adequate protective measures can and
will be taken in the event of a
radiological emergency at the site.
Under the option set forth in paragraph
(b)(2)(ii) of this section, the applicant
must make good faith efforts, where
appropriate, to obtain from the same
governmental agencies certifications
that—
(i) The proposed emergency plans are
practicable;
(ii) These agencies are committed to
participating in any further
development of the plans, including any
required field demonstrations; and
(iii) That these agencies are
committed to executing their
responsibilities under the plans in the
event of an emergency.
(c) An applicant may request that an
LWA under § 53.1130 be issued in
conjunction with the early site permit.
The application must include the
information otherwise required by
§ 53.1130.
(d) Each applicant for an early site
permit under this part must protect
safeguards information against
unauthorized disclosure in accordance
with the requirements in §§ 73.21 and
73.22 of this chapter, as applicable.
§ 53.1149
Review of applications.
(a) Standards for review of
applications. Applications filed under
this part will be reviewed according to
the applicable standards set out in this
part. In addition, the Commission must
prepare an environmental impact
statement during review of the
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application, under the applicable
provisions of 10 CFR part 51. The
Commission must determine, after
consultation with FEMA, as applicable,
whether the information required of the
applicant by § 53.1146(b)(1) shows that
there is no significant impediment to
the development of emergency plans
that cannot be mitigated or eliminated
by measures proposed by the applicant,
whether any major features of
emergency plans submitted by the
applicant under § 53.1146(b)(2)(i) are
acceptable under either § 50.160 or
appendix E to part 50 and § 50.47(b) of
this chapter, and whether any
emergency plans submitted by the
applicant under § 53.1146(b)(2)(ii)
provide reasonable assurance that
adequate protective measures can and
will be taken in the event of a
radiological emergency.
(b) Administrative review of
applications; hearings. An early site
permit application is subject to all
procedural requirements in 10 CFR part
2, including the requirements for
docketing in § 2.101(a)(1) through (4) of
this chapter, and the requirements for
issuance of a notice of hearing in
§ 2.104(a) and (d) of this chapter,
provided that the designated sections
may not be construed to require that the
environmental report, or draft or final
environmental impact statement
includes an assessment of the benefits of
construction and operation of the
reactor or reactors, or an analysis of
alternative energy sources. The
presiding officer in an early site permit
hearing must not admit contentions
proffered by any party concerning an
assessment of the benefits of
construction and operation of the
reactor or reactors, or an analysis of
alternative energy sources if those issues
were not addressed by the applicant in
the early site permit application. All
hearings conducted on applications for
early site permits filed under this part
are governed by the procedures
contained in subparts C, G, L, and N of
10 CFR part 2, as applicable.
Commission deems appropriate, if the
Commission finds that—
(1) An application for an early site
permit demonstrates compliance with
the applicable standards and
requirements of the Act and the
Commission’s regulations;
(2) Notifications, if any, to other
agencies or bodies have been duly
made;
(3) There is reasonable assurance that
the site is in conformity with the
provisions of the Act and the
Commission’s regulations;
(4) The applicant is technically
qualified to engage in any activities
authorized;
(5) The proposed ITAAC, including
any on emergency planning, are
necessary and sufficient, within the
scope of the early site permit, to provide
reasonable assurance that the facility
has been constructed and will be
operated in conformity with the license,
the provisions of the Act, and the
Commission’s regulations;
(6) Issuance of the permit will not be
inimical to the common defense and
security or to the health and safety of
the public;
(7) Any significant adverse
environmental impact resulting from
activities requested under § 53.1146(c)
can be redressed; and
(8) The findings required by subpart
A of 10 CFR part 51 have been made.
(b) The early site permit must specify
the site characteristics, design
parameters, and terms and conditions of
the early site permit the Commission
deems appropriate. Before issuance of
either a CP or COL referencing an early
site permit, the Commission must find
that any relevant terms and conditions
of the early site permit have been met.
Any terms or conditions of the early site
permit that could not be met by the time
of issuance of the CP or COL, must be
set forth as terms or conditions of the CP
or COL.
(c) The early site permit must specify
those § 53.1130(b) activities requested
under § 53.1146(c) that the permit
holder is authorized to perform.
§ 53.1155 Referral to the Advisory
Committee on Reactor Safeguards.
§ 53.1161
The Commission must refer a copy of
the application for an early site permit
to the Advisory Committee on Reactor
Safeguards (ACRS). The ACRS must
report on those portions of the
application which concern safety.
§ 53.1158
Issuance of early site permit.
(a) After conducting a hearing under
§ 53.1149(b) and receiving the report to
be submitted by the ACRS under
§ 53.1155, the Commission may issue an
early site permit, in the form the
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Extent of activities permitted.
If the activities authorized by
§ 53.1158(c) are performed and the site
is not referenced in an application for a
CP or a COL issued under this part
while the permit remains valid, then the
early site permit remains in effect solely
for the purpose of site redress, and the
holder of the permit must redress the
site under the terms of the site redress
plan required by § 53.1146(c). If, before
redress is complete, a use not envisaged
in the redress plan is found for the site
or parts thereof, the holder of the permit
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must carry out the redress plan to the
greatest extent possible consistent with
the alternate use.
§ 53.1164
Duration of permit.
(a) Except as provided in paragraph
(b) of this section, an early site permit
issued under this subpart may be valid
for not less than 10, nor more than 20
years from the date of issuance.
(b) An early site permit continues to
be valid beyond the date of expiration
in any proceeding on a CP application
or a COL application that references the
early site permit and is docketed before
the date of expiration of the early site
permit, or, if a timely application for
renewal of the permit has been
docketed, before the Commission has
determined whether to renew the
permit.
(c) An applicant for a CP or COL may,
at its own risk, reference in its
application a site for which an early site
permit application has been docketed
but not granted.
(d) Upon issuance of a CP or COL, a
referenced early site permit is
subsumed, to the extent referenced, into
the CP or COL.
§ 53.1167 Limited work authorization after
issuance of early site permit.
A holder of an early site permit may
request an LWA under § 53.1130.
§ 53.1170
Transfer of early site permit.
An application to transfer an early site
permit will be processed under
§ 53.1570.
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§ 53.1173
Application for renewal.
(a) Not less than 12, nor more than 36
months before the expiration date stated
in the early site permit, or any later
renewal period, the permit holder may
apply for a renewal of the permit. An
application for renewal must contain all
information necessary to bring up to
date the information and data contained
in the previous application.
(b) Any person whose interests may
be affected by renewal of the permit
may request a hearing on the
application for renewal. The request for
a hearing must comply with § 2.309 of
this chapter. If a hearing is granted,
notice of the hearing will be published
under § 2.309 of this chapter.
(c) An early site permit, either original
or renewed, for which a timely
application for renewal has been filed,
remains in effect until the Commission
has determined whether to renew the
permit. If the permit is not renewed, it
continues to be valid in certain
proceedings in accordance with the
provisions of § 53.1164(b).
(d) The Commission must refer a copy
of the application for renewal to the
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ACRS. The ACRS must report on those
portions of the application which
concern safety and must apply the
criteria set forth in § 53.1176.
§ 53.1176
Criteria for renewal.
(a) The Commission must grant the
renewal if it determines that—
(1) The site complies with the Act, the
Commission’s regulations, and orders
applicable and in effect at the time the
site permit was originally issued; and
(2) Any new requirements the
Commission may wish to impose—
(i) Are necessary for adequate
protection to public health and safety or
common defense and security;
(ii) Are necessary for compliance with
the Commission’s regulations, and
orders applicable and in effect at the
time the site permit was originally
issued; or
(iii) Would provide a substantial
increase in overall protection of the
public health and safety or the common
defense and security to be derived from
the new requirements, and the direct
and indirect costs of implementation of
those requirements are justified in view
of this increased protection.
(b) A denial of renewal under the
provisions of § 53.1176(a) does not bar
the permit holder or another applicant
from filing a new application for the site
which proposes changes to the site or
the way that it is used to correct the
deficiencies cited in the denial of the
renewal.
§ 53.1179
Duration of renewal.
Each renewal of an early site permit
may be for not less than 10, nor more
than 20 years, plus any remaining years
on the early site permit then in effect
before renewal.
§ 53.1182
Use of site for other purposes.
A site for which an early site permit
has been issued under this part may be
used for purposes other than those
described in the permit, including the
location of other types of energy
facilities. The permit holder must
inform the Director, Office of Nuclear
Reactor Regulation (Director), of any
significant uses for the site which have
not been approved in the early site
permit. The information about the
activities must be given to the Director
at least 30 days in advance of any actual
construction or site modification for the
activities. The information provided
could be the basis for imposing new
requirements on the permit, under the
provisions of § 53.1188. If the permit
holder informs the Director that the
holder no longer intends to use the site
for a commercial nuclear plant, the
Director may terminate the permit.
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§ 53.1188 Finality of early site permit
determinations.
(a) Commission finality. (1) While an
early site permit is in effect under
§ 53.1164 or § 53.1179, the Commission
may not change or impose new site
characteristics, design parameters, or
terms and conditions, including
emergency planning requirements, on
the early site permit unless the
Commission—
(i) Determines that a modification is
necessary to bring the permit or the site
into compliance with the Commission’s
regulations and orders applicable and in
effect at the time the permit was issued;
(ii) Determines the modification is
necessary to assure adequate protection
of the public health and safety or the
common defense and security;
(iii) Determines that a modification is
necessary based on an update under
paragraph (b) of this section; or
(iv) Issues a variance requested under
paragraph (d) of this section.
(2) In making the findings required for
issuance of a CP, COL, or OL, or the
findings required by § 53.1452(g), or in
any enforcement hearing other than one
initiated by the Commission under
paragraph (a)(1) of this section, if the
application for the CP, COL, or OL
references an early site permit, the
Commission must treat as resolved
those matters resolved in the proceeding
on the application for issuance or
renewal of the early site permit, except
as provided for in paragraphs (b), (c),
and (d) of this section.
(i) If the Commission grants a CP
application that references an early site
permit and an application for an OL
references the CP, the Commission must
treat as resolved those matters resolved
in the proceeding for the issuance or
renewal of the early site permit, except
as provided for in paragraphs (b), (c),
and (d) of this section.
(ii) If the early site permit approved
an emergency plan (or major features
thereof) that is in use by a licensee of
a commercial nuclear plant, the
Commission must treat as resolved
changes to the early site permit
emergency plan (or major features
thereof) that are identical to changes
made to the licensee’s emergency plans
under § 53.1565 occurring after issuance
of the early site permit.
(iii) If the early site permit approved
an emergency plan (or major features
thereof) that is not in use by a licensee
of a commercial nuclear plant, the
Commission must treat as resolved
changes that are equivalent to those that
could be made under § 53.1565 without
prior NRC approval had the emergency
plan been in use by a licensee.
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(b) Updating of early site permitemergency preparedness. An applicant
for a CP, OL, or COL who has filed an
application referencing an early site
permit issued under this subpart must
update the emergency preparedness
information that was provided under
§ 53.1146(b) and discuss whether the
updated information materially changes
the bases for compliance with
applicable NRC requirements.
(c) Hearings and petitions. (1) In any
proceeding for the issuance of a CP, OL,
or COL referencing an early site permit,
contentions on the following matters
may be litigated in the same manner as
other issues material to the proceeding:
(i) The nuclear reactor proposed to be
built does not fit within one or more of
the site characteristics or design
parameters included in the early site
permit;
(ii) One or more of the terms and
conditions of the early site permit have
not been met;
(iii) A variance requested under
paragraph (d) of this section is
unwarranted or should be modified;
(iv) New or additional information is
provided in the application that
substantially alters the bases for a
previous NRC conclusion or constitutes
a sufficient basis for the Commission to
modify or impose new terms and
conditions related to emergency
preparedness; or
(v) Any significant environmental
issue that was not resolved in the early
site permit proceeding, or any issue
involving the impacts of construction
and operation of the facility that was
resolved in the early site permit
proceeding for which significant new
information has been identified.
(2) Any person may file a petition
requesting that the site characteristics,
design parameters, or terms and
conditions of the early site permit be
modified, or that the permit be
suspended or revoked. The petition will
be considered under § 2.206 of this
chapter. Before construction
commences, the Commission must
consider the petition and determine
whether any immediate action is
required. If the petition is granted, an
appropriate order will be issued.
Construction under the CP or COL will
not be affected by the granting of the
petition unless the order is made
immediately effective. Any change
required by the Commission in response
to the petition must demonstrate
compliance with the requirements of
paragraph (a)(1) of this section.
(d) Variances. An applicant for a CP,
OL, or COL referencing an early site
permit may include in its application a
request for a variance from one or more
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site characteristics, design parameters,
or terms and conditions of the early site
permit, or from the Site Safety Analysis
Report. In determining whether to grant
the variance, the Commission must
apply the same technically relevant
criteria applicable to the application for
the original or renewed early site
permit. Once a CP or COL referencing
an early site permit is issued, variances
from the early site permit will not be
granted for that CP or COL.
(e) Early site permit amendment. The
holder of an early site permit may not
make changes to the early site permit or
the Site Safety Analysis Report without
prior Commission approval. The request
for a change to the early site permit
must be in the form of an application for
a license amendment and must
demonstrate compliance with the
requirements of §§ 53.1510 and 53.1520.
§ 53.1200
Standard design approvals.
Sections 53.1200 through 53.1221 set
out procedures for the filing, NRC staff
review, and referral to the ACRS of
standard designs, or major portions
thereof, for a commercial nuclear plant
under this part.
§ 53.1203
Filing of applications.
Any person may submit a proposed
standard design for a commercial
nuclear plant to the NRC staff for its
review. The submittal may consist of
either the final design for the entire
facility or the final design for major
portions thereof.
§ 53.1206 Contents of applications for
standard design approvals; general
information.
The application must contain all of
the information required by § 53.1109(a)
through (c) and (j).
§ 53.1209 Contents of applications for
standard design approvals; technical
information.
(a) Major portion of a standard design.
If the applicant seeks review of a major
portion of a standard design, the
application need only contain the
information required by this section to
the extent the requirements are
applicable to the major portion of the
standard design for which NRC staff
approval is sought. If an applicant seeks
approval of a major portion of the
design, the scope of the application for
which approval is sought must include
all functional design criteria necessary
to demonstrate compliance with the
safety criteria in §§ 53.210, 53.220 and
53.450(e), as applicable, for the major
portion of the standard design for which
NRC staff approval is sought. Such
applicants must identify conditions
related to interfaces with systems
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outside the scope of the major portion
of the standard design for which NRC
staff approval is sought, and functional
or physical boundary conditions
between the major portion of the
standard design for which NRC staff
approval is sought and the remainder of
the standard design. These conditions
must be demonstrated when the
standard design approval is
incorporated into a subsequent CP,
design certification, ML, or COL
application.
(b) Final Safety Analysis Report. The
application must contain an FSAR that
describes the facility and the limits on
its operation, presents a safety analysis
of the SSCs and of the facility, or major
portions thereof, for which the applicant
seeks design approval, and must include
the following information:
(1) Site Parameters. The site
parameters postulated for the design
under this part, including the designbasis external hazard levels for the
relevant external hazards, and an
analysis and evaluation of the design in
terms of those site parameters.
(2) Design information. Except as
specified in this paragraph, an
application for a standard design
approval for a commercial nuclear plant
must include the design information
equivalent to that required for a
standard design certification under
§ 53.1239(a)(2) through (27) for those
portions of a commercial nuclear plant
included in the standard design
approval.
§ 53.1210 Contents of applications for
standard design approvals; other
application content.
(a) In addition to the FSAR, the
application must also include the
following:
(1) Availability Controls (if not
included in the FSAR). A description of
the controls on plant operations,
including availability controls, to
provide reasonable confidence that the
configurations and special treatments
for safety-related (SR) SSCs and nonsafety-related but safety-significant
(NSRSS) SSCs provide the capabilities
and reliabilities required to demonstrate
compliance with the safety criteria of
§ 53.220.
(2) Safeguards Information. A
description of the program to protect
Safeguards Information against
unauthorized disclosure in accordance
with the requirements in §§ 73.21 and
73.22 of this chapter, as applicable.
(b) If there are SSCs of the plant
which required research and
development to confirm the adequacy of
their design, provide a report in the
application which documents the
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resolution of any safety questions
associated with such SSCs.
(c) A description of how the
performance of each design feature has
been demonstrated capable of fulfilling
functional design criteria considering
interdependent effects through either
analysis, appropriate test programs,
prototype testing, operating experience,
or a combination thereof, in accordance
with § 53.440(a).
§ 53.1212 Standards for review of
applications.
Applications filed under this part will
be reviewed under the standards set out
in 10 CFR parts 20, 53, and 73.
§ 53.1215 Referral to the Advisory
Committee on Reactor Safeguards.
The Commission must refer a copy of
the application to the ACRS. The ACRS
must report on those portions of the
application which concern safety.
§ 53.1218
Staff approval of design.
(a) Upon completion of its review of
a submittal under §§ 53.1200 through
53.1221 and receipt of a report by the
ACRS under § 53.1215, the NRC staff
must publish a determination in the
Federal Register as to whether or not
the design is acceptable, subject to
appropriate terms and conditions, and
make an analysis of the design in the
form of a report available at the NRC
website, https://www.nrc.gov.
(b) A standard design approval issued
under this section is valid for 15 years
from the date of issuance and may not
be renewed. A design approval
continues to be valid beyond the date of
expiration in any proceeding on an
application for a CP, OL, COL, or ML
under this part that references the
design approval and is docketed before
the date of expiration of the design
approval.
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§ 53.1221 Finality of standard design
approvals; information requests.
(a) An approved design must be used
by and relied upon by the NRC staff and
the ACRS in their reviews of any
standard design certification or
individual facility license application
under this part that incorporates by
reference a standard design approved
under this part unless there exists
significant new information that
substantially affects the earlier
determination or other good cause.
(b) The determination and report by
the NRC staff do not constitute a
commitment to issue a permit or
license, or in any way affect the
authority of the Commission, Atomic
Safety and Licensing Board Panel, or
presiding officers in any proceeding
under part 2 of this chapter.
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(c) Except for information requests
seeking to verify compliance with the
current licensing basis of the standard
design approval, information requests to
the holder of a standard design approval
must be evaluated before issuance to
ensure that the burden to be imposed on
respondents is justified in view of the
potential safety significance of the issue
to be addressed in the requested
information. Each evaluation performed
by the NRC staff must be in accordance
with § 53.1580 and must be approved by
the Executive Director for Operations or
authorized designee before issuance of
the request.
(d) The Commission will require,
before granting a CP, COL, OL, or ML
that references a standard design
approval, that engineering documents,
such as analyses, drawings,
procurement specifications, or
construction and installation
specifications, be completed and
available for audit if the more detailed
information is necessary for the
Commission to verify the information in
the application and make its safety
determination, including the
determination that the application is
consistent with the design approval
information. This information may be
acquired by appropriate arrangements
with the design approval applicant.
§ 53.1230
Standard design certifications.
Sections 53.1230 through 53.1263 set
forth the requirements and procedures
applicable to the Commission’s issuance
of rules granting standard design
certifications for commercial nuclear
plants under this part separate from the
filing of an application for a CP or COL
for such a facility.
§ 53.1233
Filing of applications.
(a) An application for design
certification may be filed
notwithstanding the fact that an
application for a CP, COL, or ML for
such a facility has not been filed.
(b) The application must comply with
the applicable filing requirements of
§ 53.040 and §§ 2.811 through 2.819 of
this chapter.
§ 53.1236 Contents of applications for
standard design certifications; general
information.
The application must contain all of
the information required by § 53.1109(a)
through (c) and (j).
§ 53.1239 Contents of applications for
standard design certifications; technical
information.
The application must contain a level
of design information sufficient to
enable the Commission to judge the
applicant’s proposed means of assuring
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that construction conforms to the design
and to reach a final conclusion on all
safety questions associated with the
design before the certification is
granted. The information submitted for
a design certification must include
performance requirements and design
information sufficiently detailed to
permit the preparation of acceptance
and inspection requirements by the
NRC. The Commission will require,
before design certification, that
information normally contained in
engineering documents, such as
analyses, drawings, procurement
specifications, or construction and
installation specifications, be completed
and available for audit if the more
detailed information is necessary for the
Commission to verify the information in
the application and make its safety
determination.
(a) Final Safety Analysis Report. The
application must contain an FSAR that
describes the facility and the limits on
its operation, and presents a safety
analysis of the SSCs, and must include
the following information:
(1) Site Parameters. The site
parameters postulated for the design
under this part, including the designbasis external hazard levels for the
relevant external hazards, and an
analysis and evaluation of the design in
terms of those site parameters.
(2) Plant Description and Safety
Functions—(i) General Plant
Description. A general description of the
commercial nuclear plant including
reactor type, the intended use of the
reactor, nuclear design (e.g., neutron
spectrum, reactor control, multi-unit
reactor control), overall layout of the
plant including significant plant
features and SSCs, maximum power
level and the nature and inventory of
radioactive materials.
(ii) Safety functions. A description of
the primary and additional safety
functions required under § 53.230 and a
summary of how each safety function is
satisfied.
(3) Design Features and functional
design criteria—licensing-basis events.
(i) A description of the design features
required by § 53.400 and the functional
design criteria required by §§ 53.410
and 53.420 that, when combined with
corresponding human actions and
programmatic controls, demonstrate that
the plant will demonstrate compliance
with the safety criteria defined in
§ 53.210 and established in accordance
with § 53.220, or more restrictive
alternative criteria adopted under
§ 53.470, during licensing-basis events
(LBEs).
(ii) A description of how design
features demonstrate compliance with
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the requirements of § 53.440(a) through
(i) and (k) through (m).
(4) Design Features Supporting
Normal Operations. A description of the
design features required by § 53.425 to
support the holder of an OL or COL
complying with § 53.260 during normal
operations.
(5) Design Features and Functional
Design Criteria—aircraft impact. A
description of the design features and
functional design criteria required to
demonstrate compliance with the
requirements of § 53.440(j) for
addressing the impact of a large,
commercial aircraft.
(6) Earthquake engineering. The
information necessary to demonstrate
that the commercial nuclear plant
complies with the earthquake
engineering criteria in § 53.480.
(7) Programmatic Controls and
Interfaces. (i) A description of the
corresponding programmatic controls
and interfaces necessary to achieve and
maintain the reliability and capability of
SSCs relied upon to demonstrate
compliance with the functional design
criteria required by §§ 53.410 and
53.420 and the safety criteria in
§§ 53.210 and 53.220, or more restrictive
alternative criteria adopted under
§ 53.470, and necessary to maintain
consistency with analyses required by
§ 53.450.
(ii) For an application for a multi-unit
commercial nuclear plant, the
programmatic controls and interfaces
must also be described for different
modular configurations, as required by
§ 53.440(i), including any restrictions
that will be necessary during the
construction and startup of any given
unit to ensure the safe operation of the
overall commercial nuclear plant to be
licensed under this part.
(8) Programmatic Controls for Normal
Operations. A description of how
programmatic controls, including
monitoring programs, would provide
assurance that design features and
procedures will enable the holder of an
OL or COL to comply with § 53.260.
(9) Design Features Supporting the
Protection of Plant Workers. A
description of the design features
required by § 53.430 to support the
holder of an OL or COL complying with
§ 53.270.
(10) Programmatic Controls for
Protection of Plant Workers. A
description of how programmatic
controls, including monitoring
programs, would provide assurance that
design features and procedures will
enable the holder of an OL or COL to
comply with § 53.270.
(11) Codes and Standards. A
description of generally accepted
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consensus codes and standards used to
design the design features, as required
by § 53.440(b).
(12) Materials. A description of the
materials used for SR and NSRSS SSCs
and a description of the qualification of
these materials for their service
conditions over the plant lifetime, as
required by § 53.440(c).
(13) Integrity Assessment Program. A
description of a design integrity
assessment program that addresses the
elements described in § 53.440(d).
(14) Safety and Security. Confirmation
that safety and security were considered
together in the design process, as
required by § 53.440(f).
(15) Criticality. Information
demonstrating how the applicant will
comply with requirements for criticality
accidents in § 53.440(m).
(16) Multi-unit Plants. For an
application for standard design
certification of a multi-unit commercial
nuclear plant, the possible operating
configurations of the reactor units,
including common systems, interface
requirements, and system interactions,
as required by § 53.440(i).
(17) SSC Classification. (i) The
classification of SSCs according to their
safety significance under § 53.460(a).
(ii) For SR and NSRSS SSCs, the
conditions under which they must
perform the safety functions required by
§ 53.230, including environmental
conditions.
(18) Probabilistic Risk Assessment. A
description of the probabilistic risk
assessment (PRA) required by
§ 53.450(a) and its results.
(19) Analyses. A description of the
analyses performed under § 53.450(b)
through (g) that includes the following
information:
(i) A description of the analysis of
LBEs and its results, as described in
§ 53.240. This analysis description
must—
(A) Address the elements in
§ 53.450(e) and (f); and
(B) Under § 53.460(c)—
(1) Describe any human actions that
are necessary to prevent or mitigate
LBEs;
(2) Describe how those human actions
are capable of being reliably performed
under the postulated environmental
conditions present; and
(3) Describe how those human actions
would be addressed by programs
established under subpart F of this part.
(ii)(A) A description of how SSCs
relied on to meet the safety criteria
defined in § 53.210 are protected against
or designed to withstand the effects of
external hazards under § 53.510.
(B) The information necessary to
demonstrate that the commercial
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nuclear plant complies with the
earthquake engineering criteria in
§ 53.480.
(iii) A description of the defense-indepth measures required by § 53.250.
(iv) A description of all plant
operating states where there is the
potential for the uncontrolled release of
radioactive material to the environment,
as required by § 53.450(b)(4).
(v) A description of the events that
challenge plant control and safety
systems whose failure could lead to an
undesirable end state and/or radioactive
material release, as required by
§ 53.450(b)(5).
(vi) A description of the analytical
codes used in modeling plant behavior
in analyses of LBEs and how these
codes are qualified for the range of
conditions for which they were used, as
required by § 53.450(d).
(vii) If not described in addressing
paragraph (5) of this section, the results
of other analyses required by
§ 53.450(g).
(20) Special Treatments. A
description of special treatments
established as required by § 53.460.
(21) Analytical Margins. A description
of any alternative criteria adopted to
demonstrate analytical margins
supporting operational flexibilities, if
applicable, as required by § 53.470.
(22) Quality Assurance. A description
of the QAP applied to the design of the
SSCs of the commercial nuclear plant,
as required by § 53.460(b). The
description of the QAP for a commercial
nuclear plant must include a discussion
of how the applicable requirements of
appendix B to part 50 of this chapter
were satisfied.
(23) Design Features and Controls to
Address the Minimization of
Contamination. The information
required by § 20.1406 of this chapter.
(24) Interface Requirements. (i) A
description analysis, and evaluation of
the interfaces between the standard
design and the balance of the
commercial nuclear plant that may
impact the ability of the plant to
demonstrate compliance with the
functional design criteria or the safety
criteria of subparts B and C of this part.
(ii) Confirmation that interface
requirements are verifiable through
inspections, testing, or analysis. These
requirements must be sufficiently
detailed to allow for completion of the
final safety analysis by license
applicants that reference the certified
design under this subpart. The method
to be used for verification of interface
requirements must be included as part
of the proposed ITAAC required by
§ 53.1241(a)(3).
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(iii) A representative conceptual
design for those portions of the plant for
which the application does not seek
certification to aid the NRC in its review
of the FSAR and to permit assessment
of the adequacy of the interface
requirements under paragraph (a)(24)(i)
of this section.
(25) Technical Qualifications. A
description of the technical
qualifications of the applicant to engage
in the proposed activities in accordance
with the regulations in this chapter.
(26) Technical Specifications.
Proposed technical specifications
prepared under § 53.710(a) for those
areas addressed by the design
certification.
(27) Role of personnel. Information to
address the following for the role of
personnel in ensuring safe operations:
(i) A description of how the human
factors engineering design requirements
of § 53.440(n)(1) are addressed;
(ii) A description of how the human
system interface design requirements of
§ 53.440(n)(2) are addressed;
(iii) A concept of operations that is of
sufficient scope and detail to address
the requirements of § 53.440(n)(3);
(iv) A functional requirements
analysis and function allocation that is
of sufficient scope and detail to address
the requirements of § 53.440(n)(4).
(b) [Reserved]
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§ 53.1241 Contents of applications for
standard design certifications; other
application content.
(a) In addition to the FSAR, the
application must also include the
following:
(1) Environmental report. An
environmental report as required by
§ 51.55 of this chapter.
(2) Availability Controls (if not
included in the FSAR). A description of
the controls on plant operations,
including availability controls, to
provide reasonable confidence that the
configurations and special treatments
for SR and NSRSS SSCs provide the
capabilities and reliabilities required to
demonstrate compliance with the safety
criteria of § 53.220, or more restrictive
alternative criteria adopted under
§ 53.470.
(3) Inspections, tests, analyses, and
acceptance criteria. The proposed
ITAAC that are necessary and sufficient
to provide reasonable assurance that, if
the inspections, tests, and analyses are
performed and the acceptance criteria
met, a facility that incorporates the
design certification has been
constructed and will be operated in
conformity with the design certification,
the provisions of the Act, and the
Commission’s rules and regulations.
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(4) Safeguards information. A
description of the program to protect
Safeguards Information against
unauthorized disclosure in accordance
with the requirements in §§ 73.21 and
73.22 of this chapter, as applicable.
(b) If there are SSCs of the plant
which required research and
development to confirm the adequacy of
their design, provide a report in the
application which documents the
resolution of any safety questions
associated with such SSCs.
(c) A description of how the
performance of each design feature has
been demonstrated capable of fulfilling
functional design criteria considering
interdependent effects through either
analysis, appropriate test programs,
prototype testing, operating experience,
or a combination thereof, in accordance
with § 53.440(a).
§ 53.1242
Review of applications.
(a) Standards for review of
applications. Applications filed under
this part will be reviewed for
compliance with the standards set out
in this part and 10 CFR parts 20, 51, and
73.
(b) Administrative review of
applications; hearings. (1) A standard
design certification is a rule that will be
issued under the provisions of subpart
H of 10 CFR part 2, as supplemented by
the provisions of this section. The
Commission must initiate the
rulemaking after an application has
been filed under § 53.1233 and must
specify the procedures to be used for the
rulemaking. The notice of proposed
rulemaking published in the Federal
Register must provide an opportunity
for the submission of comments on the
proposed design certification rule. If, at
the time a proposed design certification
rule is published in the Federal Register
under this paragraph, the Commission
decides that a legislative hearing should
be held, the information required by
§ 2.1502(c) of this chapter must be
included in the Federal Register
document for the proposed design
certification.
(2) Following the submission of
comments on the proposed design
certification rule, the Commission may,
at its discretion, hold a legislative
hearing under the procedures in subpart
O of part 2 of this chapter. The
Commission must publish a document
in the Federal Register of its decision to
hold a legislative hearing. The
document must contain the information
specified in § 2.1502(c) of this chapter
and specify whether the Commission or
a presiding officer will conduct the
legislative hearing.
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(3) Notwithstanding anything in
§ 2.390 of this chapter to the contrary,
proprietary information will be
protected in the same manner and to the
same extent as proprietary information
submitted in connection with
applications for licenses, provided that
the design certification will be
published in chapter I of this title.
(c) Reference to an issued operating
license or combined license. In those
cases where a design certification
application is preceded by the issuance
of an OL or custom COL for a
commercial nuclear plant that is
essentially the same as the standard
design for which certification is being
requested, the NRC review will follow
the processes for referencing a standard
design approval in § 53.1221, to the
extent practicable.
§ 53.1245 Referral to the Advisory
Committee on Reactor Safeguards.
The Commission must refer a copy of
the application to the ACRS. The ACRS
must report on those portions of the
application which concern safety.
§ 53.1248 Issuance of standard design
certification.
(a) After conducting a rulemaking
proceeding under § 53.1242 on an
application for a standard design
certification and receiving the report to
be submitted by the ACRS under
§ 53.1245, the Commission may issue a
standard design certification in the form
of a rule for the design that is the subject
of the application, if the Commission
determines that—
(1) The application demonstrates
compliance with the applicable
standards and requirements of the Act
and the Commission’s regulations;
(2) Notifications, if any, to other
agencies or bodies have been duly
made;
(3) There is reasonable assurance that
the standard design conforms with the
provisions of the Act and the
Commission’s regulations;
(4) The applicant is technically
qualified;
(5) The proposed ITAAC are
necessary and sufficient, within the
scope of the standard design, to provide
reasonable assurance that, if the
inspections, tests, and analyses are
performed and the acceptance criteria
met, the facility has been constructed
and will be operated in accordance with
the design certification, the provisions
of the Act, and the Commission’s
regulations;
(6) Issuance of the standard design
certification will not be inimical to the
common defense and security or to the
health and safety of the public;
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(7) The findings required by subpart
A of part 51 of this chapter have been
made; and
(8) The applicant has implemented
the QAP described or referenced in the
Safety Analysis Report.
(b) The design certification rule must
specify the site parameters, design
characteristics, and any additional
requirements and restrictions of the
design certification rule.
(c) After the Commission has adopted
a final design certification rule, the
applicant must not permit any
individual to have access to or any
facility to possess restricted data or
classified National Security Information
until the individual and/or facility has
been approved for access under the
provisions of 10 CFR parts 25 and/or 95,
as applicable.
§ 53.1251
Duration of certification.
(a) Except as provided in paragraph
(b) of this section, a standard design
certification issued under this subpart is
valid for 15 years from the effective date
of the rule.
(b) A standard design certification
continues to be valid beyond the date of
expiration in any proceeding on an
application for a COL or an OL under
this part that references the standard
design certification and is docketed
either before the date of expiration of
the certification, or, if a timely
application for renewal of the
certification has been filed, before the
Commission has determined whether to
renew the certification. A design
certification also continues to be valid
beyond the date of expiration in any
hearing held under § 53.1452 before
operation begins under a COL that
references the design certification.
(c) An applicant for a CP, OL, COL,
or ML under this part may, at its own
risk, reference in its application a design
for which a design certification
application has been docketed but not
granted.
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§ 53.1254
Application for renewal.
(a) Not less than 12 nor more than 36
months before the expiration of the
initial 15-year period, or any later
renewal period, any person may apply
for renewal of the certification. An
application for renewal must contain all
information necessary to bring up to
date the information and data contained
in the previous application. The
Commission will require, before
renewal of certification, that engineering
documents, such as analyses, drawings,
procurement specifications, or
construction and installation
specifications, be completed and
available for audit if the more detailed
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information is necessary for the
Commission to verify the information in
the application and make its safety
determination. Notice and comment
procedures must be used for a
rulemaking proceeding on the
application for renewal. The
Commission, in its discretion, may
require the use of additional procedures
in individual renewal proceedings.
(b) A design certification, either
original or renewed, for which a timely
application for renewal has been filed
remains in effect until the Commission
has determined whether to renew the
certification. If the certification is not
renewed, it continues to be valid in
certain proceedings under § 53.1251.
(c) The Commission must refer a copy
of the application for renewal to the
ACRS. The ACRS must report on those
portions of the application which
concern safety and must apply the
criteria set forth in § 53.1257.
§ 53.1257
Criteria for renewal.
(a) The Commission must issue a rule
granting the renewal if the design, either
as originally certified or as modified
during the rulemaking on the renewal,
complies with the Act and the
Commission’s regulations applicable
and in effect at the time the certification
was issued.
(b) The Commission may impose
other requirements if it determines
that—
(1) They are necessary for adequate
protection to public health and safety or
common defense and security;
(2) They are necessary for compliance
with the Commission’s regulations and
orders applicable and in effect at the
time the design certification was issued;
or
(3) There is a substantial increase in
overall protection of the public health
and safety or the common defense and
security to be derived from the new
requirements, and the direct and
indirect costs of implementing those
requirements are justified in view of this
increased protection.
(c) In addition, the applicant for
renewal may request an amendment to
the design certification. The
Commission must grant the amendment
request if it determines that the
amendment will comply with the Act
and the Commission’s regulations in
effect at the time of renewal. If the
amendment request entails such an
extensive change to the design
certification that an essentially new
standard design is being proposed, an
application for a design certification
must be filed in accordance with this
subpart.
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(d) Denial of renewal does not bar the
applicant, or another applicant, from
filing a new application for certification
of the design, which proposes design
changes that correct the deficiencies
cited in the denial of the renewal.
§ 53.1260
Duration of renewal.
Each renewal of certification for a
standard design will be for not less than
10, nor more than 15 years.
§ 53.1263 Finality of standard design
certifications.
(a)(1) While a standard design
certification rule is in effect under
§§ 53.1251 or 53.1260, the Commission
may not modify, rescind, or impose new
requirements on the certification
information, whether on its own
motion, or in response to a petition from
any person, unless the Commission
determines in a rulemaking that the
change—
(i) Is necessary either to bring the
certification information or the
referencing plants into compliance with
the Commission’s regulations applicable
and in effect at the time the certification
was issued;
(ii) Is necessary to provide adequate
protection of the public health and
safety or the common defense and
security;
(iii) Reduces unnecessary regulatory
burden and maintains protection to
public health and safety and the
common defense and security;
(iv) Provides the detailed design
information to be verified under those
ITAAC that are directed at certification
information (i.e., design acceptance
criteria);
(v) Is necessary to correct material
errors in the certification information;
(vi) Substantially increases overall
safety, reliability, or security of facility
design, construction, or operation, and
the direct and indirect costs of
implementation of the rule change are
justified in view of this increased safety,
reliability, or security; or
(vii) Contributes to increased
standardization of the certification
information.
(2)(i) In a rulemaking under
§ 53.1263(a)(1), except for
§ 53.1263(a)(1)(ii), the Commission will
give consideration to whether the
benefits justify the costs for plants that
are already licensed or for which an
application for a permit or license is
under consideration.
(ii) The rulemaking procedures for
changes under § 53.1263(a)(1) must
provide for notice and opportunity for
public comment.
(3) Any modification the NRC
imposes on a design certification rule
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under paragraph (a)(1) of this section
will be applied to all plants referencing
the certified design, except those to
which the modification has been
rendered technically irrelevant by
action taken under paragraphs (a)(4) or
(b) of this section.
(4) The Commission may not impose
new requirements by plant-specific
order on any part of the design of a
specific plant referencing the design
certification rule if that part was
approved in the design certification
while a design certification rule is in
effect under § 53.1248, unless—
(i) A modification is necessary to
secure compliance with the
Commission’s regulations applicable
and in effect at the time the certification
was issued, or to assure adequate
protection of the public health and
safety or the common defense and
security; and
(ii) Special circumstances as defined
in § 53.080 are present. In addition to
the factors listed in § 53.080, the
Commission must consider whether the
special circumstances which § 53.080
requires to be present outweigh any
decrease in safety that may result from
the reduction in standardization caused
by the plant-specific order.
(5) Except as provided in § 2.335 of
this chapter, in making the findings
required for issuance of a COL, CP, OL,
or ML, or for any hearing under
§ 53.1452, the Commission must treat as
resolved those matters resolved in
connection with the issuance or renewal
of a design certification rule.
(b) An applicant who references a
design certification rule may request an
exemption from one or more elements of
the certification information. The
Commission may grant such a request
only if it determines that the exemption
will comply with the requirements of
§ 53.080. In addition to the factors listed
in § 53.080, the Commission must
consider whether the special
circumstances that § 53.080 requires to
be present outweigh any decrease in
safety that may result from the
reduction in standardization caused by
the exemption. The granting of an
exemption on request of an applicant is
subject to litigation in the same manner
as other issues in the OL or COL
hearing.
(c) The Commission will require,
before granting a CP, COL, OL, or ML
that references a design certification
rule, that information normally
contained in engineering documents,
such as analyses, drawings,
procurement specifications, or
construction and installation
specifications, be completed and
available for audit if the more detailed
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information is necessary for the
Commission to verify the information in
the application and make its safety
determination, including the
determination that the application is
consistent with the certification
information. This information may be
acquired by appropriate arrangements
with the design certification applicant.
§ 53.1270
Manufacturing licenses.
Sections 53.1270 through 53.1295 set
out the requirements and procedures
applicable to Commission issuance of a
license under this part authorizing
manufacture of manufactured reactors to
be installed at sites not identified in the
ML application.
§ 53.1273
Filing of applications.
Any person, except one excluded by
§ 53.1118, may file an application for an
ML under this part with the Director,
Office of Nuclear Reactor Regulation.
§ 53.1276 Contents of applications for
manufacturing licenses; general
information.
Each application for an ML must
include the information contained in
§ 53.1109(a) through (e), and (j).
§ 53.1279 Contents of applications for
manufacturing licenses; technical
information.
(a) Final Safety Analysis Report-siting
and design. The application must
include an FSAR containing the
information set forth below, with a level
of design information sufficient to
enable the Commission to judge the
applicant’s proposed means of ensuring
that the manufacturing conforms to the
design and to reach a final conclusion
on all safety questions associated with
the design, permit the preparation of
construction and installation
specifications by an applicant who
seeks to use the manufactured reactor,
and permit the preparation of
acceptance and inspection requirements
by the NRC. The application must
include the following information:
(1) Site parameters. The site
parameters postulated for the design
under this part, including the designbasis external hazard levels for the
relevant external hazards, and an
analysis and evaluation of the design in
terms of those site parameters.
(2) Design information. Except as
specified in this paragraph, the design
information equivalent to that required
for a standard design certification as
defined in § 53.1239(a)(2) through (27)
for those portions of a commercial
nuclear plant included in the
manufactured reactor.
(3) Quality assurance program. A
description of the QAP applied to the
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design, and to be applied to the
fabrication and testing of the SSCs of the
manufactured reactor under
§ 53.620(a)(6), including a discussion of
how the applicable requirements of
appendix B to part 50 of this chapter
will be satisfied;
(4) Conceptual designs.
Representative conceptual designs for
one or more commercial nuclear plants
using the manufactured reactor;
(5) Operating configurations. If
multiple manufactured reactors may be
installed at a commercial nuclear plant,
a description of the possible operating
configurations, including common
systems, interface requirements, and
system interactions. The final safety
analysis must also account for
differences among the possible
configurations, including any
restrictions that will be necessary
during the construction and startup of a
given manufactured reactor to ensure
the safe operation of any commercial
nuclear reactor already operating;
(6) Interface requirements. (i) The
interface requirements between the
manufactured reactor and the remaining
portions of the commercial nuclear
plant or connections to other facilities
outside of the commercial nuclear plant.
(ii) Confirmation that interface
requirements are verifiable through
inspections, testing, or analysis. These
requirements must be sufficiently
detailed to allow for completion of the
final safety analysis by license
applicants that reference the
manufactured reactor manufactured
under this subpart. Applicants for a
COL under this part will need to verify
the interface requirements at the
installation site. The method to be used
for verification of interface requirements
must be included as part of the
proposed ITAAC required by
§ 53.1282(a).
(iii) Information to support
development of radiation monitoring
programs required under subpart F of
this part by an applicant for a COL,
including potential pathways for
radionuclides produced within the
manufactured reactor to enter
interfacing systems.
(b) Final Safety Analysis Report—
manufacturing information. The FSAR
must include the following information
related to the manufacturing processes,
organization, controls, and inspections:
(1) A description, including
references to generally accepted
consensus codes and standards, of the
processes that will be used to procure,
fabricate, and assemble components that
make up the manufactured reactor. The
description should clearly define which
activities are proposed to be within the
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scope of the ML and those, such as the
making of a component to be procured
from a separate company for installation
in the manufactured reactor, that are not
considered to be within the scope of the
ML;
(2) A description of the organizational
and management structure singularly
responsible for direction of design and
manufacture of the manufactured
reactor. The information should include
a description of the management plans,
technical qualifications, and controls in
place to demonstrate compliance with
the requirements of § 53.620;
(3) A description of the inspections
and tests to be performed as part of the
manufacturing process, including the
inspection of procured components,
inspection and testing of fabrication
processes such as the molding, welding,
or coating of components, and
inspections and testing of the assembled
manufactured reactor or portions of the
manufactured reactor;
(4) A description of the fitness-forduty program required by part 26 of this
chapter and its implementation.
(c) Deployment of the completed
manufactured reactor. The application
must include the following information
related to the deployment of a
manufactured reactor:
(1) Procedures governing the
preparation of the manufactured reactor
or portions of the manufactured reactor
for shipping to the site where it is to be
operated; the conduct of shipping; and
verifying the condition of the shipped
items upon receipt at the site;
(2) Details of the interaction of the
design, manufacture, and installation of
a manufactured reactor within the
applicant’s organization and the manner
by which the applicant will ensure close
integration between the designer,
contractors, and any facility in which
the manufactured reactor is to be
installed;
(3) A description of the measures to
be used for the control of interfaces,
including the consideration of key site
parameters, between the holder of the
ML and the holder of the COL for the
commercial nuclear plant at which the
manufactured reactor is to be installed.
(d) Special considerations for factory
fueling. In addition to the above
paragraphs, an application for an ML for
a manufactured reactor that will be
fueled at the factory under a 10 CFR part
70 license must include the following
information related to loading fuel and
the required independent physical
mechanisms to prevent criticality and to
otherwise provide assurance that the
fueled manufactured reactor can be
successfully transported, installed, and
operated at a site for which the
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Commission has issued a COL that
authorizes construction and operation of
a commercial nuclear plant using the
manufactured reactor:
(1) A description of the procedures
used during the fueling of the
manufactured reactor that ensure that
the configuration of fuel within the
fueled manufactured reactor is
consistent with the design and analyses
supporting operation of the
manufactured reactor under the COL at
the place of operation. The description
may reference the applicable 10 CFR
part 70 application and other sections of
the Safety Analysis Report supporting
the ML license application.
(i) The application must describe the
measures taken for in-factory
inspections and non-nuclear testing
performed to ensure that the
configuration of fuel within the fueled
manufactured reactor is consistent with
the design and analyses supporting
operation of the manufactured reactor
under the COL at the place of operation.
(ii) The application must describe the
design features included in the
manufactured reactor to prevent
criticality, including at least two
independent mechanisms each of which
is sufficient to prevent criticality, the
associated functional design criteria
applied to those design features, and the
physical and programmatic controls
implemented during manufacturing,
storage, and transport that are credited
to assure the features function as
designed when subject to potential
hazards and human errors. The
descriptions must include how those
measures will be controlled during
installation under the ML and removal
under the COL at the place of operation.
(2) A description of the procedures
governing the transfer of responsibilities
for the fueled manufactured reactor
from the holder of the ML to the holder
of the COL for the installation site.
(3) If available at the time of filing the
ML application or, if not available at the
time of filing the ML application,
submitted as an amendment to the ML
or ML application at the time of filing
the Part 70 application, a description of
the programs needed to demonstrate
compliance with the requirements of
§ 53.620(d) and 10 CFR parts 70, 71, and
73 for the receipt, storage, and loading
of SNM into a manufactured reactor and
the transport of the fueled manufactured
reactor to a site for which the
Commission has issued a COL that
authorizes construction and operation of
a commercial nuclear plant using the
manufactured reactor, including the
following.
(i) A physical security program in
accordance with § 53.620(d)(2)(i).
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(ii) A cybersecurity program in
accordance with § 53.620(d)(2)(i).
§ 53.1282 Contents of applications for
manufacturing licenses; other application
content.
(a) Inspections, tests, analyses, and
acceptance criteria. (1) The application
must contain proposed inspections,
tests, and analyses that the COL holder
must perform, and the acceptance
criteria that are necessary and sufficient
to provide reasonable assurance that, if
the inspections, tests, and analyses are
performed and the acceptance criteria
met:
(i) The reactor has been manufactured
in conformity with the ML, the
provisions of the Act, and the
Commission’s rules and regulations; and
(ii) The manufactured reactor will be
operated in conformity with the
approved design and any license
authorizing operation of the
manufactured reactor.
(2) If the application references a
standard design certification, the ITAAC
contained in the certified design must
apply to those portions of the facility
design that are covered by the design
certification.
(3) If the application references a
standard design certification, the
application may include a notification
that a required inspection, test, or
analysis in the design certification
ITAAC has been successfully completed
and that the corresponding acceptance
criterion has been met. The Federal
Register notification required by
§ 53.1285 must indicate that the
application includes this notification.
(b) Environmental report. (1) The
application must contain an
environmental report as required by
§ 51.54 of this chapter.
(2) If the ML application references a
standard design certification, the
environmental report need not contain a
discussion of severe accident mitigation
design alternatives for the manufactured
reactor as used in a commercial nuclear
plant.
(c) Safeguards information. The
application must contain a description
of the program to protect safeguards
information against unauthorized
disclosure in accordance with the
requirements in §§ 73.21 and 73.22 of
this chapter, as applicable.
(d) Performance demonstration. A
description of how the performance of
each design feature has been
demonstrated capable of fulfilling
functional design criteria considering
interdependent effects through either
analysis, appropriate test programs,
prototype testing, operating experience,
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or a combination thereof, in accordance
with § 53.440(a).
§ 53.1285
Review of applications.
(a) Standards for review of
applications. Applications for MLs
under this part will be reviewed
according to the applicable standards
set out in this subpart as well as
applicable standards in this part and 10
CFR parts 20, 25, 26, 51, 70, 71, 73, and
75.
(b) Administrative review of
applications, hearings. A proceeding on
an ML is subject to all applicable
procedural requirements contained in
10 CFR part 2, including the
requirements for docketing in
§ 2.101(a)(1) through (4) of this chapter,
and the requirements for issuance of a
notice of proposed action in § 2.105 of
this chapter, provided, however, that the
designated sections may not be
construed to require that the
environmental report or draft or final
environmental impact statement include
an assessment of the benefits of
constructing and/or operating the
manufactured reactor or an evaluation
of alternative energy sources. All
hearings on MLs are governed by the
hearing procedures contained in 10 CFR
part 2, subparts C, E, G, L, and N.
§ 53.1286 Referral to the Advisory
Committee on Reactor Safeguards.
The Commission must refer a copy of
the application to the ACRS. The ACRS
must report on those portions of the
application which concern safety.
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§ 53.1287
licenses.
Issuance of manufacturing
(a) After completing any hearing
under § 53.1285(b), and receiving the
report submitted by the ACRS, the
Commission may issue an ML if the
Commission finds that—
(1) Applicable standards and
requirements of the Act and the
Commission’s regulations have been
met;
(2) There is reasonable assurance that
the manufactured reactor will be
manufactured, and can be transported,
incorporated into a commercial nuclear
plant, and operated in conformity with
the ML, the provision of the Act, and
the Commission’s regulations;
(3) The proposed manufactured
reactor can be incorporated into a
commercial nuclear plant and operated
at sites having characteristics that fall
within the site parameters postulated for
the design of the manufactured reactors
without undue risk to the health and
safety of the public;
(4) The applicant is technically
qualified to design and manufacture the
proposed manufactured reactor;
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(5) The proposed ITAAC are
necessary and sufficient, within the
scope of the ML, to provide reasonable
assurance that the manufactured reactor
has been manufactured and will be
operated in conformity with the license,
the provisions of the Act, and the
Commission’s regulations;
(6) The issuance of a license to the
applicant will not be inimical to the
common defense and security or to the
health and safety of the public; and
(7) The findings required by subpart
A of 10 CFR part 51 have been made.
(b) Each ML issued under this subpart
must specify—
(1) Terms and conditions as the
Commission deems necessary and
appropriate;
(2) Technical specifications for
operation of the manufactured reactor,
as the Commission deems necessary and
appropriate;
(3) Site parameters and design
characteristics for the manufactured
reactor;
(4) The interface requirements to be
met by the site-specific elements of the
facility, such as the energy conversions
systems and ultimate heat sink, not
within the scope of the manufactured
reactor; and
(5) The entity with design authority
for the manufactured reactor covered by
the license.
§ 53.1288
licenses.
Finality of manufacturing
(a)(1) Notwithstanding any provision
in § 53.1590, during the term of an ML
issued under this part the Commission
may not modify, rescind, or impose new
requirements on the design of the
manufactured reactor, or the
requirements for the manufacture of the
manufactured reactor, unless the
Commission determines that a
modification is necessary to bring the
design of the reactor or its manufacture
into compliance with the Commission’s
requirements applicable and in effect at
the time the ML was issued, or to
provide reasonable assurance of
adequate protection to public health and
safety or common defense and security.
(2) Any modification to the design of
a manufactured reactor that is imposed
by the Commission under paragraph
(a)(1) of this section will be applied to
all manufactured reactors manufactured
under the license, including those that
have already been transported and sited,
except those manufactured reactors to
which the modification has been
rendered technically irrelevant by
action taken under § 53.1530 or
paragraph (b) of this section.
(3) In making the findings required
under this part for issuance of a COL,
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in any hearing under § 53.1452, or in
any enforcement hearing other than one
initiated by the Commission under
paragraph (a)(1) of this section, for
which a manufactured reactor
manufactured under this subpart is
referenced or used, the Commission
must treat as resolved those matters
resolved in the proceeding on the
application for issuance or renewal of
the ML, including the adequacy of
design of the manufactured reactor, the
costs and benefits of severe accident
mitigation design alternatives, and the
bases for not incorporating severe
accident mitigation design alternatives
into the design of the manufactured
reactor to be manufactured.
(b) An applicant who references or
uses a manufactured reactor
manufactured under an ML under this
part may include in the application a
request for a departure from the design
characteristics, site parameters, terms
and conditions, or approved design of
the manufactured reactor. The
Commission may grant a request only if
it determines that the departure will
comply with the requirements of
§ 53.080, and that the special
circumstances outweigh any decrease in
safety that may result from the
reduction in standardization caused by
the departure. The granting of a
departure on request of an applicant is
subject to litigation in the same manner
as other issues in the COL hearing.
§ 53.1291
licenses.
Duration of manufacturing
An ML issued under this part is valid
for not less than 5, nor more than 15
years from the date of issuance. Upon
expiration of the ML, the manufacture of
any uncompleted manufactured reactors
must cease unless a timely application
for renewal has been docketed with the
NRC.
§ 53.1293
licenses.
Transfer of manufacturing
An ML may be transferred under
§ 53.1570.
§ 53.1295
licenses.
Renewal of manufacturing
(a)(1) Not less than 12 months, nor
more than 5 years before the expiration
of the ML, or any later renewal period,
the holder of the ML issued under this
part may apply for a renewal of the
license. An application for renewal
must contain all information necessary
to bring up to date the information and
data contained in the previous
application.
(2) The filing of an application for a
renewed license must be in accordance
with subpart A of 10 CFR part 2 and
§ 53.1100.
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(3) An ML issued under this part,
either original or renewed, for which a
timely application for renewal has been
filed, remains in effect until the
Commission has made a final
determination on the renewal
application, provided, however, that the
holder of an ML may not begin
manufacture of a manufactured reactor
less than 6 months before the expiration
of the license.
(4) Any person whose interest may be
affected by renewal of the license may
request a hearing on the application for
renewal. The request for a hearing must
comply with § 2.309 of this chapter. If
a hearing is granted, notice of the
hearing will be published in accordance
with § 2.104 of this chapter.
(5) The Commission must refer a copy
of the application for renewal to the
ACRS. The ACRS must report on those
portions of the application which
concern safety.
(b) The Commission may grant the
renewal if the Commission
determines—
(1) The ML complies with the Act and
the Commission’s regulations and
orders applicable and in effect at the
time the ML was originally issued; and
(2) Any new requirements the
Commission may wish to impose are—
(i) Necessary for adequate protection
to public health and safety or common
defense and security;
(ii) Necessary for compliance with the
Commission’s regulations and orders
applicable and in effect at the time the
ML was originally issued; or
(iii) A substantial increase in overall
protection of the public health and
safety or the common defense and
security to be derived from the new
requirements, and the direct and
indirect costs of implementation of
those requirements are justified in view
of this increased protection.
(c) A renewed ML may be issued for
a term of not less than 5, nor more than
15 years, plus any remaining years on
the ML then in effect before renewal.
The renewed license must be subject to
the requirements of § 53.1288.
An application for a CP must include
the information required by § 53.1109
and the following information:
(a) Information sufficient to
demonstrate to the Commission the
financial qualification of the applicant
to carry out, under the regulations in
this chapter, the activities for which the
permit is sought. As applicable, the
following should be provided:
(1) The information that demonstrates
that the applicant possesses or has
reasonable assurance of obtaining the
funds necessary to cover estimated
construction costs and related fuel cycle
costs, including estimates of the total
construction costs and related fuel cycle
costs of the facility and must indicate
the source(s) of funds to cover these
costs.
(2) Each application for a CP
submitted by a newly formed entity
organized for the primary purpose of
constructing and operating a facility
must also include information showing:
(i) The legal and financial
relationships the entity has or proposes
to have with its stockholders or owners;
(ii) The stockholders’ or owners’
financial ability to meet any contractual
obligation to the entity that they have
incurred or proposed to incur; and
(iii) Any other information considered
necessary by the Commission to enable
it to determine the applicant’s financial
qualification; and
(3) The Commission may request an
established entity or newly-formed
entity to submit additional or more
detailed information respecting its
financial arrangements and status of
funds if the Commission considers this
information appropriate. This may
include information regarding an
applicant’s ability to continue the
conduct of the activities authorized by
the CP and to decommission the facility.
(b) If the applicant proposes to
construct or alter a facility, the
application must state the earliest and
latest dates for completion of the
construction or alteration.
Construction permits.
§ 53.1309 Contents of applications for
construction permits; technical information.
Sections 53.1300 through 53.1348 set
out the requirements and procedures
applicable to Commission issuance of a
CP for commercial nuclear plants. A CP
for the construction of a commercial
nuclear plant under this part will be
issued before the issuance of an OL if
the application is otherwise acceptable
and will be converted upon completion
of the facility and Commission action,
into an OL as provided under
§§ 53.1360 through 53.1405.
The application must contain a
Preliminary Safety Analysis Report
(PSAR) that describes the facility and
the limits on its operation and presents
a preliminary safety analysis of the SSCs
of the facility as a whole. The PSAR
must include the following information,
at a level of detail sufficient to enable
the Commission to reach a conclusion
on safety matters that must be resolved
by the Commission before issuance of a
CP:
§ 53.1300
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§ 53.1306 Contents of applications for
construction permits; general information.
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(a)(1) Site information. An application
for a CP for a commercial nuclear
reactor must include the site
information equivalent to that required
for an early site permit in
§ 53.1146(a)(1)(iv) through (x).
(2) Design information. Except as
specified in this paragraph, an
application for a CP for a commercial
nuclear plant must include the design
information equivalent to that required
for a standard design certification as
defined in § 53.1239(a)(2) through (27).
(i) Quality assurance program. A
description of the QAP to be applied to
the design, fabrication, construction,
and testing of the SSCs of the facility
under § 53.610(a)(6), including a
discussion of how the requirements of
appendix B to part 50 of this chapter
will be satisfied.
(ii) Preliminary design information.
The information provided in the
application may include some aspects of
the design that are not fully developed,
and the information is therefore
preliminary. The completed design,
including any changes during
construction, must be described in the
FSAR required in § 53.1369 that
supports an application for an OL.
(iii) Planned research or testing.
Descriptions of how design features and
related functional design criteria will
fulfill the safety criteria in subpart B, or
more restrictive alternative criteria
adopted under § 53.470, and how that
has been or will be demonstrated
through either analysis, appropriate test
programs, experience, or a combination
thereof. Where any design feature has
not been fully developed or
demonstrated to fulfill the functional
design criteria at the time of an
application for a CP, the applicant must
provide a plan for future analysis,
research and development, test
programs, gathering of experience, or a
combination thereof to provide
reasonable confidence that the required
demonstration will be available for an
application for an OL.
(iv) Programmatic controls.
Descriptions of the programmatic
controls may include those to be
provided in the FSAR or other licensing
basis documents because they are
necessary to achieve and maintain the
reliability and capability of SSCs relied
upon to demonstrate compliance with
the established safety criteria and
functional design criteria required in
subpart B, and to maintain consistency
with analyses required by § 53.450.
(3) Technical qualifications. A
description of the technical
qualifications of the applicant to engage
in the proposed activities under the
regulations in this chapter.
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(4) Emergency preparedness. A
description of the applicant’s
preliminary plans for coping with
emergencies based on:
(i) Except as provided in paragraph
(a)(4)(ii) of this section, the
requirements in appendix E to part 50.
(ii) For a commercial nuclear plant
consisting of either small modular
reactors or non-light-water reactors, the
requirements in either § 50.160 or
appendix E to part 50.
(5) Physical security. A report that
provides a preliminary description of
how the site characteristics support the
development of adequate security plans
and measures consistent with the
requirements in § 53.540.
(6) Fitness-for-duty program. A
description of the fitness-for-duty (FFD)
program required by 10 CFR part 26 and
its implementation.
(b) A description of the program to
protect Safeguards Information against
unauthorized disclosure in accordance
with the requirements in §§ 73.21 and
73.22 of this chapter, as applicable.
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§ 53.1312 Contents of applications for
construction permits; other application
content.
(a) In addition to the PSAR, the
application must include the following:
(1) An environmental report either
under § 51.50(a) of this chapter if an
LWA under § 53.1130 is not requested
in conjunction with the CP application,
or under §§ 51.49 and 51.50(a) of this
chapter if an LWA is requested in
conjunction with the CP application; or
(2) If the applicant wishes to request
that an LWA under § 53.1130 be issued
before issuance of the CP, the
information otherwise required by
§ 53.1130, in accordance with either
§ 2.101(a)(1) through (a)(5), or
§ 2.101(a)(9) of this chapter.
(b) If the CP application references an
early site permit, standard design
approval, or standard design
certification issued under this part, then
the following requirements apply:
(1) The PSAR need not contain
information or analyses submitted to the
Commission in connection with the
referenced NRC approval, permit, or
certification, provided, however, that
the PSAR incorporates the material by
reference and confirms that the site and
design of the facility falls within
parameter values postulated in the
referenced NRC approval, permit, or
certification.
(2) The PSAR must provide a means
to demonstrate that all terms and
conditions that have been included in
the referenced NRC approval, permit, or
certification will be satisfied by the date
of issuance of the OL, as appropriate. If
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the PSAR does not demonstrate that
each site characteristic falls within the
corresponding postulated site parameter
and each design characteristic of the
facility falls within the corresponding
postulated design parameter, the
application must justify a departure,
variance, or exemption from the
referenced NRC approval, license, or
certification in regard to that particular
site or design characteristic in
compliance with the requirements of
this part.
(3) If a referenced early site permit
approves complete and integrated
emergency plans, or major features of
emergency plans, then the PSAR must
include any new or additional
information that updates and corrects
the information that was provided
under § 53.1146(b)(2) and discuss
whether the new or additional
information materially changes the
bases for compliance with the
applicable requirements.
§ 53.1315
Review of applications.
(a) Standards for review of
applications. Applications filed under
this part will be reviewed according to
the standards set out in this part and 10
CFR parts 20, 51, 73, and 140.
(b) Administrative review of
applications; hearings. A proceeding on
a CP application is subject to all
applicable procedural requirements
contained in 10 CFR part 2, including
the requirements for docketing (§ 2.101
of this chapter) and issuance of a notice
of hearing (§ 2.104 of this chapter). All
hearings on CP applications are
governed by the procedures contained
in 10 CFR part 2.
§ 53.1318 Finality of referenced NRC
approvals, permits, and certifications.
If the application for a CP under this
part references an early site permit,
standard design approval, or standard
design certification, the scope and
nature of matters resolved for the
application are governed by the relevant
provisions addressing finality, including
§§ 53.1188, 53.1221, and 53.1263.
§ 53.1324 Referral to the Advisory
Committee on Reactor Safeguards.
The Commission must refer a copy of
the application to the ACRS. The ACRS
must report on those portions of the
application that concern safety and
must apply the standards referenced in
§ 53.1315, in accordance with the
finality provisions in § 53.1318.
§ 53.1327 Authorization to conduct limited
work authorization activities.
(a) If the application does not
reference an early site permit which
authorizes the holder to perform the
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activities under § 53.1130, the applicant
may not perform those activities
without obtaining the separate
authorization required by § 53.1130.
Authorization may be granted only after
the presiding officer in the proceeding
on the application has made the
findings and determination required by
§ 53.1130(b)(1)(ii) and (iv), and the
Director, Office of Nuclear Reactor
Regulation makes the determination
required by § 53.1130(b)(1)(iii).
(b) If, after an applicant has performed
the activities permitted by paragraph (a)
of this section, the application for the
CP is withdrawn or denied, then the
applicant must implement an approved
site redress plan.
§ 53.1330 Exemptions, departures, and
variances.
(a) Applicants for a CP under this
part, or any amendment to a CP, may
include in the application a request for
an exemption from one or more of the
Commission’s regulations. The
Commission may grant a request if it
determines that the exemption complies
with § 53.080.
(b) An applicant for a CP who has
filed an application referencing an NRC
approval, permit, or certification issued
under this part may include in the
application a request for exemptions,
departures, or variances related to the
subject referenced NRC approval,
permit, or certification. In determining
whether to grant the departure,
variance, or exemption, the Commission
must apply the same technically
relevant criteria as were applicable to
the application for the original or
renewed approval, license, or
certification.
§ 53.1333
permits.
Issuance of construction
(a) After conducting a hearing in
accordance with § 53.1315 and receiving
the report submitted by the ACRS, the
Commission may issue a CP only if the
Commission finds that—
(1) The applicant has described the
proposed design of the facility and has
identified the major features or
components incorporated therein for the
protection of the health and safety of the
public;
(2) Such further technical or design
information as may be required to
complete the safety analysis, and which
can reasonably be left for later
consideration, will be supplied in the
FSAR;
(3) Safety features or components, if
any, that require research and
development have been described by
the applicant and the applicant has
identified, and there will be conducted,
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a research and development program
reasonably designed to resolve any
safety questions associated with such
features or components; and
(4) On the basis of the foregoing, there
is reasonable assurance of the
following—
(i) Such safety questions will be
satisfactorily resolved at or before the
latest date stated in the application for
completion of construction of the
proposed facility; and
(ii) Taking into consideration the site
criteria contained subpart D to this part,
the proposed facility can be constructed
and operated at the proposed location
without undue risk to the health and
safety of the public.
(b) A CP must contain the terms and
conditions for the permit, as the
Commission deems necessary and
appropriate. The Commission may, in
its discretion, incorporate in any CP
provisions requiring the applicant to
furnish periodic reports of the progress
and results of research and development
programs designed to resolve safety
questions.
§ 53.1336
Finality of construction permits.
Notwithstanding any provision in
§ 53.1590, a CP constitutes an
authorization to proceed with
construction but does not constitute
Commission approval of the safety of
any design feature or specification
unless the applicant specifically
requests such approval and such
approval is incorporated in the permit.
The applicant, at its option, may request
such approvals in the CP or by
amendment to the CP. If approved by
the NRC and included in the permit, the
NRC will consider modifications to the
approved design features or
specifications in accordance with
§ 53.1590.
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§ 53.1342
permits.
Duration of construction
(a) A CP will state the earliest and
latest dates for completion of
construction or alteration of the facility,
not to exceed 40 years from date of
issuance.
(b) If the proposed construction or
alteration of the facility is not
completed by the latest completion date,
the CP shall expire, and all rights are
forfeited. However, upon good cause
shown, the Commission will extend the
completion date for a reasonable period
of time. The Commission will recognize,
among other things, developmental
problems attributed to the experimental
nature of the facility or fire, flood
explosion, strike, sabotage, domestic
violence, enemy action an act of the
elements, and other acts beyond the
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control of the permit holder, as a basis
for extending the completion date.
§ 53.1345
permits.
Transfer of construction
A CP may be transferred under
§ 53.1570.
§ 53.1348
permits.
Termination of construction
When a permit holder has determined
to permanently cease construction, the
holder must, within 30 days, submit a
written certification to the NRC.
§ 53.1360
Operating licenses.
Sections 53.1360 through 53.1405 set
out the requirements and procedures
applicable to Commission issuance of
an OL for a nuclear power facility.
§ 53.1366 Contents of applications for
operating licenses; general information.
An application for an OL must
include the information required by
§ 53.1109 and the following
information:
(a) Except for an electric utility
applicant, information sufficient to
demonstrate to the Commission the
financial qualification of the applicant
to carry out, in accordance with the
regulations in this chapter, the activities
for which the license is sought. As
applicable, the following should be
provided:
(1) The applicant must submit
information that demonstrates the
applicant possesses or has reasonable
assurance of obtaining the funds
necessary to cover estimated operation
costs for the period of the license. The
applicant must submit estimates for
total annual operating costs for each of
the first 5 years of operation of the
facility. The applicant must also
indicate the source(s) of funds to cover
these costs.
(2) Each application for an OL
submitted by a newly-formed entity
organized for the primary purpose of
operating the facility must also include
information showing—
(i) The legal and financial
relationships the entity has or proposes
to have with its stockholders or owners;
(ii) The stockholders’ or owners’
financial ability to meet any contractual
obligation to the entity which they have
incurred or proposed to incur; and
(iii) Any other information considered
necessary by the Commission to enable
it to determine the applicant’s financial
qualification.
(3) The Commission may request an
established entity or newly formed
entity to submit additional or more
detailed information respecting its
financial arrangements and status of
funds if the Commission considers this
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information appropriate. This may
include information regarding a
licensee’s ability to continue the
conduct of the activities authorized by
the license and to decommission the
facility.
(b) The application must include
information in the form of a report, as
described in subpart G, indicating how
reasonable assurance will be provided
that funds will be available to
decommission the facility, including a
copy of the financial instrument
obtained to satisfy the requirements of
§ 53.1040.
§ 53.1369 Contents of applications for
operating licenses; technical information.
Final Safety Analysis Report. The
application must contain an FSAR that
describes the facility and the limits on
its operation and presents a safety
analysis of the SSCs of the facility as a
whole. The FSAR must include the
following information, at a level of
detail sufficient to enable the
Commission to reach a final conclusion
on all safety matters that must be
resolved by the Commission before
issuance of an OL. The FSAR must
include the following information:
(a) Site information. An application
for an OL for a commercial nuclear
reactor must include the site
information equivalent to that required
for an early site permit in
§ 53.1146(a)(1)(iv) through (x), including
all current information, such as the
results of environmental and
meteorological monitoring programs,
which has been developed since
issuance of the CP, relating to site
evaluation factors identified in this part.
(b) Design information. Except as
specified in this paragraph, an FSAR for
an OL for a commercial nuclear plant
must include the final design
information equivalent to that required
for a standard design certification as
defined in § 53.1239(a)(2) through (7),
(a)(9), and (a)(11) through (a)(27).
(1) The completed design, including
any changes during construction, must
be described.
(2) Where any design feature had not
been fully developed or demonstrated at
the time of application for the CP, the
applicant must provide the analysis,
research and development, test
programs, gathering of experience, or a
combination thereof to provide the
required demonstration to fulfill the
functional design criteria.
(c) Technical qualifications. A
description of the technical
qualifications of the applicant to engage
in the proposed activities in accordance
with the regulations in this chapter.
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(d) Integrity assessment program. A
description of an Integrity Assessment
Program that addresses the elements
described in § 53.870.
(e) Safeguards information. A
description of the program to protect
Safeguards Information against
unauthorized disclosure in accordance
with the requirements in §§ 73.21 and
73.22 of this chapter, as applicable.
(f) Emergency response facility or
facilities. Description of location and
capabilities to be established for
command and control, support, and
coordination of onsite and offsite, as
applicable, functions during reactor
accident conditions.
(g) Role of personnel. (1) A
description of the completed
assessments related to the role of
personnel in ensuring safe operations
considering the analyses required by
§ 53.730. These assessments must
include the following:
(i) Human factors engineering design
requirements of § 53.730(a);
(ii) Human system interface design
requirements of § 53.730(b);
(iii) Concept of operations of
§ 53.730(c);
(iv) Functional requirements analysis
and function allocation of § 53.730(d);
(2) A description of the program to be
used for evaluating and applying
operating experience as required by
§ 53.730(e);
(3) A staffing plan and supporting
analyses as required by § 53.730(f).
(h) Training, examination, and
proficiency programs. (1) A description
of the training, examination, and
proficiency programs required by
§ 53.730(g);
(2) A description of the training
programs required by § 53.830.
(i) Emergency plan. Emergency plans
complying with the requirements of
§ 53.855.
(1) Include all emergency plan
certifications, as applicable, that have
been obtained from the State, local, and
participating Tribal governmental
agencies with emergency planning
responsibilities that are wholly or
partially within the EPZ plume
exposure pathway. These certifications
must state that—
(i) The proposed emergency plans are
practicable;
(ii) These agencies are committed to
participating in any further
development of the plans, including any
required field demonstrations; and
(iii) These agencies are committed to
executing their responsibilities under
the plans in the event of an emergency.
(2) If certifications cannot be obtained
after sustained, good faith efforts by the
applicant, then the application must
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contain information, including a utility
plan, sufficient to show that the
proposed plans provide reasonable
assurance that adequate protective
measures can and will be taken in the
event of a radiological emergency at the
site.
(3) If complete and integrated
emergency plans were approved as part
of an early site permit, or submitted,
reviewed, and approved as part of the
CP application, new certifications that
demonstrate compliance with the
requirements of paragraph (i)(1) of this
section are not required.
(j) Organization. A description of the
applicant’s organizational structure,
allocations of responsibilities and
authorities, and personnel qualifications
requirements for operation.
(k) Maintenance program. A
description of a maintenance program
under § 53.715.
(l) Quality assurance. A description of
the QAP that demonstrates compliance
with the requirements under § 53.865.
(m) Radiation protection program. A
radiation protection program
description under § 53.850.
(n) Security program. A physical
security plan that describes how the
applicant will comply with § 53.860
(and 10 CFR part 11, if applicable,
including the identification and
description of jobs as required by
§ 11.11(a) of this chapter, at the
proposed facility). The plan must list
tests, inspections, audits, and other
means to be used to demonstrate
compliance with the requirements of 10
CFR parts 11 and 73, if applicable.
(o) Safeguards contingency plan. A
safeguards contingency plan in
accordance with the criteria set forth in
appendix C to 10 CFR part 73. The
safeguards contingency plan must
include plans for dealing with threats,
thefts, and radiological sabotage, as
defined in 10 CFR part 73, relating to
the SNM and nuclear facilities licensed
under this chapter and in the
applicant’s possession and control. Each
application for this type of license must
include the information contained in
the applicant’s safeguards contingency
plan. (Implementing procedures
required for this plan need not be
submitted for approval.) 1
(p) Security training and
qualification. A training and
qualification plan that describes how
the applicant will demonstrate
compliance with the criteria set forth in
§ 73.100 of this chapter or appendix B
to 10 CFR part 73.
(q) Cybersecurity plan. A
cybersecurity plan in accordance with
the criteria set forth in § 73.54 or
§ 73.110 of this chapter.
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(r) Security, safeguards and
cybersecurity plan implementation. A
description of the implementation of the
physical security plan, safeguards
contingency plan, training and
qualification plan, and cybersecurity
plan. Each applicant who prepares a
physical security plan, a safeguards
contingency plan, a training and
qualification plan, or a cybersecurity
plan must protect the plans and other
related Safeguards Information against
unauthorized disclosure in accordance
with the requirements of §§ 73.21 and
73.22 of this chapter.
(s) Fire protection program. A
description of the fire protection
program under § 53.875.
(t) Inservice inspection/inservice
testing program. A description of the
inservice inspection and inservice
testing programs under § 53.880.
(u) [Reserved]
(v) [Reserved]
(w) General employee training. A
description of the training program
required to demonstrate compliance
with § 53.830 and its implementation.
(x) Fitness-for-duty program. A
description of the FFD program required
by 10 CFR part 26 and its
implementation.
(y) Other programs. A description and
evaluation of the results of the
applicant’s programs, including
research and development, if any, to
demonstrate that any safety questions
identified at the CP stage have been
resolved.
(z) Safety design feature performance.
A description of how the performance of
each safety design feature has been
demonstrated capable of fulfilling
functional design criteria considering
interdependent effects through either
analysis, appropriate test programs,
prototype testing, operating experience,
or a combination thereof, in accordance
with § 53.440(a).
(aa) Technical specifications.
Proposed technical specifications
prepared in accordance with the
requirements of § 53.710(a).
1 A physical security plan that contains all
the information required in both § 73.55 or
§ 73.100 of this chapter and appendix C to 10
CFR part 73 satisfies the requirement for a
contingency plan.
§ 53.1372 Contents of applications for
operating licenses; other application
content.
In addition to the FSAR, the
application must also include the
following:
(a) Environmental report. An
environmental report in accordance
with § 51.53(b) of this chapter.
(b) Availability controls (if not
included in the FSAR). A description of
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the controls on plant operations,
including availability controls, to
provide reasonable confidence of safe
operation and that the configurations
and special treatments for SR and
NSRSS SSCs provide the capabilities
and reliabilities required to satisfy the
safety criteria of § 53.220, or more
restrictive alternative criteria adopted
under § 53.470, if not addressed by
Technical Specifications under
§ 53.1369(aa).
§ 53.1375
Review of applications.
(a) Standards for review of
applications. Applications filed under
this part will be reviewed according to
the standards set out in 10 CFR parts 20,
26, 51, 53, 73, and 140.
(b) Administrative review of
applications; hearings. A proceeding on
an OL is subject to all applicable
procedural requirements contained in
10 CFR part 2, including the
requirements for docketing (§ 2.101 of
this chapter) and issuance of a notice of
hearing (§ 2.104 of this chapter). All
hearings on OLs are governed by the
procedures contained in 10 CFR part 2.
§ 53.1381 Referral to the Advisory
Committee on Reactor Safeguards.
The Commission must refer a copy of
the application to the ACRS. The ACRS
must report on those portions of the
application that concern safety and
must apply the standards referenced in
§ 53.1375.
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§ 53.1384 Exemptions, departures, and
variances.
(a) Applicants for an OL under this
part, or any amendment to an OL, may
include in the application a request for
an exemption from one or more of the
Commission’s regulations. The
Commission may grant an exemption
request if it determines that the
exemption complies with § 53.080.
(b) An applicant for an OL who has
filed an application referencing an NRC
approval, permit, license, or
certification issued under this part may
include in the application a request for
departures, variances, or exemptions
related to the subject referenced NRC
approval, permit, license, or
certification. In determining whether to
grant the departure, variance, or
exemption, the Commission must apply
the same technically relevant criteria as
were applicable to the application for
the original or renewed approval,
license, or certification.
§ 53.1387
Issuance of operating licenses.
Upon completion of the construction
or alteration of a facility, in compliance
with the terms and conditions of the
construction permit and subject to any
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necessary testing of the facility for
health or safety purposes, the
Commission will, in the absence of good
cause shown to the contrary, issue an
OL or an appropriate amendment of the
license, as the case may be.
(a)(1) After receiving the report
submitted by the ACRS, the
Commission may issue an OL if the
Commission finds that—
(i) Construction of the facility has
been substantially completed in
conformity with the CP and the
application as amended, the provisions
of the Act, and the rules and regulations
of the Commission;
(ii) Any required notifications to other
agencies or bodies have been duly
made;
(iii) The facility will operate in
conformity with the application as
amended, the provisions of the Act, and
the rules and regulations of the
Commission;
(iv) There is reasonable assurance
that—
(A) the activities authorized by the OL
can be conducted without endangering
the health and safety of the public; and
(B) such activities will be conducted
in compliance with the regulations in
this chapter.
(v) The applicant is technically and
financially qualified to engage in the
activities authorized, however, no
finding of financial qualification is
necessary for an electric utility
applicant for an OL;
(vi) Issuance of the license will not be
inimical to the common defense and
security or to the health and safety of
the public;
(vii) The applicable provisions of 10
CFR part 140 have been satisfied; and
(viii) The findings required by subpart
A of 10 CFR part 51 have been made.
(2) [Reserved]
(b) [Reserved]
(c) The OL will include appropriate
provisions with respect to any
uncompleted items of construction and
such limitations or conditions as are
required to assure that operation during
the period of the completion of such
items will not endanger public health
and safety.
(d) The Commission will issue an OL
in such form and containing such
conditions and limitations, including
technical specifications, as it deems
necessary and appropriate.
§ 53.1390
licenses.
Backfitting of operating
After issuance of an OL, the
Commission may not modify, add, or
delete any term or condition of the OL,
except in accordance with the
provisions of § 53.1590.
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§ 53.1396
Duration of operating licenses.
The Commission will issue an OL
under this part for the term requested by
the applicant, not to exceed 40 years
from the date of issuance, or for the
estimated useful life of the facility if the
Commission determines that the
estimated useful life is less than the
term requested.
§ 53.1399
Transfer of an operating license.
An OL may be transferred under
§ 53.1570.
§ 53.1402
Application for renewal.
The filing of an application for a
renewed license must be in accordance
with § 53.1595.
§ 53.1405
license.
Continuation of an operating
Each OL for a facility that has
permanently ceased operations
continues in effect beyond the
expiration date to authorize ownership
and possession of the facility until the
Commission notifies the licensee in
writing that the license is terminated.
During this period of continued
effectiveness, the licensee must—
(a) Take actions necessary to
decommission and decontaminate the
facility and continue to maintain the
facility, including, where applicable, the
storage, control, and maintenance of the
spent fuel in a safe condition; and
(b) Conduct activities in accordance
with all other restrictions applicable to
the facility in accordance with the
NRC’s regulations and the provisions of
the OL for the facility.
§ 53.1410
Combined licenses.
Sections 53.1410 through 53.1461 set
out the requirements and procedures
applicable to Commission issuance of
COLs for commercial nuclear plants
under this part.
§ 53.1413 Contents of applications for
combined licenses; general information.
An application for a COL must
include the information required by
§ 53.1109 and the following
information:
(a) Except for an electric utility
applicant in regard to financial
assurance required after a Commission
finding under § 53.1452, the application
must include information sufficient to
demonstrate to the Commission the
financial qualification of the applicant
to carry out, in accordance with the
regulations in this chapter, the activities
for which the permit or license is
sought. As applicable, the following
should be provided:
(1) The applicant must submit
information that demonstrates that the
applicant possesses or has reasonable
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assurance of obtaining the funds
necessary to cover estimated
construction costs and related fuel cycle
costs. The applicant must submit
estimates of the total construction costs
of the facility and related fuel cycle
costs and must indicate the source(s) of
funds to cover these costs.
(2) The applicant must submit
information that demonstrates the
applicant possesses or has reasonable
assurance of obtaining the funds
necessary to cover estimated operation
costs for the period of the license. The
applicant must submit estimates for
total annual operating costs for each of
the first 5 years of operation of the
facility. The applicant must also
indicate the source(s) of funds to cover
these costs.
(3) Each application for a COL
submitted by a newly-formed entity
organized for the primary purpose of
constructing and operating a facility
must also include information
showing—
(i) The legal and financial
relationships the entity has or proposes
to have with its stockholders or owners;
and
(ii) The stockholders’ or owners’
financial ability to meet any contractual
obligation to the entity which they have
incurred or proposed to incur; and
(iii) Any other information considered
necessary by the Commission to enable
it to determine the applicant’s financial
qualification.
(4) The Commission may request an
established entity or newly formed
entity to submit additional or more
detailed information respecting its
financial arrangements and status of
funds if the Commission considers this
information appropriate. This may
include information regarding a
licensee’s ability to continue the
conduct of the activities authorized by
the license and to decommission the
facility.
(b) The application must include
information in the form of a report, as
described in subpart G of this part,
indicating how reasonable assurance
will be provided that funds will be
available to decommission the facility.
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§ 53.1416 Contents of applications for
combined licenses; technical information.
(a) Final Safety Analysis Report. The
application must contain an FSAR that
describes the facility and the limits on
its operation and presents a safety
analysis of the SSCs of the facility as a
whole. The Commission will require,
before issuance of a COL, that
engineering documents, such as
analyses, drawings, procurement
specifications, or construction and
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installation specifications, be completed
and available for audit if the more
detailed information is necessary for the
Commission to verify the information in
the application and make its safety
determination. The FSAR must include
the following information, at a level of
detail sufficient to enable the
Commission to reach a final conclusion
on all safety matters that must be
resolved by the Commission before
issuance of a COL:
(1) Site information. An application
for a COL for a commercial nuclear
reactor must include the site
information required for an early site
permit in § 53.1146(a)(1)(iv) through (x).
(2) Design information. An
application for a COL for a commercial
nuclear plant must include the design
information equivalent to that required
for a standard design certification as
defined in § 53.1239(a)(2) through (7),
(a)(9), and (a)(11) through (27).
(3) Technical qualifications. A
description of the technical
qualifications of the applicant to engage
in the proposed activities in accordance
with the regulations in this chapter.
(4) Integrity assessment program. A
description of an Integrity Assessment
Program that addresses the elements
described in § 53.870.
(5) Safeguards information. A
description of the program to protect
Safeguards Information against
unauthorized disclosure in accordance
with the requirements in §§ 73.21 and
73.22 of this chapter, as applicable.
(6) Emergency response facility or
facilities. Description of the locations
and capabilities to be established for
command and control, support, and
coordination of onsite and offsite, as
applicable, functions during reactor
accident conditions.
(7) Role of personnel. (i) A description
of the completed assessments related to
the role of personnel in ensuring safe
operations considering the analyses
required by § 53.730. These assessments
must include the following:
(A) Human factors engineering design
requirements of § 53.730(a);
(B) Human system interface design
requirements of § 53.730(b);
(C) Concept of operations of
§ 53.730(c); and
(D) Functional requirements analysis
and function allocation of § 53.730(d);
(ii) A description of the program to be
used for evaluating and applying
operating experience as required by
§ 53.730(e);
(iii) A staffing plan and supporting
analyses as required by § 53.730(f).
(8) Training, examination, and
proficiency programs. (i) A description
of the training, examination, and
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proficiency programs required by
§ 53.730(g); and
(ii) A description of the training
programs required by § 53.830.
(9) Emergency plan. Emergency plans
complying with the requirements of
§ 53.855.
(i) The emergency plan must include,
as applicable, all emergency plan
certifications that have been obtained
from the State, local, and participating
Tribal governmental agencies with
emergency planning responsibilities.
The certifications must state that—
(A) The proposed emergency plans
are practicable;
(B) These agencies are committed to
participating in any further
development of the plans, including any
required field demonstrations; and
(C) These agencies are committed to
executing their responsibilities under
the plans in the event of an emergency.
(ii) If certifications cannot be obtained
after sustained, good faith efforts by the
applicant, then the application must
contain information, including a utility
plan, sufficient to show that the
proposed plans provide reasonable
assurance that adequate protective
measures can and will be taken in the
event of a radiological emergency at the
site.
(10) Organization. A description of
the applicant’s organizational structure,
allocations of responsibilities and
authorities, and personnel qualifications
requirements for operation.
(11) Maintenance program. A
description of a maintenance program
under § 53.715.
(12) Quality assurance. A description
of the QAP under § 53.865.
(13) Radiation protection program. A
radiation protection program
description under § 53.850.
(14) Security program. A physical
security plan that describes how the
applicant will comply with § 53.860
(and 10 CFR part 11, if applicable,
including the identification and
description of jobs as required by
§ 11.11(a) of this chapter, at the
proposed facility). The plan must list
tests, inspections, audits, and other
means to be used to demonstrate
compliance with the requirements of 10
CFR parts 11 and 73, if applicable.
(15) Safeguards contingency plan. A
safeguards contingency plan in
accordance with the criteria set forth in
appendix C to 10 CFR part 73. The
safeguards contingency plan must
include plans for dealing with threats,
thefts, and radiological sabotage, as
defined in 10 CFR part 73, relating to
the SNM and nuclear facilities licensed
under this chapter and in the
applicant’s possession and control. Each
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application for this type of license must
include the information contained in
the applicant’s safeguards contingency
plan.1 (Implementing procedures
required for this plan need not be
submitted for approval.)
(16) Security training and
qualification. A training and
qualification plan that describes how
the applicant will demonstrate
compliance with the criteria set forth in
§ 73.100 of this chapter or appendix B
to 10 CFR part 73.
(17) Cybersecurity plan. A
cybersecurity plan in accordance with
the criteria set forth in § 73.54 or
§ 73.110 of this chapter.
(18) Security, safeguards and
cybersecurity plan implementation. A
description of the implementation of the
physical security plan, safeguards
contingency plan, training and
qualification plan, and cybersecurity
plan. Each applicant who prepares a
physical security plan, a safeguards
contingency plan, a training and
qualification plan, or a cybersecurity
plan must protect the plans and other
related Safeguards Information against
unauthorized disclosure in accordance
with the requirements of §§ 73.21 and
73.22 of this chapter.
(19) Fire protection program. A
description of the fire protection
program under § 53.875.
(20) Inservice inspection/inservice
testing program. Descriptions of
inservice inspection and inservice
testing programs under § 53.880.
(21) [Reserved]
(22) [Reserved]
(23) General employee training. A
description of the training program
required to demonstrate compliance
with § 53.830 and its implementation.
(24) Fitness-for-duty program. A
description of the FFD program under
part 26 of this chapter and its
implementation.
(25) Technical specifications.
Proposed technical specifications
prepared in accordance with the
requirements of § 53.710(a).
(b) If there are SSCs of the plant for
which research and development is
necessary to confirm the adequacy of
their design, a report which documents
the resolution of any safety questions
associated with such SSCs.
(c) A description of how the
performance of each safety design
feature has been demonstrated capable
of fulfilling functional design criteria
considering interdependent effects
through either analysis, appropriate test
programs, prototype testing, operating
experience, or a combination thereof, in
accordance with § 53.440(a).
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(d) If the COL application references
an early site permit, then the following
requirements apply:
(1) The FSAR need not contain
information or analyses submitted to the
Commission in connection with the
early site permit provided that the FSAR
must either include or incorporate by
reference the early site permit Site
Safety Analysis Report and contain, in
addition to the information and analyses
otherwise required, information
sufficient to demonstrate that the design
of the facility falls within the site
characteristics and design parameters
specified in the early site permit.
(2) If the FSAR does not demonstrate
that design of the facility falls within
the site characteristics and design
parameters, the application must
include a request for a variance that
complies with the requirements of
§§ 53.1188(d) and 53.1437.
(3) The FSAR must demonstrate that
all terms and conditions that have been
included in the early site permit will be
satisfied by the date of issuance of the
COL. Any terms or conditions of the
early site permit that could not be met
by the time of issuance of the COL must
be set forth as terms or conditions of the
COL.
(4) If the early site permit approves
complete and integrated emergency
plans, or major features of emergency
plans, then the FSAR must include any
new or additional information that
updates and corrects the information
that was provided under § 53.1146(b)(2)
and discuss whether the new or
additional information materially
changes the bases for compliance with
the applicable requirements. The
application must identify changes to the
emergency plans or major features of
emergency plans that have been
incorporated into the proposed facility
emergency plans and that constitute or
would constitute a change in an
emergency plan that results in reducing
the licensee’s capability to perform an
emergency planning function in the
event of a radiological emergency.
(5) If complete and integrated
emergency plans are approved as part of
the early site permit, new certifications
meeting the requirements of paragraph
(a)(9)(i) of this section are not required.
(e) If the COL application references
a standard design approval, then the
following requirements apply:
(1) The FSAR need not contain
information or analyses submitted to the
Commission in connection with the
design approval, provided, however,
that the FSAR must either include or
incorporate by reference the standard
design approval FSAR and must
contain, in addition to the information
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and analyses otherwise required,
information sufficient to demonstrate
that the characteristics of the site fall
within the site parameters specified in
the design approval. In addition, the
plant-specific PRA information must
use the PRA information for the design
approval and must be updated to
account for site specific design
information and any design changes or
departures.
(2) The FSAR must demonstrate that
all terms and conditions that have been
included in the design approval will be
satisfied by the date of issuance of the
COL.
(f) If the COL application references a
standard design certification, then the
following requirements apply:
(1) The FSAR need not contain
information or analyses submitted to the
Commission in connection with the
standard design certification, provided,
however, that the FSAR must either
include or incorporate by reference the
standard design certification FSAR and
must contain, in addition to the
information and analyses otherwise
required, information sufficient to
demonstrate that the site characteristics
fall within the site parameters specified
in the standard design certification. In
addition, the plant-specific PRA
information must use the PRA
information for the standard design
certification and must be updated to
account for site-specific design
information and any design changes or
departures.
(2) The FSAR must demonstrate that
the interface requirements established
for the design under § 53.1239(a)(24)
have been met.
(3) The FSAR must demonstrate that
all requirements and restrictions set
forth in the referenced standard design
certification rule must be satisfied by
the date of issuance of the COL. Any
requirements and restrictions set forth
in the referenced standard design
certification rule that could not be
satisfied by the time of issuance of the
COL, must be set forth as terms or
conditions of the COL.
(g) If the COL application references
the use of one or more manufactured
reactors licensed under § 53.1270, then
the following requirements apply:
(1) The FSAR need not contain
information or analyses submitted to the
Commission in connection with the ML,
provided, however, that the FSAR must
either include or incorporate by
reference the ML FSAR and must
contain, in addition to the information
and analyses otherwise required,
information sufficient to demonstrate
that the site characteristics fall within
the site parameters specified in the ML.
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In addition, the plant-specific PRA
information must use the PRA
information for the manufactured
reactor and must be updated to account
for site-specific design information and
any design changes or departures.
(2) The FSAR must demonstrate that
the interface requirements established
for the design have been met.
(3) The FSAR must demonstrate that
all terms and conditions that have been
included in the ML will be satisfied by
the date of issuance of the COL. Any
terms or conditions of the ML that could
not be met by the time of issuance of the
COL, must be set forth as terms or
conditions of the COL.
(h) Each applicant for a COL under
this part must protect Safeguards
Information against unauthorized
disclosure in accordance with the
requirements in §§ 73.21 and 73.22 of
this chapter, as applicable.
1 A physical security plan that contains all
the information required in both § 73.55 or
§ 73.100 of this chapter and appendix C to 10
CFR part 73 demonstrates compliance with
the requirement for a contingency plan.
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§ 53.1419 Contents of applications for
combined licenses; other application
content.
(a) In addition to the FSAR, the
application must also include the
following:
(1) Environmental report. (i) An
environmental report either in
accordance with § 51.50(c) of this
chapter if an LWA under § 53.1130 is
not requested in conjunction with the
COL application, or in accordance with
§§ 51.49 and 51.50(c) of this chapter if
an LWA is requested in conjunction
with the COL application; or
(ii) If the applicant wishes to request
that an LWA under § 53.1130 be issued
before issuance of the COL, the
information otherwise required by
§ 53.1130, in accordance with either
§ 2.101(a)(1) through (a)(4), or
§ 2.101(a)(9) of this chapter;
(2) Availability controls (if not
included in the FSAR). A description of
the controls on plant operations,
including availability controls, to
provide reasonable confidence of safe
operation and that the configurations
and special treatments for SR SSCs and
NSRSS SSCs provide the capabilities
and reliabilities required to satisfy the
safety criteria of § 53.220, or more
restrictive alternative criteria adopted
under § 53.470, if not addressed by
Technical Specifications under
§ 53.1416(a)(25); and
(3) Inspections, tests, analyses, and
acceptance criteria. The proposed
inspections, tests, and analyses,
including those applicable to emergency
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planning, that the licensee must
perform, and the acceptance criteria that
are necessary and sufficient to provide
reasonable assurance that, if the
inspections, tests, and analyses are
performed and the acceptance criteria
met, the facility has been constructed
and will be operated in conformity with
the COL, the provisions of the Act, and
the Commission’s rules and regulations.
(i) If the application references an
early site permit with ITAAC, the early
site permit ITAAC must apply to those
aspects of the COL which are approved
in the early site permit.
(ii) If the application references a
standard design certification, the ITAAC
contained in the certified design must
apply to those portions of the facility
design which are approved in the
standard design certification.
(iii) If the application references an
ML, the ITAAC contained in the ML
must apply to those portions of the
facility design which are approved in
the ML.
(iv) If the application references an
early site permit with ITAAC, a
standard design certification, an ML, or
combination thereof, the application
may include a notification that a
required inspection, test, or analysis in
the ITAAC has been successfully
completed and that the corresponding
acceptance criterion has been met. The
Federal Register notification required
by § 53.1422 of this chapter must
indicate that the application includes
this notification.
(b) [Reserved]
§ 53.1422
Review of applications.
(a) Standards for review of
applications. Applications filed under
this part will be reviewed according to
the standards set out in this part and 10
CFR parts 20, 51, 73, and 140.
(b) Administrative review of
applications; hearings. A proceeding on
a COL is subject to all applicable
procedural requirements contained in
10 CFR part 2, including the
requirements for docketing (§ 2.101 of
this chapter) and issuance of a notice of
hearing (§ 2.104 of this chapter). If an
applicant requests a Commission
finding on certain ITAAC with the
issuance of the COL, then those ITAAC
will be identified in the notice of
hearing. All hearings on COLs are
governed by the procedures contained
in 10 CFR part 2.
§ 53.1425 Finality of referenced NRC
approvals.
If the application for a COL under this
part references an early site permit,
standard design certification rule,
standard design approval, or ML, issued
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under this part, the scope and nature of
matters resolved for the application and
any COL issued are governed by the
relevant provisions addressing finality,
including §§ 53.1188, 53.1221, 53.1263,
and 53.1288.
§ 53.1431 Referral to the Advisory
Committee on Reactor Safeguards.
The Commission must refer a copy of
the application to the ACRS. The ACRS
must report on those portions of the
application that concern safety and
must apply the standards referenced in
§ 53.1422, in accordance with the
finality provisions in § 53.1425.
§ 53.1434 Authorization to conduct limited
work authorization activities.
(a) If the application for a COL under
this part does not reference an early site
permit which authorizes the holder to
perform the activities under
§ 53.1130(b), the applicant may not
perform those activities without
obtaining the separate authorization
required by § 53.1130(a). Authorization
may be granted only after the presiding
officer in the proceeding on the
application has made the findings and
determination required by
§ 53.1130(b)(1)(ii) and (b)(1)(iv), and the
Director, Office of Nuclear Reactor
Regulation makes the determination
required by § 53.1130(b)(1)(iii).
(b) If, after an applicant has performed
the activities permitted by a LWA
issued under § 53.1130, the application
for the COL is withdrawn or denied,
then the applicant must implement the
approved site redress plan.
§ 53.1437 Exemptions, departures, and
variances.
(a) An applicant for a COL, or any
amendment to a COL, may include in
the application a request for an
exemption from one or more of the
Commission’s regulations.
(1) If the request is for an exemption
from any part of a referenced standard
design certification rule, the
Commission may grant the request if it
determines that the exemption complies
with any exemption provisions of the
referenced standard design certification
rule, or with § 53.1263 if there are no
applicable exemption provisions in the
referenced standard design certification
rule.
(2) For all other requests for
exemptions, the Commission may grant
a request if it determines that the
exemption complies with § 53.080.
(b) An applicant for a COL who has
filed an application referencing an early
site permit issued under § 53.1158 may
include in the application a request for
a variance from one or more site
characteristics, design parameters, or
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terms and conditions of the permit, or
from the Site Safety Analysis Report. In
determining whether to grant the
variance, the Commission must apply
the same technically relevant criteria as
were applicable to the application for
the original or renewed site permit.
Once a COL referencing an early site
permit is issued, variances from the
early site permit will not be granted for
that CP or COL.
(c) An applicant for a COL who has
filed an application referencing use of a
manufactured reactor may include in
the application a request for a departure
from one or more design characteristics,
site parameters, terms and conditions,
or approved design of the manufactured
reactor under the ML issued under
§ 53.1287. The Commission may grant
such a request only if it determines that
the departure will comply with the
requirements of § 53.080, and that the
special circumstances outweigh any
decrease in safety that may result from
the reduction in standardization caused
by the departure.
(d) Issuance of a variance under
paragraph (b) of this section or a
departure under paragraph (c) of this
section is subject to litigation during the
COL proceeding in the same manner as
other issues material to that proceeding.
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§ 53.1440
Issuance of combined licenses.
(a)(1) After conducting a hearing
under § 53.1422(b) and receiving the
report submitted by the ACRS, the
Commission may issue a COL if the
Commission finds that—
(i) The applicable standards and
requirements of the Act and the
Commission’s regulations have been
met;
(ii) Any required notifications to other
agencies or bodies have been duly
made;
(iii) There is reasonable assurance that
the facility will be constructed and will
operate in conformity with the license,
the provisions of the Act, and the
Commission’s regulations;
(iv) The applicant is technically and
financially qualified to engage in the
activities authorized; however, no
finding of financial qualification is
necessary for an electric utility
applicant for a COL;
(v) Issuance of the license will not be
inimical to the common defense and
security or to the health and safety of
the public; and
(vi) The findings required by subpart
A of 10 CFR part 51 have been made.
(2) The Commission may also find, at
the time it issues the COL, that certain
acceptance criteria in one or more of the
ITAAC in a referenced early site permit,
standard design certification, or ML
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have been met. This finding will finally
resolve that those acceptance criteria
have been met, those acceptance criteria
will be deemed to be excluded from the
COL, and findings under § 53.1452(g)
with respect to those acceptance criteria
are unnecessary.
(b) The Commission must identify
within the COL the inspections, tests,
and analyses, including those applicable
to emergency planning, that the licensee
must perform, and the acceptance
criteria that, if met, are necessary and
sufficient to provide reasonable
assurance that the facility has been
constructed and will be operated in
conformity with the license, the
provisions of the Act, and the
Commission’s rules and regulations.
(c) A COL must contain the terms and
conditions, including technical
specifications, as the Commission
deems necessary and appropriate.
§ 53.1443
Finality of combined licenses.
(a) After issuance of a COL, the
Commission may not modify, add, or
delete any term or condition of the COL,
the design of the facility, the ITAAC
contained in the license that are not
derived from a referenced standard
design certification or ML, except under
the provisions of § 53.1452 or § 53.1590.
(b) If the COL does not reference a
standard design certification or use of a
manufactured reactor under an ML
issued under § 53.1287, then a licensee
may make changes in the facility as
described in the FSAR (as updated) and
make changes in the procedures as
described in the FSAR (as updated)
under the applicable change processes
in § 53.1550.
(c) If the COL references a certified
design, then—
(1) Changes to or departures from
information within the scope of the
referenced standard design certification
rule are subject to the applicable change
processes in that rule; and
(2) Changes that are not within the
scope of the referenced standard design
certification rule are subject to the
applicable change processes in subpart
I of this part, unless they also involve
changes to or noncompliance with
information within the scope of the
referenced standard design certification
rule. In these cases, the applicable
provisions of this section and the
standard design certification rule apply.
(d) If the COL references use of a
manufactured reactor under an ML
issued under this part, then—
(1) Changes to or departures from
information within the scope of the
manufactured reactor’s design are
subject to the change processes in
§ 53.1288; and
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(2) Changes that are not within the
scope of the manufactured reactor’s
design are subject to the applicable
change processes in subpart I.
(e) The Commission may issue and
make immediately effective any
amendment to a COL upon a
determination by the Commission that
the amendment involves no significant
hazards consideration, notwithstanding
the pendency before the Commission of
a request for a hearing from any person.
The amendment may be issued and
made immediately effective in advance
of the holding and completion of any
required hearing. The amendment will
be processed under the procedures
specified in § 53.1515.
(f) Any modification to, addition to, or
deletion from the terms and conditions
of a COL, including any modification to,
addition to, or deletion from the
inspections, tests, and analyses, or
related acceptance criteria contained in
the license is a proposed amendment to
the license. There must be an
opportunity for a hearing on the
amendment.
§ 53.1449
Inspection during construction.
(a) Licensee schedule for inspections,
tests, or analyses. The licensee must
submit to the NRC, no later than 1 year
after issuance of the COL or at the start
of construction as defined at § 53.020,
whichever is later, its schedule for
completing the inspections, tests, or
analyses in the ITAAC. The licensee
must submit updates to the ITAAC
schedules every 6 months thereafter
and, within 1 year of its scheduled date
for initial loading of fuel (or, for a fueled
manufactured reactor, within 1 year of
its scheduled date for initiating the
physical removal of any one of the
independent physical mechanisms to
prevent criticality required under
§ 53.620(d)(1)), the licensee must submit
updates to the ITAAC schedule every 30
days until the final notification is
provided to the NRC under paragraph
(c)(1) of this section.
(b) Licensee and applicant conduct of
activities subject to ITAAC. With respect
to activities subject to an ITAAC, an
applicant for a COL may proceed at its
own risk with design and procurement
activities, and a licensee may proceed at
its own risk with design, procurement,
construction, and preoperational
activities, even though the NRC may not
have found that any one of the
prescribed acceptance criteria are met.
(c) Licensee notifications. (1) ITAAC
closure notification. The licensee must
notify the NRC that prescribed
inspections, tests, and analyses have
been performed and that the prescribed
acceptance criteria are met. The
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notification must contain sufficient
information to demonstrate that the
prescribed inspections, test, and
analyses have been performed and that
the prescribed acceptance criteria are
met.
(2) ITAAC post-closure notifications.
Following the licensee’s ITAAC closure
notifications under paragraph (c)(1) of
this section until the Commission makes
the finding under § 53.1452(g), the
licensee must notify the NRC, in a
timely manner, of new information that
materially alters the basis for
determining that either inspections,
tests, or analyses were performed as
required, or that acceptance criteria are
met. The notification must contain
sufficient information to demonstrate
that, notwithstanding the new
information, the prescribed inspections,
tests, and analyses have been performed
as required, and the prescribed
acceptance criteria are met.
(3) Uncompleted ITAAC notification.
If the licensee has not provided, by the
date 225 days before the scheduled date
for initial loading of fuel (or, for a fueled
manufactured reactor, by the date 225
days before the scheduled date for
initiating the physical removal of any
one of the independent physical
mechanisms to prevent criticality
required under § 53.620(d)(1)), the
notification required by paragraph (c)(1)
of this section for all ITAAC, then the
licensee must notify the NRC that the
prescribed inspections, tests, or analyses
for all uncompleted ITAAC will be
performed and that the prescribed
acceptance criteria will be met prior to
operation. The notification must be
provided no later than the date 225 days
before the scheduled date for initial
loading of fuel (or, for a fueled
manufactured reactor, no later than the
date 225 days before the scheduled date
for initiating the physical removal of
any one of the independent physical
mechanisms to prevent criticality
required under § 53.620(d)(1)), and must
provide sufficient information to
demonstrate that the prescribed
inspections, tests, or analyses will be
performed and the prescribed
acceptance criteria for the uncompleted
ITAAC will be met, including, but not
limited to, a description of the specific
procedures and analytical methods to be
used for performing the prescribed
inspections, tests, and analyses and
determining that the prescribed
acceptance criteria are met.
(4) All ITAAC complete notification.
The licensee must notify the NRC that
all ITAAC are complete.
(d) Licensee determination of
noncompliance with ITAAC. (1) In the
event that an activity is subject to an
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ITAAC derived from a referenced
standard design certification and the
licensee has not demonstrated that the
prescribed acceptance criteria are met,
the licensee may take corrective actions
to successfully complete that ITAAC or
request an exemption from the standard
design certification ITAAC, as
applicable. A request for an exemption
must also be accompanied by a request
for a license amendment under subpart
I.
(2) In the event that an activity is
subject to an ITAAC not derived from a
referenced standard design certification
and the licensee has not demonstrated
that the prescribed acceptance criteria
are met, the licensee may take corrective
actions to successfully complete that
ITAAC or request a license amendment
under subpart I.
(e) NRC inspection, publication of
notices, and availability of licensee
notifications. The NRC must ensure that
the prescribed inspections, tests, and
analyses in the ITAAC are performed.
(1) At appropriate intervals until the
last date for submission of requests for
hearing under § 53.1452, the NRC must
publish notices in the Federal Register
of the NRC staff’s determination of the
successful completion of inspections,
tests, and analyses.
(2) The NRC must make publicly
available the licensee notifications
under paragraph (c) of this section. The
NRC must, no later than the date of
publication of the notice of intended
operation required by § 53.1452(a),
make publicly available those licensee
notifications under paragraph (c) of this
section that have been submitted to the
NRC at least 7 days before that notice.
§ 53.1452
license.
Operation under a combined
(a) The licensee must notify the NRC
of its scheduled date for initial loading
of fuel no later than 270 days before the
scheduled date and must notify the NRC
of updates to its schedule every 30 days
thereafter.1 Not less than 180 days
before the date scheduled for initial
loading of fuel into a plant by a licensee
that has been issued a COL under this
part, the Commission must publish
notice of intended operation in the
Federal Register.2 The notice must
provide that any person whose interest
may be affected by operation of the
plant may, within 60 days, request that
the Commission hold a hearing on
whether the facility as constructed
complies, or on completion will
comply, with the acceptance criteria in
the COL, except that a hearing must not
be granted for those ITAAC that the
Commission found were met under
§ 53.1440(a)(2).
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(b) A request for hearing under
paragraph (a) of this section must show,
prima facie—
(1) That one or more of the acceptance
criteria of the ITAAC in the COL have
not been, or will not be, met; and
(2) The specific operational
consequences of nonconformance that
would be contrary to providing
reasonable assurance of adequate
protection of the public health and
safety.
(c) The Commission, acting as the
presiding officer, must determine
whether to grant or deny the request for
hearing under the applicable
requirements of § 2.309 of this chapter.
If the Commission grants the request,
the Commission, acting as the presiding
officer, must determine whether during
a period of interim operation there will
be reasonable assurance of adequate
protection to the public health and
safety. The Commission’s determination
must consider the petitioner’s prima
facie showing and any answers thereto.
If the Commission determines there is
such reasonable assurance, it must
allow operation during an interim
period under the COL.
(d) The Commission, in its discretion,
must determine appropriate hearing
procedures, whether informal or formal
adjudicatory, for any hearing under
paragraph (a) of this section, and must
state its reasons therefore.
(e) The Commission must, to the
maximum possible extent, render a
decision on issues raised by the hearing
request within 180 days of the
publication of the notice provided by
paragraph (a) of this section or by the
anticipated date for initial loading of
fuel into the reactor, whichever is later.
(f) A petition to modify the terms and
conditions of the COL will be processed
as a request for action under § 2.206 of
this chapter. The petitioner must file the
petition with the Secretary of the
Commission. Before the licensed
activity allegedly affected by the
petition (fuel loading, low power
testing, etc.) commences, the
Commission must determine whether
any immediate action is required. If the
petition is granted, then an appropriate
order will be issued. Fuel loading and
operation under the COL will not be
affected by the granting of the petition
unless the order is made immediately
effective.
(g) The licensee must not operate the
facility until the Commission makes a
finding that the acceptance criteria in
the COL are met, except for those
acceptance criteria that the Commission
found were met under § 53.1440(a)(2). If
the COL is for a modular design, each
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reactor unit may require a separate
finding as construction proceeds.
(h) After the Commission has made
the finding in paragraph (g) of this
section, the ITAAC do not, by virtue of
their inclusion in the COL, constitute
regulatory requirements either for
licensees or for renewal of the license;
except for the specific ITAAC for which
the Commission has granted a hearing
under paragraph (a) of this section, all
ITAAC expire upon final Commission
action in the proceeding. However,
subsequent changes to the facility or
procedures described in the FSAR (as
updated) must comply with the
requirements in § 53.1443(e) or (f), as
applicable.
1 For licensees installing fueled
manufactured reactors under a COL, the COL
holder must instead notify the NRC of its
scheduled date for initiating the physical
removal of any one of the independent
physical mechanisms to prevent criticality
required under § 53.620(d)(1) no later than
270 days before the scheduled date and must
notify the NRC of updates to its schedule
every 30 days thereafter.
2 For licensees installing fueled
manufactured reactors under a COL, the
Commission must instead publish notice of
intended operation in the Federal Register
not less than 180 days before the date
scheduled for initiating the physical removal
of any one of the independent physical
mechanisms to prevent criticality required
under § 53.620(d)(1).
§ 53.1455
Duration of combined license.
A COL is issued for a specified period
not to exceed 40 years from the date on
which the Commission makes a finding
that acceptance criteria are met under
§ 53.1452(g) or allowing operation
during an interim period under the COL
under § 53.1452(c).
§ 53.1456
Transfer of a combined license.
A COL may be transferred under
§ 53.1570.
§ 53.1458
Application for renewal.
The filing of an application for a
renewed license must be in accordance
with § 53.1595.
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§ 53.1461
license.
Continuation of combined
Each COL for a facility that has
permanently ceased operations
continues in effect beyond the
expiration date to authorize ownership
and possession of the facility until the
Commission notifies the licensee in
writing that the license is terminated.
During this period of continued
effectiveness, the licensee must—
(a) Take actions necessary to
decommission and decontaminate the
facility and continue to maintain the
facility, including, where applicable, the
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storage, control and maintenance of the
spent fuel, in a safe condition; and
(b) Conduct activities in accordance
with all other restrictions applicable to
the facility in accordance with the
NRC’s regulations and the provisions of
the COL for the facility.
§ 53.1470 Standardization of commercial
nuclear plant designs: licenses to construct
and operate nuclear power reactors of
identical design at multiple sites.
(a) Except as otherwise specified in
this section, the provisions of this
section apply to CP, OL, and COL
applications for commercial nuclear
plants of identical design (the ‘‘common
design’’) under this part.
(b) Each application for a CP, OL, or
COL submitted pursuant to this section
must be submitted as specified in
§§ 53.1300, 53.1360, or 53.1410,
respectively, and § 2.101 of this chapter.
Each application must state that the
applicant wishes to construct a facility
identical to a facility proposed for one
or more sites other than the applicant’s
(the ‘‘common design’’), and the
applicant wishes to have the application
considered under this section. Each
application must list each of the other
applications to be treated together under
this section.
(c) Each application must include the
information required by the applicable
sections of this subpart, provided
however, that the application must
identify the common design, and, if
applicable, reference a standard design
certification or standard design approval
under this part, or the use of a reactor
manufactured under this part. The
FSAR for each application must either
incorporate by reference or include the
final safety analysis of the common
design, including, if applicable, the
FSAR for the referenced standard design
certification, standard design approval,
or the manufactured reactor.
(d) Each application submitted
pursuant to this section must contain an
environmental report under
§§ 53.1312(a)(1), 53.1372(a), or
53.1419(a)(1), as applicable, that
complies with the applicable provisions
of 10 CFR part 51, provided, however,
that the application may incorporate by
reference a single environmental report
on the environmental impacts of the
common design that are applicable to
each site.
(e) Upon a determination that each
application is acceptable for docketing
under § 2.101 of this chapter, each
application will be docketed and a
notice of docketing for each application
will be published in the Federal
Register, under § 2.104 of this chapter,
provided, however, that the notice must
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state that the application will be
processed under the provisions of this
section and subpart D of 10 CFR part 2.
At the discretion of the Commission, a
single notice of docketing for multiple
applications may be published in the
Federal Register.
(f) The NRC must prepare an
environmental assessment or draft and
final environmental impact statements
for each of the applications under 10
CFR part 51. Scoping under §§ 51.28
and 51.29 of this chapter for each of the
license applications may be conducted
simultaneously and joint scoping may
be conducted with respect to the
environmental issues relevant to the
common design. If the applications
reference a standard design certification,
then the environmental assessment or
environmental impact statement for
each of the applications must
incorporate by reference the standard
design certification environmental
assessment. If the applications do not
reference a standard design certification,
then the NRC must prepare
environmental assessments or draft and
final supplemental environmental
impact statements which address severe
accident mitigation design alternatives
for the common design, which must be
incorporated by reference into the
environmental assessment or
environmental impact statement
prepared for each application. Scoping
under §§ 51.28 and 51.29 of this chapter
for the supplemental environmental
impact statement may be conducted
simultaneously and may be part of the
scoping for each of the applications.
(g) The ACRS must report on each of
the applications as required by the
applicable sections of this subpart. Each
report must be limited to those safety
matters for each application that are not
relevant to the common design. In
addition, the ACRS must separately
report on the safety of the common
design, provided, however, that the
report need not address the safety of a
referenced standard design certification
or reactor manufactured under this part.
(h) The Commission must designate a
presiding officer to conduct the
proceeding with respect to the health
and safety, common defense and
security, and environmental matters
relating to the common design and
affecting at least two applications. The
hearing will be governed by the
applicable provisions of subparts A, C,
G, L, N, and O of 10 CFR part 2 relating
to applications for CPs, OLs, and COLs.
The presiding officer must issue a
partial initial decision on the common
design.
(i) If the design for the power
reactor(s) proposed in a particular
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application is not identical to the others,
that application may not be processed
under this section and subpart D of 10
CFR part 2.
(j) As used in this section, the design
of a nuclear power reactor included in
a single referenced Safety Analysis
Report means the design of those SSCs
important to radiological health and
safety and the common defense and
security.
Subpart I—Maintaining and Revising
Licensing-Basis Information
§ 53.1500
Licensing-basis information.
This subpart provides the
requirements for each holder of a
license for a commercial nuclear plant
licensed under this part to maintain
licensing-basis information as defined
in § 53.020; evaluate changes to site
characteristics, plant design features,
and programmatic controls to determine
needed approvals and revisions; and
submit appropriate updates to the U.S.
Nuclear Regulatory Commission (NRC).
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§ 53.1502
licenses.
Specific terms and conditions of
(a) Each license issued under this part
is subject to the provisions of the
Atomic Energy Act of 1954, as amended,
(the Act) and to all rules, regulations,
and orders of the Commission. The
terms and conditions of the license will
be subject to amendment, revision, or
modification, by reason of amendments
of the Act or by reason of rules,
regulations, and orders issued in
accordance with the terms of the Act.
(b) Each license issued under this part
must be subject to all conditions
imposed as a matter of law by sections
401(a)(2) and 401(d) of the Federal
Water Pollution Control Act, as
amended (33 U.S.C.A. 1341(a)(2) and
(d)).
(c) A holder of an operating license
(OL) or combined license (COL) under
this part may take reasonable action that
departs from a license condition or a
technical specification included in a
license issued under this part in a
national security emergency established
by a law enacted by the Congress or by
an order or directive issued by the
President pursuant to statutes or the
Constitution of the United States. The
authority under this paragraph must be
exercised in accordance with law,
including section 57e of the Act, and is
in addition to the authority granted
under § 53.740(h), which remains in
effect unless otherwise directed by the
Commission during a national security
emergency. The authority under this
paragraph may be exercised—
(1) When this action is immediately
needed to implement national security
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objectives as designated by the national
command authority through the
Commission; and
(2) No action consistent with license
conditions and technical specifications
that can satisfy national security
objectives is immediately apparent.
(d)(1) If the NRC finds that the state
of emergency preparedness does not
provide reasonable assurance that
adequate protective measures can and
will be taken in the event of a
radiological emergency (including
findings based on requirements of 10
CFR part 50, appendix E, section IV.D.3)
and if the deficiencies (including
deficiencies based on requirements of
10 CFR part 50, appendix E, section
IV.D.3) are not corrected within 4
months of that finding, the Commission
will determine whether the facility must
be shut down or cease operations until
such deficiencies are remedied or
whether other enforcement action is
appropriate. In determining whether a
shutdown or other enforcement action is
appropriate, the Commission will take
into account, among other factors,
whether the licensee can demonstrate to
the Commission’s satisfaction that the
deficiencies in the plan are not
significant for the plant in question, or
that adequate interim compensating
actions have been or will be taken
promptly, or that that there are other
compelling reasons for continued
operation.
(2) If the planning standards for
radiological emergency preparedness
apply to offsite emergency response
plans, or if the planning activities in
§ 50.160(b)(1)(iv)(B) apply, then the
NRC will base its finding on a review of
the Federal Emergency Management
Agency findings and determinations as
to whether State, participating Tribal
and local emergency plans are adequate
and capable of being implemented, and
on the NRC assessment as to whether
the licensee’s emergency plans are
adequate and capable of being
implemented. Nothing in this paragraph
must be construed as limiting the
authority of the Commission to take
action under any other regulation or
authority of the Commission or at any
time other than that specified in this
paragraph.
§ 53.1505 Changes to licensing-basis
information requiring prior NRC approval.
(a) Sections 53.1510 through 53.1520
provide the process for a licensee to
request and the NRC to issue
amendments to licenses, including any
conditions contained therein, technical
specifications or other attachments to a
license, and any orders issued by the
NRC modifying a license. Sections
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53.1525 and 53.1530 govern proposed
changes to a commercial nuclear plant
referencing a certified design or
manufacturing license (ML).
(b) A licensee may propose changing
licensing-basis information established
by NRC regulations by requesting an
exemption in accordance with § 53.080.
§ 53.1510
license.
Application for amendment of
Whenever a holder of a license under
this part desires to amend the license,
an application for an amendment must
be filed with the Commission, as
specified in § 53.040, that fully
describes the changes desired and,
following as far as applicable, the form
prescribed for original applications.
Applications for amendments involving
changes to plant structures, systems,
and components (SSCs), programmatic
controls, or the role of plant personnel
must include an assessment of the
changes in relation to the safety
requirements in subpart B of this part
and the analyses requirements of
§ 53.450 as applicable, an analysis of
whether the amendment involves no
significant hazards consideration using
the standards in § 53.1520, and a
consideration of environmental factors.
§ 53.1515 Public notices; State
consultation.
The Commission will use the
following procedures for an application
requesting an amendment to an OL or
COL issued under this part.
(a) Public notices. (1)(i) The
Commission may publish in the Federal
Register under § 2.105 of this chapter an
individual notice of proposed action for
an amendment for which it makes a
proposed determination that no
significant hazards consideration is
involved, or, at least once every 30 days,
publish a periodic Federal Register
notice of proposed actions, which
identifies each amendment issued and
each amendment proposed to be issued
since the last such periodic notice, or it
may publish both such notices.
(ii) For each amendment proposed to
be issued, the notice will
(A) Contain the staff’s proposed
determination under the standards in
§ 53.1520;
(B) Provide a brief description of the
amendment and of the facility involved;
(C) Solicit public comments on the
proposed determination; and
(D) Provide for a 30-day comment
period.
(iii) The comment period will begin
on the day after the date of the
publication of the first notice, and,
normally, the amendment will not be
granted until after this comment period
expires.
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(2) The Commission may inform the
public about the final disposition of an
amendment request for which it has
made a proposed determination of no
significant hazards consideration either
by issuing an individual notice of
issuance under § 2.106 of this chapter or
by publishing such a notice in its
periodic system of Federal Register
notices. In either event, it will not make
and will not publish a final
determination of no significant hazards
consideration unless it receives a
request for a hearing on that amendment
request.
(3) Where the Commission makes a
final determination that no significant
hazards consideration is involved and
that the amendment should be issued,
the amendment will be effective on
issuance, even if adverse public
comments have been received and even
if an interested person meeting the
provisions for intervention called for in
§ 2.309 of this chapter has filed a
request for a hearing. The Commission
need hold any required hearing only
after it issues an amendment, unless it
determines that a significant hazards
consideration is involved, in which case
the Commission will provide an
opportunity for a prior hearing.
(4) Where the Commission finds that
an emergency situation exists, in that
failure to act in a timely way would
result in derating or shutdown of a
commercial nuclear reactor, or in
prevention of either resumption of
operation or of increase in power output
up to the plant’s licensed power level,
it may issue a license amendment
involving no significant hazards
consideration without prior notice and
opportunity for a hearing or for public
comment. In such a situation, the
Commission will not publish a notice of
proposed determination on no
significant hazards consideration but
will publish a notice of issuance under
§ 2.106 of this chapter providing for
opportunity for a hearing and for public
comment after issuance. The
Commission expects its licensees to
apply for license amendments in a
timely fashion. It will decline to
dispense with notice and comment on
the determination of no significant
hazards consideration if it determines
that the licensee has abused the
emergency provision by failing to make
timely application for the amendment
and thus itself creating the emergency.
Whenever an emergency situation
exists, a licensee requesting an
amendment must explain why this
emergency situation occurred and why
it could not avoid this situation, and the
Commission will assess the licensee’s
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reasons for failing to file an application
sufficiently in advance of that event.
(5) Where the Commission finds that
exigent circumstances exist, in that a
licensee and the Commission must act
quickly and that time does not permit
the Commission to publish a Federal
Register notice allowing 30 days for
prior public comment, and it also
determines that the amendment
involves no significant hazards
considerations, it—
(i)(A) Will either issue a Federal
Register notice providing notice of an
opportunity for hearing and allowing at
least 2 weeks from the date of the notice
for prior public comment; or
(B) Will use local media to provide
reasonable notice to the public in the
area surrounding a licensee’s facility of
the licensee’s amendment and of its
proposed determination as described in
paragraph (a)(1) of this section,
consulting with the licensee on the
proposed media release and on the
geographical area of its coverage;
(ii) Will provide for a reasonable
opportunity for the public to comment,
using its best efforts to make available
to the public whatever means of
communication it can for the public to
respond quickly, and, in the case of
telephone comments, have these
comments recorded or transcribed, as
necessary and appropriate;
(iii) When it has issued a local media
release, may inform the licensee of the
public’s comments, as necessary and
appropriate;
(iv) Will publish a notice of issuance
under § 2.106 of this chapter;
(v) Will provide a hearing after
issuance, if one has been requested by
a person who satisfies the provisions for
intervention specified in § 2.309 of this
chapter; and
(vi) Will require the licensee to
explain the exigency and why the
licensee cannot avoid it and use its
normal public notice and comment
procedures in paragraph (a)(1) of this
section if it determines that the licensee
has failed to use its best efforts to make
a timely application for the amendment
in order to create the exigency and to
take advantage of this procedure.
(6) Where the Commission finds that
significant hazards considerations are
involved, it will issue a Federal Register
notice providing an opportunity for a
prior hearing even in an emergency
situation, unless it finds an imminent
danger to the health or safety of the
public, in which case it will issue an
appropriate order or rule under 10 CFR
part 2.
(b) State consultation. (1) At the time
a licensee requests an amendment, it
must notify the State in which its
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facility is located of its request by
providing that State with a copy of its
application and its reasoned analysis
about no significant hazards
considerations and indicate on the
application that it has done so.
(2) The Commission will advise the
State of its proposed determination
about no significant hazards
consideration normally by sending it a
copy of the Federal Register notice.
(3) The Commission will make the
names of the Project Manager or other
NRC personnel it designated to consult
with the State available to the State
official designated to consult about its
proposed determination. The
Commission will consider any
comments of that State official. If it does
not hear from the State in a timely
manner, it will consider that the State
has no interest in its determination;
nonetheless, to ensure that the State is
aware of the application, before it issues
the amendment, it will make a good
faith effort to communicate directly
with that official. (Inability to consult
with a responsible State official
following good faith attempts will not
prevent the Commission from making
effective a license amendment involving
no significant hazards consideration.)
(4) The Commission will make a good
faith attempt to consult with the State
before it issues a license amendment
involving no significant hazards
consideration. If, however, it does not
have time to use its normal consultation
procedures because of an emergency
situation, it will attempt to
communicate directly with the
appropriate State official. (Inability to
consult with a responsible State official
following good faith attempts will not
prevent the Commission from making
effective a license amendment involving
no significant hazards consideration, if
the Commission deems it necessary in
an emergency situation.)
(5) After the Commission issues the
requested amendment, it will send a
copy of its determination to the State.
(c) Caveats about State consultation.
(1) The State consultation procedures in
paragraph (b) of this section do not give
the State a right—
(i) To veto the Commission’s
proposed or final determination;
(ii) To a hearing on the determination
before the amendment becomes
effective; or
(iii) To insist upon a postponement of
the determination or upon issuance of
the amendment.
(2) These procedures do not alter
present provisions of law that reserve to
the Commission exclusive responsibility
for setting and enforcing radiological
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health and safety requirements for
commercial nuclear plants.
§ 53.1520
Issuance of amendment.
(a) In determining whether an
amendment to a license will be issued
to the applicant, the Commission will be
guided by the considerations which
govern the issuance of initial licenses to
the extent applicable and appropriate. If
the application is for amendment of an
OL or COL and involves the material
alteration of a commercial nuclear plant,
a construction permit (CP) will be
issued before the issuance of the
amendment to the license, provided
however, that if the application involves
a material alteration to a manufactured
reactor under this part before its
installation at a site, or a COL before the
date that the Commission makes the
finding under § 53.1452(g), no
application for or issuance of a CP is
required. If the amendment involves a
significant hazards consideration, the
Commission will give notice of its
proposed action—
(1) Under § 2.105 of this chapter
before acting thereon; and
(2) As soon as practicable after the
application has been docketed.
(b) The Commission will be
particularly sensitive to a license
amendment request that involves
irreversible consequences (such as one
that permits a significant increase in the
amount of effluents or radiation emitted
by a commercial nuclear plant).
(c) The Commission may make a final
determination, under the procedures in
§ 53.1515, that a proposed amendment
to an OL or a COL for a commercial
nuclear plant under this part involves
no significant hazards consideration, if
operation of the plant in accordance
with the proposed amendment would
not—
(1) Involve a significant increase in
the probability or consequences of an
accident previously evaluated; or
(2) Create the possibility of a new or
different kind of an accident from any
accident previously evaluated; or
(3) Involve a significant reduction in
a margin of safety.
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§ 53.1525 Revising certification
information within a design certification
rule.
(a) A holder of an OL or COL who
references a design certification rule
issued under this part must request an
exemption if proposing to change one or
more elements of the certification
information. The Commission may grant
such a request only if it determines that
the exemption will comply with the
requirements of § 53.080 and that the
special circumstances outweigh any
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decrease in safety that may result from
the reduction in standardization caused
by the departure.
(b) The request for an exemption must
be included with any associated license
amendment request, which must be
requested and processed in accordance
with §§ 53.1510, 53.1515, and 53.1520.
(c) Licensees must evaluate changes to
the design as described in the Final
Safety Analysis Report (FSAR) not
involving changes to the certification
information using the criteria in
§ 53.1550.
§ 53.1530 Revising design information
within a manufacturing license.
(a) The holder of an ML may not make
changes to the design of the
manufactured reactor authorized to be
manufactured without obtaining an
amendment pursuant to § 53.1510 and,
as applicable, § 53.1520.
(b) The holder of a COL under this
part who references or uses a
manufactured reactor under this part
must request approval for any proposed
departure from the design
characteristics, site parameters, terms
and conditions, or approved design of
the manufactured reactor. The
application for such departures must be
submitted and processed in accordance
with §§ 53.1510, 53.1515, and 53.1520.
In those cases where an ML references
a design certification rule, the
amendment application from the holder
of the COL must also request an
exemption from the design certification
rule under § 53.1525 if one or more
elements of the certification information
are adversely affected by the proposed
change. The holder of the COL must
evaluate changes to the commercial
nuclear plant as described in the FSAR
but outside of the scope of the
referenced ML using the criteria in
§ 53.1550.
§ 53.1535 Amendments during
construction.
(a) The holder of a CP or limited work
authorization (LWA) under this part
may request an amendment to the CP or
LWA in order to gain Commission
approval of the safety of selected design
features or specifications, including
proposed departures from a design
certification rule or ML. Amendments to
CPs or LWAs under this part must be
requested and processed under
§§ 53.1510 and 53.1520.
(b) The holder of a COL under this
part for which the NRC has not yet
made a finding in accordance with
§ 53.1452(g) must request amendments
required by § 53.1525 or § 53.1550 no
later than 45 days from the date the
licensee begins the construction of the
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SSCs to implement the change or
departure requiring NRC approval. The
licensee proceeds with such changes at
its own risk recognizing that there is a
possibility that the amendment will not
be granted.
§ 53.1540 Updating licensing-basis
information and determining the need for
NRC approval.
(a) Sections 53.1545 through 53.1565
provide the process for a holder of an
OL or COL to modify licensing-basis
information and to evaluate potential
changes to its facilities, procedures,
programs, and organizations to
determine if NRC approval is required.
(b) Definitions for the purposes of
§§ 53.1545 through 53.1565—
Change means a modification or
addition to, or removal from, the
commercial nuclear plant or procedures
that affects a design feature or related
functional design criteria, method of
performing or controlling the functions
of design features, or an evaluation that
demonstrates that intended functions
will be accomplished.
Departure from a method of
evaluation described in the Final Safety
Analysis Report (FSAR) (as updated)
used in establishing the functional
design criteria for safety-related
structures, systems, or components or in
the safety analyses means—
(1) Changing any of the elements of
the method described in the FSAR (as
updated) unless the results of the
analysis are conservative or essentially
the same; or
(2) Changing from a method described
in the FSAR to another method unless
that method has been approved by NRC
for the intended application.
Facility as described in the FSAR (as
updated) means—
(1) The SSCs that are described in the
FSAR (as updated),
(2) The design and performance
requirements for such SSCs described in
the FSAR (as updated), and
(3) The evaluations or methods of
evaluation included in the FSAR (as
updated) for such SSCs which
demonstrate that their intended
function(s) will be accomplished.
Final Safety Analysis Report (as
updated) means the FSAR submitted
under § 53.1369 or § 53.1416, as
amended and supplemented, and as
updated under § 53.1545, as applicable.
Procedures as described in the Final
Safety Analysis Report (as updated)
means those procedures that contain
information described in the FSAR (as
updated) such as how SSCs are operated
and controlled (including assumed
operator actions and response times).
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Updating Final Safety Analysis
(a) Each holder of an OL or COL
under this part for which the
Commission has made the finding under
§ 53.1452(g) must update the FSAR
originally submitted as part of the
application for the license every 24
months or more frequently to assure that
the information included in the report
contains the latest information
developed. The submittal must include
the effects on the content of the FSAR
of—
(1) Changes made to the facility or
procedures as described in the FSAR;
(2) Safety analyses and evaluations
performed by the licensee either in
support of approved license
amendments or in support of
conclusions that changes did not require
a license amendment under § 53.1550;
(3) Updates to the probabilistic risk
assessments required under § 53.450;
(4) The cumulative effects of the
changes to the facility or procedures on
the margins to the safety criteria in
§§ 53.210, 53.220, 53.450(e), and 53.470
since the last FSAR update; and
(5) Analyses of new safety issues
performed by or on behalf of the
licensee at Commission request.
(b)(1) The licensee must submit
revisions containing updated
information to the Commission, under
§ 53.040, identifying the location of
revised or new information.
(2) The submittal must include—
(i) A certification by a duly authorized
officer of the licensee that either the
information accurately presents changes
made since the previous submittal,
necessary to reflect information and
analyses submitted to the Commission
or prepared pursuant to Commission
requirement, or that no such changes
were made; and
(ii) An identification of changes made
under the provisions of § 53.1550 but
not previously submitted to the
Commission.
(c) Each applicant for or holder of a
COL under this part for which the
Commission has not made the finding
under § 53.1452(g) must submit an
update to the FSAR annually by
providing the information required in
(a)(1) through (a)(5) of this section and
meeting the requirements of paragraph
(b) of this section. Combined license
applicants who have requested the NRC
to suspend its review of the COL
application and COL holders who have
informed the NRC that they do not plan
to pursue construction need not submit
an annual update of the FSAR. If a COL
applicant requests that the NRC resume
its review, or a COL holder notifies the
NRC that the COL holder plans to
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commence or resume construction, then
the COL applicant or holder must
submit to NRC an update to its FSAR
within 90 days of the request or
notification, as applicable, and annually
thereafter.
(d) The FSAR (as updated) must be
retained by the licensee until the
Commission terminates its license.
(e) Each holder of an ML under this
part must submit an update of the FSAR
reflecting any modification to the design
that is directed or approved by the
Commission under § 53.1288 or
§ 53.1530, and any new analyses of the
design requested by the Commission
under § 53.1580.
§ 53.1550 Evaluating changes to facility as
described in Final Safety Analysis Reports.
(a) The holder of an OL or COL may
make changes in the facility as
described in the FSAR (as updated) and
make changes in the procedures as
described in the FSAR (as updated)
without obtaining a license amendment
pursuant to § 53.1510 only if—
(1) A change to the technical
specifications incorporated in the
license is not required; and
(2) The change meets all of the
following criteria:
(i) Does not result in an increase to
the frequency or consequences of an
event sequence such that an event
sequence not previously identified as
risk significant becomes risk significant
by the analyses performed in
accordance with § 53.450(e).
(ii) Does not result in an increase to
the frequency or consequences of an
event sequence such that an event
sequence identified as risk significant in
accordance with § 53.450(e) exceeds the
licensing-basis event evaluation criteria
required to be established in accordance
with § 53.450(e).
(iii) Does not involve either of the
following: (A) a change to the NRCapproved comprehensive risk metric(s)
or associated risk performance objective
under § 53.220(b), or (B) an increase to
the frequency or consequences of one or
more event sequences such that there is
more than a minimal reduction in the
margin between the calculated
comprehensive risks posed by the
commercial nuclear plant and the safety
criteria of § 53.220.
(iv) Does not involve a departure from
a method of evaluation described in the
FSAR (as updated) used in assessing
licensing-basis events in accordance
with § 53.450 unless the results of the
analysis under § 53.450 are conservative
or essentially the same, the revised
method of evaluation has been
previously approved by the NRC for the
intended application, or the revised
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method of evaluation can be used under
an NRC-endorsed consensus code or
standard.
(v) Does not result in the escalation in
the safety classification of an SSC from
non-safety-related to non-safety-related
but safety-significant or from non-safetyrelated but safety-significant to safetyrelated.
(vi) Does not result in more than a
minimal decrease in defense in depth.
(vii) For commercial nuclear plants
licensed under this part for which
alternative evaluation criteria are
adopted in accordance with § 53.470,
does not result in a change to the
frequency or consequences of event
sequences such that the calculated
margins between the results for event
sequences evaluated in accordance with
§ 53.450(e) and the alternative
evaluation criteria decreases by 25
percent or more.
(viii) Does not result in the
identification of a new design-basis
accident in accordance with § 53.450(f).
(ix) Does not result in a decrease by
10 percent or more in the margin
between the consequence of any designbasis accident and the safety criteria in
§ 53.210.
(x) Does not prevent meeting the
design requirements in § 53.440(j) to
limit the release of radionuclides from
reactor systems, waste stores, or other
significant inventories of radioactive
materials assuming the impact of a
large, commercial aircraft.
(3) In implementing this paragraph,
the FSAR (as updated) is considered to
include FSAR changes since submittal
of the last update of the FSAR under
§ 53.1545.
(4) The provisions in this section do
not apply to changes to the facility or
procedures when the applicable
regulations establish more specific
criteria for accomplishing such changes.
(b)(1) A licensee who references a
design certification rule may make
departures from the standard design,
without prior Commission approval,
unless the proposed departure involves
a change to the design as described in
the rule certifying the design, in which
case the requirements of § 53.1525 are
applicable.
(2) The licensee must maintain
records of all departures from the
certified design of the facility and these
records must be maintained and
available for audit until the termination
of the license. The licensee must
identify the location and nature of
departures from licensing-basis
information within supporting
documents for a certified design within
the updates to the Safety Analysis
Report required by § 53.1545.
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(3) Licensees for which the NRC has
docketed the certifications required
under § 53.1070 need not retain records
of departures from the design of the
facility associated with SSCs that have
been permanently removed from service
using an NRC-approved change process.
(c)(1) The licensee must maintain
records of changes in the facility and
procedures made under paragraph (a) of
this section. These records must include
a written evaluation which provides the
bases for the determination that the
change does not require a license
amendment under paragraph (a)(2) of
this section.
(2) The licensee must submit, as
specified in § 53.040, a report
containing a brief description of any
departures and changes, including a
summary of the evaluation of each. A
report must be submitted at intervals
not to exceed 24 months. For COLs, the
report must be submitted at intervals
not to exceed 6 months during the
period from the date of application for
a COL to the date the Commission
makes its findings under § 53.1452(g).
(3) The records of changes in the
facility must be maintained until the
termination of an OL or COL issued
under this part, or the termination of a
renewed license issued under
§ 53.1595—whichever is later. Records
of changes in procedures must be
maintained for a period of 5 years.
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§ 53.1560 Updating program documents
included in licensing-basis information.
(a) Each holder under this part of an
OL or COL for which the Commission
has made the finding under § 53.1452(g)
must biennially or more frequently
update the program documents
submitted as part of an application to
obtain or maintain the license to assure
that the information included in the
documents contains the latest
information developed. The submittals
must include the effects on the content
of the program documents of—
(1) Changes made in the facility,
procedures, licensee’s organization, or
site environs;
(2) Safety analyses and evaluations
performed by the applicant or licensee
either in support of approved license
amendments or in support of
conclusions that changes did not require
a license amendment in accordance
with § 53.1550;
(3) Analyses of new safety issues
performed by or on behalf of the
licensee at Commission request; and
(4) Changes to the programs as a result
of operating experience, corrective
actions, or other reasons deemed
appropriate to ensure the programs
serve their underlying purpose to
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support the requirements in subpart B of
this part or other NRC regulations.
(b)(1) The licensee must submit
revisions containing updated
information to the Commission, as
specified in § 53.040, identifying the
location of revised or new information.
(2) The submittal must include—
(i) A certification by a duly authorized
officer of the licensee that either the
information accurately presents changes
made since the previous submittals,
necessary to reflect information and
analyses submitted to the Commission
or prepared pursuant to Commission
requirement, or that no such changes
were made; and
(ii) An identification of changes made
under the provisions of § 53.1550 but
not previously submitted to the
Commission.
(c) The updated program documents
must be retained by the licensee until
the Commission terminates their
license.
§ 53.1565 Evaluating changes to programs
included in licensing-basis information.
(a) A licensee may make changes to
the facility, procedures, or organizations
or address changes to site environs as
described in the program documents
included in licensing-basis information
without obtaining prior NRC approval
only if—
(1) A change to the technical
specifications incorporated in the
license is not required;
(2) An exemption from an NRC
regulation is not required; and
(3) The change conforms to programspecific requirements included in
regulations in this part, technical
specifications, or the NRC-approved
program document included and
reviewed as part of a license application
under subpart H or an amendment
under this subpart.
(b) In implementing this section, the
program documents (as updated)
include changes since submittal of the
last updates of the program documents
pursuant to § 53.1560.
(c) The provisions in this section do
not apply to changes to the program
documents when the applicable
regulations establish more specific
criteria for accomplishing such changes.
(d) To make changes to the facility,
procedures, or organizations or to
address changes to site environs as
described in the program documents
included in licensing-basis information
for individual programs, the following
requirements must be satisfied:
(1) Quality assurance program—
operation. (i) Each holder under this
part of an OL or COL, after the
Commission makes the finding under
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§ 53.1452(g), may make a change to a
previously accepted quality assurance
program (QAP) description included or
referenced in the Safety Analysis Report
without prior NRC approval, provided
the change does not reduce the
commitments in the program
description as accepted by the NRC.
Changes to the QAP description that do
not reduce the commitments must be
submitted to the NRC in accordance
with the requirements of § 53.1545. In
addition to QAP changes involving
administrative improvements and
clarifications, spelling corrections,
punctuation, or editorial items, the
following changes are not considered to
be reductions in commitment:
(A) The use of a quality assurance
(QA) standard approved by the NRC
which is more recent than the QA
standard in the licensee’s QAP at the
time of the change;
(B) The use of a QA alternative or
exception approved by an NRC safety
evaluation, provided that the bases of
the NRC approval are applicable to the
licensee’s facility;
(C) The use of generic organizational
position titles that clearly denote the
position function, supplemented as
necessary by descriptive text, rather
than specific titles;
(D) The use of generic organizational
charts to indicate functional
relationships, authorities, and
responsibilities, or, alternately, the use
of descriptive text;
(E) The elimination of QAP
information that duplicates language in
QA regulatory guides and QA standards
to which the licensee is committed; and
(F) Organizational revisions that
ensure that persons and organizations
performing QA functions continue to
have the requisite authority and
organizational freedom, including
sufficient independence from cost and
schedule when opposed to safety
considerations.
(ii) Changes to the QAP description
that do reduce the commitments must
be submitted to the NRC and receive
NRC approval prior to implementation,
as follows:
(A) Changes made to the QAP
description as presented in the Safety
Analysis Report or in a topical report
must be submitted as specified in
§ 53.040.
(B) The submittal of a change to the
Safety Analysis Report QAP description
must include all pages affected by that
change and must be accompanied by a
forwarding letter identifying the change,
the reason for the change, and the basis
for concluding that the revised program
incorporating the change continues to
satisfy the criteria of appendix B to part
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50 of this chapter and the Safety
Analysis Report QAP description
commitments previously accepted by
the NRC (the letter need not provide the
basis for changes that correct spelling,
punctuation, or editorial items).
(C) A copy of the forwarding letter
identifying the change must be
maintained as a facility record for 3
years.
(D) Changes to the QAP description
included or referenced in the Safety
Analysis Report shall be regarded as
accepted by the Commission upon
receipt of a letter to this effect from the
appropriate reviewing office of the
Commission or 60 days after submittal
to the Commission, whichever occurs
first.
(2) Quality assurance program—
siting, construction, and manufacturing.
Each holder of an LWA, early site
permit, CP, ML, or COL, before the
Commission makes the finding under
§ 53.1452(g) of this chapter, under this
part may make a change to a previously
accepted QAP description included or
referenced in the Safety Analysis Report
without prior NRC approval, provided
the change does not reduce the
commitments in the program
description previously accepted by the
NRC. Changes to the QAP description
that do not reduce the commitments
must be submitted to NRC within 90
days. Changes to the QAP description
that reduce the commitments must be
submitted to NRC and receive NRC
approval before implementation, as
follows:
(i) Changes to the Safety Analysis
Report must be submitted for review as
specified in § 53.040. Changes made to
NRC-accepted QA topical report
descriptions must be submitted as
specified in § 53.040.
(ii) The submittal of a change to the
Safety Analysis Report QAP description
must include all pages affected by that
change and must be accompanied by a
forwarding letter identifying the change,
the reason for the change, and the basis
for concluding that the revised program
incorporating the change continues to
satisfy the criteria of appendix B of part
50 of this chapter and the Safety
Analysis Report QAP description
commitments previously accepted by
the NRC (the letter need not provide the
basis for changes that correct spelling,
punctuation, or editorial items).
(iii) A copy of the forwarding letter
identifying the changes must be
maintained as a facility record for 3
years.
(iv) Changes to the QAP description
included or referenced in the Safety
Analysis Report shall be regarded as
accepted by the Commission upon
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receipt of a letter to this effect from the
appropriate reviewing office of the
Commission or 60 days after submittal
to the Commission, whichever occurs
first.
(3) Emergency preparedness program.
(i) Definitions for the purpose of
paragraph (d)(3) of this section:
(A) Change means an action that
results in modification or addition to, or
removal from, the licensee’s emergency
plan. All such changes are subject to the
provisions of this section except where
the applicable regulations establish
specific criteria for accomplishing a
particular change.
(B) Emergency plan means the
document(s), prepared and maintained
by the licensee, that identify and
describe the licensee’s methods for
maintaining emergency preparedness
and responding to emergencies. An
emergency plan includes the plan as
originally approved by the NRC and all
subsequent changes made by the
licensee with, and without, prior NRC
review and approval under paragraph
(d)(3) of this section.
(C) Emergency planning function
means a capability or resource necessary
to prepare for and respond to a
radiological emergency.
(D) Reduction in effectiveness means
a change in an emergency plan that
results in reducing the licensee’s
capability to perform an emergency
planning function in the event of a
radiological emergency.
(ii)(A) Except as provided in
paragraph (d)(3)(ii)(B) of this section, a
holder of an OL under this part, or a
COL under this part after the
Commission makes the finding under
§ 53.1452(g), must follow and maintain
the effectiveness of an emergency plan
that meets the requirements in appendix
E to part 50 of this chapter and the
planning standards of § 50.47(b).
(B) A holder of an OL under this part
for a commercial nuclear plant
consisting of small modular reactors
(SMRs) or non-light-water reactors, or a
holder of a COL under this part after the
Commission makes the finding under
§ 53.1452(g) for a commercial nuclear
plant consisting of either SMRs or nonlight-water reactors, must follow and
maintain the effectiveness of either an
emergency plan that meets the
requirements in § 50.160 or an
emergency plan that meets the
requirements in appendix E to part 50
of this chapter and the planning
standards of § 50.47(b).
(iii)(A) Except as provided in
paragraph (d)(3)(iii)(B) of this section,
the licensee may make changes to its
emergency plan without NRC approval
only if the licensee performs and retains
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an analysis demonstrating that the
changes do not reduce the effectiveness
of the plan and the plan, as changed,
continues to meet the requirements in
appendix E to part 50 of this chapter
and the planning standards of
§ 50.47(b).
(B) A license under this part for a
commercial nuclear plant consisting of
either SMRs or non-light-water reactors
may make changes to its emergency
plan without NRC approval only if the
licensee performs and retains an
analysis demonstrating that the changes
do not reduce the effectiveness of the
plan and the plan, as changed,
continues to meet either the
requirements in § 50.160 or the
requirements in appendix E to part 50
and the planning standards of
§ 50.47(b).
(iv) The changes to a licensee’s
emergency plan that reduce the
effectiveness of the plan as defined in
paragraph (d)(3)(i)(D) of this section
may not be implemented without prior
approval by the NRC. A licensee
desiring to make such a change must
submit an application for an
amendment to its license. In addition to
the filing requirements of §§ 53.1510
and 53.1515, the request must include
all emergency plan pages affected by
that change and must be accompanied
by a forwarding letter identifying the
change, the reason for the change, and
the basis for concluding that the
licensee’s emergency plan, as revised,
will continue to meet either the
requirements in § 50.160 to this chapter
or the requirements in appendix E to
part 50 of this chapter and the planning
standards of § 50.47(b) of this chapter.
(v) The licensee must retain a record
of each change to the emergency plan
made without prior NRC approval for a
period of three years from the date of
the change and shall submit, as
specified in § 53.040, a report of each
such change, including a summary of its
analysis, within 30 days after the change
is put in effect.
(vi) The licensee must retain the
emergency plan and each change for
which prior NRC approval was obtained
pursuant to paragraph (d)(3)(iv) of this
section as a record until the
Commission terminates the license for
the nuclear power reactor.
(vii)(A) The licensee must provide for
the development, revision,
implementation, and maintenance of its
emergency preparedness program. The
licensee must ensure that all program
elements are reviewed by persons who
have no direct responsibility for the
implementation of the emergency
preparedness program either—
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(1) At intervals not to exceed 12
months; or
(2) As necessary, based on an
assessment by the licensee against
performance indicators, and as soon as
reasonably practicable after a change
occurs in personnel, procedures,
equipment, or facilities that potentially
could adversely affect emergency
preparedness, but no longer than 12
months after the change. In any case, all
elements of the emergency preparedness
program must be reviewed at least once
every 24 months.
(B) The review must include an
evaluation for adequacy of interfaces
with State participating Tribal and local
governments and of licensee drills,
exercises, capabilities, and procedures.
The results of the review, along with
recommendations for improvements,
must be documented, reported to the
licensee’s corporate and plant
management, and retained for a period
of 5 years. The part of the review
involving the evaluation for adequacy of
interface with State, participating Tribal
and local governments must be available
to the appropriate State, participating
Tribal and local governments.
(4) Security programs. (i) The licensee
must prepare and maintain safeguards
contingency plan procedures in
accordance with appendix C of part 73
of this chapter for affecting the actions
and decisions contained in the
Responsibility Matrix of the safeguards
contingency plan. The licensee may not
make a change that would decrease the
safeguard effectiveness of a physical
security plan, or guard training and
qualification plan, or cybersecurity plan
submitted under subpart H or part 73 of
this chapter, or of the first four
categories of information (Background,
Generic Planning Base, Licensee
Planning Base, Responsibility Matrix)
contained in a licensee safeguards
contingency plan submitted under
subpart H or part 73 of this chapter, as
applicable, without prior approval of
the Commission. A licensee desiring to
make such a change must submit an
application for amendment to the
licensee’s license under §§ 53.1510,
53.1515, and 53.1520.
(ii) The licensee may make changes to
the plans referenced in paragraph (4)(i)
of this section without prior
Commission approval if the changes do
not decrease the safeguards
effectiveness of the plan. The licensee
must maintain records of changes to the
plans made without prior Commission
approval for a period of 3 years from the
date of the change, and must submit, as
specified in § 53.040, a report
containing a description of each change
within 2 months after the change is
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made. Prior to the safeguards
contingency plan being put into effect,
the licensee must have—
(A) All safeguards capabilities
specified in the safeguards contingency
plan available and functional;
(B) Detailed procedures developed
according to appendix C to part 73 of
this chapter available at the licensee’s
site; and
(C) All appropriate personnel trained
to respond to safeguards incidents as
outlined in the plan and specified in the
detailed procedures.
(iii) The licensee must provide for the
development, revision, implementation,
and maintenance of its safeguards
contingency plan. The licensee must
ensure that all program elements are
reviewed by individuals independent of
both security program management and
personnel who have direct
responsibility for implementation of the
security program either—
(A) At intervals not to exceed 12
months; or
(B) As necessary, based on an
assessment by the licensee against
performance indicators, and as soon as
reasonably practicable after a change
occurs in personnel, procedures,
equipment, or facilities that potentially
could adversely affect security, but no
longer than 12 months after the change.
In any case, all elements of the
safeguards contingency plan must be
reviewed at least once every 24 months.
(iv) The review must include a review
and audit of safeguards contingency
procedures and practices, an audit of
the security system testing and
maintenance program, and a test of the
safeguards systems along with
commitments established for response
by local law enforcement authorities.
The results of the review and audit,
along with recommendations for
improvements, must be documented,
reported to the licensee’s corporate and
plant management, and kept available at
the plant for inspection for a period of
3 years.
§ 53.1570
Transfer of licenses.
(a) No commercial nuclear plant
license issued under this part, or any
right thereunder, shall be transferred,
assigned, or in any manner disposed of,
either voluntarily or involuntarily,
directly or indirectly, through transfer of
control of the license to any person,
unless the Commission gives its consent
in writing.
(b)(1) An application for transfer of a
license must include—
(i) As much of the information
described in §§ 53.1109, 53.1306,
53.1366, and 53.1413 with respect to the
identity and technical and financial
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qualifications of the proposed transferee
as would be required by those sections
if the application were for an initial
license. The Commission may require
additional information such as data
respecting proposed safeguards against
hazards from radioactive materials and
the applicant’s qualifications to protect
against such hazards.
(ii) A statement of the purposes for
which the transfer of the license is
requested, the nature of the transaction
necessitating or making desirable the
transfer of the license, and an agreement
to limit access to Restricted Data or
Classified National Security Information
pursuant to § 53.1115. The Commission
may require any person who submits an
application for license pursuant to the
provisions of this section to file a
written consent from the existing
licensee or a certified copy of an order
or judgment of a court of competent
jurisdiction attesting to the person’s
right (subject to the licensing
requirements of the Act and these
regulations) to possession of the facility
or site involved.
(2) [Reserved]
(c) After appropriate notice to
interested persons, including the
existing licensee, and observance of
such procedures as may be required by
the Act or regulations or orders of the
Commission, the Commission will
approve an application for the transfer
of a license, if the Commission
determines—
(1) That the proposed transferee is
qualified to be the holder of the license;
and
(2) That transfer of the license is
otherwise consistent with applicable
provisions of law, regulations, and
orders issued by the Commission
pursuant thereto.
§ 53.1575
Termination of licenses.
(a) When the holder of an OL or COL
under this part has determined to
permanently cease operations the
licensee must, within 30 days, submit a
written certification to the NRC,
consistent with the requirements of
§ 53.1070.
(b) Once fuel has been permanently
removed from the reactor system, the
licensee must submit a written
certification to the NRC that meets the
requirements of § 53.1070.
(c)(1) Upon docketing of the
certifications for permanent cessation of
operations and permanent removal of
fuel from the reactor system, or when a
final legally effective order to
permanently cease operations has come
into effect, the license no longer
authorizes operation of the reactor or
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emplacement or retention of fuel into
the reactor system.
(2) Activities associated with
decommissioning will be carried out in
accordance with the requirements and
procedures in subpart G of this part.
(3) The Commission shall terminate
the license if it determines that—
(i) The remaining dismantlement has
been performed in accordance with the
approved license termination plan
required in subpart G of this part; and
(ii) The final radiation survey and
associated documentation, including an
assessment of dose contributions
associated with parts released for use
before approval of the license
termination plan, demonstrate that the
facility and site have met the criteria for
decommissioning in subpart E of 10
CFR part 20.
(d) A holder of a CP or COL under this
part may request the termination of the
license as well as licenses issued by the
NRC under parts 30, 40, or 70 of this
chapter prior to plant operations. Such
requests may support an immediate
NRC approval of the site for unrestricted
use.
§ 53.1580
Information requests.
Each licensee under this part must at
any time before termination of the
license, upon request of the
Commission, submit, as specified in
§ 53.040 written statements, signed
under oath or affirmation, to enable the
Commission to determine whether or
not the license should be modified,
suspended, or revoked. Except for
information sought to verify licensee
compliance with the current licensing
basis for that facility, the NRC must
prepare the reason or reasons for each
information request prior to issuance to
ensure that the burden to be imposed on
respondents is justified in view of the
potential safety significance of the issue
to be addressed in the requested
information. Each such justification
provided for an evaluation performed by
the NRC staff must be approved by the
Executive Director for Operations or his
or her designee prior to issuance of the
request.
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§ 53.1585 Revocation, suspension,
modification of licenses and approvals for
cause.
A license or standard design approval
issued under this part may be revoked,
suspended, or modified, in whole or in
part, for any material false statement in
the application or in the supplemental
or other statement of fact required of the
applicant; or because of conditions
revealed by the application or statement
of fact of any report, record, inspection,
or other means which would warrant
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the Commission to refuse to grant a
license or approval on an original
application; or for failure to
manufacture a reactor, or construct or
operate a facility in accordance with the
terms of the license, provided, however,
that failure to make timely completion
of the proposed construction or
alteration of a facility under a CP under
this part shall be governed by the
provisions of § 53.1342(b); or for
violation of, or failure to observe, any of
the terms and provisions of the Act,
regulations, license, approval, or order
of the Commission.
§ 53.1590
Backfitting.
(a)(1) Backfitting means the
modification of or addition to systems,
structures, components, or design of a
facility; or the design approval or ML for
a facility; or the procedures or
organization required to design,
construct or operate a facility; any of
which may result from a new or
amended provision in the Commission’s
regulations or the imposition of a
regulatory staff position interpreting the
Commission’s regulations that is either
new or different from a previously
applicable staff position after the date of
the commercial nuclear plant license
issued under this part.
(2) Except as provided in paragraph
(a)(4) of this section, the Commission
shall require a systematic and
documented analysis pursuant to
paragraph (b) of this section for backfits
which it seeks to impose.
(3) Except as provided in paragraph
(a)(4) of this section, the Commission
shall require the backfitting of a facility
only when it determines, based on the
analysis described in paragraph (b) of
this section, that there is a substantial
increase in the overall protection of the
public health and safety or the common
defense and security to be derived from
the backfit and that the direct and
indirect costs of implementation for that
facility are justified in view of this
increased protection.
(4) The provisions of paragraphs (a)(2)
and (a)(3) of this section are
inapplicable and, therefore, backfit
analysis is not required and the
standards in paragraph (a)(3) of this
section do not apply where the
Commission or staff, as appropriate,
finds and declares, with appropriate
documented evaluation for its finding,
either—
(i) That a modification is necessary to
bring a facility into compliance with a
license or the rules or orders of the
Commission, or into conformance with
written commitments by the licensee; or
(ii) That regulatory action is necessary
to ensure that the facility provides
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adequate protection to the health and
safety of the public and is in accord
with the common defense and security;
or
(iii) That the regulatory action
involves defining or redefining what
level of protection to the public health
and safety or common defense and
security should be regarded as adequate.
(5) The Commission must always
require the backfitting of a facility if it
determines that such regulatory action
is necessary to ensure that the facility
provides adequate protection to the
health and safety of the public and is in
accord with the common defense and
security.
(6) The documented evaluation
required by paragraph (a)(4) of this
section must include a statement of the
objectives of and reasons for the
modification and the basis for invoking
the exception. If immediately effective
regulatory action is required, then the
documented evaluation may follow
rather than precede the regulatory
action.
(7) If there are two or more ways to
achieve compliance with a license or
the rules or orders of the Commission,
or with written licensee commitments,
or there are two or more ways to reach
a level of protection which is adequate,
then ordinarily the applicant or licensee
is free to choose the way which best
suits its purposes. However, should it be
necessary or appropriate for the
Commission to prescribe a specific way
to comply with its requirements or to
achieve adequate protection, then cost
may be a factor in selecting the way,
provided that the objective of
compliance or adequate protection is
met.
(b) In reaching the determination
required by paragraph (a)(3) of this
section, the Commission will consider
how the backfit should be scheduled in
light of other ongoing regulatory
activities at the facility and, in addition,
will consider information available
concerning any of the following factors
as may be appropriate and any other
information relevant and material to the
proposed backfit:
(1) The statement of the specific
objectives that the proposed backfit is
designed to achieve;
(2) The general description of the
activity that would be required by the
licensee or applicant in order to
complete the backfit;
(3) The potential change in the risk to
the public from the accidental off-site
release of radioactive material;
(4) The potential impact on
radiological exposure of facility
employees;
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(5) The installation and continuing
costs associated with the backfit,
including the cost of facility downtime
or the cost of construction delay;
(6) The potential safety impact of
changes in plant or operational
complexity, including the relationship
to proposed and existing regulatory
requirements;
(7) The estimated resource burden on
the NRC associated with the proposed
backfit and the availability of such
resources;
(8) The potential impact of differences
in facility type, design or age on the
relevancy and practicality of the
proposed backfit;
(9) Whether the proposed backfit is
interim or final and, if interim, the
justification for imposing the proposed
backfit on an interim basis.
(c) No licensing action will be
withheld during the pendency of backfit
analyses required by the Commission’s
rules.
(d) The Executive Director for
Operations shall be responsible for
implementation of this section, and all
analyses required by this section shall
be approved by the Executive Director
for Operations or his or her designee.
§ 53.1595
Renewal.
Licenses may be renewed by the
Commission upon expiration of the
period of the license.
Subpart J—Reporting and Other
Administrative Requirements
§ 53.1600
General information.
Each applicant and licensee under
this part must ensure that U.S. Nuclear
Regulatory Commission (NRC)
inspectors have unfettered access to
sites and facilities licensed or proposed
to be licensed in § 53.1610, must
maintain records and make reports to
the NRC in accordance with
requirements in §§ 53.1620 through
53.1650, must satisfy financial
qualification and reporting requirements
in §§ 53.1660 through 53.1700, and
must obtain and maintain required
financial protections in case of an
accident in §§ 53.1720 and 53.1730.
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§ 53.1610 Unfettered access for
inspections.
(a) Each applicant for or holder of a
manufacturing license (ML), operating
license (OL), combined license (COL),
construction permit (CP), or early site
permit must permit inspection, by duly
authorized representatives of the
Commission, of its records, premises,
activities, and of licensed materials in
possession or use, related to the license
or CP or early site permit as may be
necessary to effectuate the purposes of
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the Atomic Energy Act of 1956, as
amended, (the Act) and the Energy
Reorganization Act of 1974, as
amended.
(b)(1) Each holder of an ML, OL, COL,
or CP must, upon request by the
Director, Office of Nuclear Reactor
Regulation, provide rent-free office
space for the exclusive use of the
Commission inspection personnel. Heat,
air conditioning, light, electrical outlets,
and janitorial services must be
furnished by each licensee and each
holder of a CP. The office must be
convenient to and have full access to the
facility and must provide the inspectors
both visual and acoustic privacy.
(2) For a site or facility with an
assigned resident inspector, the space
provided must be adequate to
accommodate a full-time inspector, a
part-time secretary, and transient NRC
personnel and must be generally
commensurate with other office
facilities at the site. For sites or facilities
assigned multiple resident inspectors,
additional space may be requested. The
office space that is provided must be
subject to the approval of the Director,
Office of Nuclear Reactor Regulation.
All furniture, supplies, and
communication equipment will be
furnished by the Commission.
(3) For a site or facility without an
assigned resident inspector, temporary
space to accommodate periodic or
special inspections must be provided.
The office space must be generally
commensurate with other office
accommodations at the site.
(4) The licensee or permit holder must
afford any NRC resident inspector
assigned to that site, or other NRC
inspectors identified by the Regional
Administrator as likely to inspect the
facility, immediate unfettered access,
equivalent to access provided regular
plant employees, following proper
identification and compliance with
applicable access control measures for
security, radiological protection, and
personal safety.
(5) The licensee or permit holder must
ensure that the arrival and presence of
an NRC inspector, who has been
properly authorized facility access as
described in paragraph (b)(4) of this
section, is not announced or otherwise
communicated by its employees or
contractors to other persons at the
facility unless specifically requested by
the NRC inspector.
§ 53.1620 Maintenance of records, making
of reports.
(a) Each holder of an ML, OL, COL,
CP, or early site permit must maintain
all records and make all reports, in
connection with the activity, as may be
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required by the conditions of the license
or permit or by the regulations and
orders of the Commission in effectuating
the purposes of the Act and the Energy
Reorganization Act of 1974, as
amended. Reports must be submitted in
accordance with § 53.040.
(b) [Reserved]
(c) Records that are required by the
regulations in this part, by license
condition, or by technical specifications
must be retained for the period specified
by the appropriate regulation, license
condition, or technical specification. If
a retention period is not otherwise
specified, these records must be
retained until the Commission
terminates the facility license or, in the
case of an early site permit, until the
permit expires.
(d)(1) Records which must be retained
under this part may be the original or a
reproduced copy or a microform if the
reproduced copy or microform is duly
authenticated by authorized personnel
and the microform is capable of
producing a clear and legible copy after
storage for the period specified by
Commission regulations. The record
may also be stored in electronic media
with the capability of producing legible,
accurate, and complete records during
the required retention period. Records
such as letters, drawings, and
specifications, must include all
pertinent information such as stamps,
initials, and signatures. The licensee
must maintain adequate safeguards
against tampering with, and loss of
records.
(2) If there is a conflict between the
Commission’s regulations in this part,
license condition, or technical
specification, or other written
Commission approval or authorization
pertaining to the retention period for the
same type of record, the retention
period specified in the regulations in
this part for such records shall apply
unless the Commission, under § 53.080
of this part, has granted a specific
exemption from the record retention
requirements in the regulations in this
part.
(e) Each licensee must notify the
Commission as specified in § 53.040 of
this part, of successfully completing
power ascension testing or startup
testing as applicable within 30 calendar
days of completing the testing.
§ 53.1630 Immediate notification
requirements for operating commercial
nuclear plants.
(a) General requirements.1 (1) Each
holder of an OL under this part or a COL
under this part after the Commission
makes the finding under § 53.1452(g),
must notify the NRC Operations Center
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via the Emergency Notification System
(ENS) of—
(i) The declaration of any of the
Emergency Classes specified in the
licensee’s approved Emergency Plan; or
(ii) Those non-emergency events
specified in paragraph (b) of this section
that occurred within 3 years of the date
of discovery.
(2) If the ENS is inoperative, the
licensee must make the required
notifications via commercial telephone
service, other dedicated telephone
system, or any other method which will
ensure that a report is made as soon as
practical to the NRC Headquarters
Operations Center at the numbers
specified in appendix A to part 73 of
this chapter.
(3) The licensee must notify the NRC
immediately after notification of the
appropriate State or local agencies and
not later than 1 hour after the time the
licensee declares one of the Emergency
Classes.
(4) The licensee must activate the data
links with the NRC as specified in their
emergency plans after declaring an
Emergency Class for events of actual or
potential substantial degradation of
plant safety or security, probable risk to
site personnel life, or site equipment
damage caused by hostile action. The
data links may also be activated by the
licensee during emergency drills or
exercises if the licensee’s computer
system has the capability to transmit the
exercise data.
(5) When making a report under
paragraph (a)(1) of this section, the
licensee must identify—
(i) The Emergency Class declared; or
(ii) Paragraph (b)(1), ‘‘One-hour
reports,’’ paragraph (b)(2), ‘‘Four-hour
reports,’’ or paragraph (b)(3), ‘‘Eighthour reports,’’ as the paragraph of this
section requiring notification of the nonemergency event.
(b) Non-emergency events. (1) Onehour reports. If not reported as a
declaration of an Emergency Class
under paragraph (a) of this section, the
licensee must notify the NRC as soon as
practical and in all cases within one
hour of the occurrence of any deviation
from the plant’s Technical
Specifications authorized under
§ 53.740(h) of this part.
(2) Four-hour reports. If not reported
under paragraphs (a) or (b)(1) of this
section, the licensee must notify the
NRC as soon as practical, and in all
cases, within 4 hours of the occurrence
of any of the following:
(i) The initiation of any commercial
nuclear plant shutdown required by the
plant’s Technical Specifications.
(ii) Any event or condition that results
in actuation of the reactor protection
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system when the reactor is critical
except when the actuation results from
and is part of a pre-planned sequence
during testing or reactor operation.
(iii) Any event or condition that
results in an unplanned actuation of a
safety-related (SR) standby cooling
system or the unplanned sole reliance
on an SR standby cooling system for
those systems that are in constant
operation.
(iv) Any event or condition that
results in an unplanned movement of,
change of state in, or chemical
interaction involving a significant
amount of radioactive material within
the commercial nuclear plant.
(v) Any event or situation, related to
the health and safety of the public or
onsite personnel, or protection of the
environment, for which a news release
is planned or notification to other
government agencies has been or will be
made. Such an event may include an
onsite fatality or inadvertent release of
radioactively contaminated materials.
(3) Eight-hour reports. If not reported
under paragraphs (a), (b)(1), or (b)(2) of
this section, the licensee must notify the
NRC as soon as practical and in all cases
within 8 hours of the occurrence of any
of the following:
(i) Any event or condition that results
in—
(A) The condition of the commercial
nuclear plant, including its principal
safety barriers, being seriously
degraded; or
(B) The commercial nuclear plant
being in a condition not analyzed under
§ 53.450 that significantly degrades
plant safety.
(ii) Any event or condition that results
in valid actuation of an SR system,
except when the actuation results from
and is part of a pre-planned sequence
during testing or reactor operation.
(iii) Any event or condition that at the
time of discovery could have prevented
the fulfilment of the safety functions
identified under § 53.230. Events
covered may include one or more
procedural errors, equipment failures,
and/or discovery of design, analysis,
fabrication, construction, and/or
procedural inadequacies. However,
individual component failures need not
be reported pursuant to this paragraph
if other equipment was operable and
available to perform the required safety
function.
(iv) Any event requiring the transport
of a radioactively contaminated person
to an offsite medical facility for
treatment.
(v) Any event that results in a major
loss of emergency assessment capability,
offsite response capability, or offsite
communications capability (e.g.,
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significant portion of control room
indication, ENS, or offsite notification
system).
(c) Follow-up notification: With
respect to the notifications made under
paragraphs (a) and (b) of this section, in
addition to making the required initial
notification, each licensee, must during
the course of the event—
(1) Immediately Report:
(i) any further degradation in the level
of safety of the plant or other worsening
plant conditions, including those that
require the declaration of any of the
Emergency Classes, if such a declaration
has not been previously made, or
(ii) any change from one Emergency
Class to another, or
(iii) a termination of the Emergency
Class.
(2) Immediately Report:
(i) the results of ensuing evaluations
or assessments of plant conditions,
(ii) the effectiveness of response or
protective measures taken, and
(iii) important information related to
plant behavior that is not understood.
(3) Maintain an open, continuous
communication channel with the NRC
Operation Center upon request by the
NRC.
1 Other requirements for immediate
notification of the NRC by licensed operating
commercial nuclear plants are contained
elsewhere in this chapter, in particular
§§ 20.1906, 20.2202, 72.216, 73.77, and
73.1200 of this chapter.
§ 53.1640
Licensee event report system.
(a) Reportable events. (1) Each
commercial nuclear plant licensee
holding an OL under this part or a COL
under this part after the Commission
makes the finding under § 53.1452(g),
must submit a Licensee Event Report
(LER) for any event of the type
described in this paragraph within 60
days after discovery of the event. In the
case of an invalid actuation reported
under § 53.1640(a)(2), other than
automatic reactor shutdown when the
reactor is critical, the licensee may, at
its option, provide a telephone
notification to the NRC Operations
Center within 60 days after discovery of
the event instead of submitting a written
LER. Unless otherwise specified in this
section, the licensee must report an
event if it occurred within 3 years of the
date of discovery regardless of the plant
mode or power level, and regardless of
the significance of the structure, system,
or component that initiated the event.
(2) The licensee must report—
(i)(A) The completion of any
commercial nuclear plant shutdown
required by the plant’s Technical
Specifications.
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(B) Any operation or condition which
was prohibited by the plant’s Technical
Specifications except when—
(1) The Technical Specification is
administrative in nature;
(2) The event consisted solely of a
case of a late surveillance test where the
oversight was corrected, the test was
performed, and the equipment was
found to be capable of performing its
specified safety functions; or
(3) The Technical Specification was
revised prior to discovery of the event
such that the operation or condition was
no longer prohibited at the time of the
event.
(C) Any deviation from the plant’s
Technical Specifications authorized
under § 53.740(h).
(ii) Any event or condition that
resulted in—
(A) The condition of the commercial
nuclear plant, including its principal
safety barriers, being seriously
degraded; or
(B) The commercial nuclear plant
being in a condition not analyzed under
§ 53.450 that significantly degrades
plant safety.
(iii) Any natural phenomena or other
external condition that posed an actual
threat to the safety of the commercial
nuclear plant or significantly hampered
site personnel in the performance of
duties necessary for the safe operation
of the commercial nuclear plant.
(iv) Any event or condition that
resulted in inadvertent operation of any
structures, systems, and component
classified as SR for an identified safety
function under § 53.460 or the
unplanned sole reliance on an SR
system for those systems that are in
constant operation, except when—
(A) The actuation resulted from and
was part of a pre-planned sequence
during testing; or
(B) The actuation was invalid and—
(1) Occurred while the system was
properly removed from service; or
(2) Occurred after the safety function
had been already completed.
(v) Any event or condition that could
have prevented the fulfillment of the
safety functions identified under
§ 53.230.
(vi) Events covered in paragraph
(a)(2)(v) of this section may include one
or more procedural errors, equipment
failures, and/or discovery of design,
fabrication, construction, and/or
procedural inadequacies. However,
individual component failures need not
be reported pursuant to paragraph
(a)(2)(v) of this section if any other
equipment was operable and available
to perform the required safety function.
(vii)(A) Any event or condition that as
a result of a single cause could have
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prevented the fulfillment of any of the
safety functions identified under
§ 53.230.
(B) Events covered in paragraph
(a)(2)(vii)(A) of this section may include
cases of procedural error, equipment
failure, and/or discovery of a design,
analysis, fabrication, construction,
and/or procedural inadequacy.
However, licensees are not required to
report an event pursuant to paragraph
(a)(2)(vii)(A) of this section if the event
results from—
(1) A shared dependency among
trains or channels that is a natural or
expected consequence of the approved
plant design; or
(2) Normal and expected wear or
degradation.
(viii)(A) Any airborne radioactive
release that, when averaged over a time
period of 1-hour, resulted in airborne
radionuclide concentrations in an
unrestricted area that exceeds 20 times
the applicable concentration limits
specified in appendix B to 10 CFR part
20, table 2, column 1.
(B) Any liquid effluent release that,
when averaged over a time period of
1-hour, exceeds 20 times the applicable
concentrations specified in appendix B
to 10 CFR part 20, table 2, column 2, at
the point of entry into the receiving
waters (i.e., unrestricted area) for all
radionuclides except tritium and
dissolved noble gases.
(ix) Any event that posed an actual
threat to the safety of the commercial
nuclear plant or significantly hampered
site personnel in the performance of
duties necessary for the safe operation
of the plant, including fires, toxic gas
releases, or radioactive releases.
(b) Contents. The LER must contain—
(1) A brief abstract describing the
major occurrences during the event,
including all component or system
failures that contributed to the event
and significant corrective action taken
or planned to prevent recurrence.
(2)(i) A clear, specific narrative
description of what occurred so that
knowledgeable readers conversant with
the design of commercial nuclear plants,
but not familiar with the details of a
particular plant, can understand the
complete event.
(ii) The narrative description must
include the following specific
information as appropriate for the
particular event:
(A) Plant operating conditions before
the event.
(B) Status of systems, structures, or
components that were inoperable at the
start of the event and that contributed to
the event.
(C) Dates and approximate time of the
occurrences.
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(D) The cause of each component or
system failure or personnel error, if
known.
(E) The failure mode, mechanism, and
effect of each failed component, if
known.
(F) [Reserved]
(G) For failures of components with
multiple functions, include a list of
systems or secondary functions that
were also affected.
(H) For failure that rendered a
component or system classified as SR or
non-safety-related but safety-significant
inoperable, an estimate of the elapsed
time from the discovery of the failure
until the component or system was
returned to service.
(I) The method of discovery of each
component or system failure or
procedural error.
(J) For each human performance
related root cause, the licensee must
discuss the cause(s) and circumstances.
(K) Automatically and manually
initiated safety system responses.
(L) The manufacturer and model
number (or other identification) of each
component that failed during the event.
(3) An assessment of the safety
consequences and implications of the
event. This assessment must include—
(i) The availability of systems or
components that could have performed
the same function as the components
and systems that failed during the event,
and
(ii) For events that occurred when the
reactor was shut down, the availability
of systems or components that are
needed to shut down the reactor and
maintain safe shutdown conditions,
remove residual heat, control the release
of radioactive material, or mitigate the
consequences of an accident.
(4) A description of any corrective
actions planned as a result of the event,
including those to reduce the
probability of similar events occurring
in the future.
(5) Reference to any previous similar
events at the same plant that are known
to the licensee.
(6) The name and contact information
of a person within the licensee’s
organization who is knowledgeable
about the event and can provide
additional information concerning the
event and the plant’s characteristics.
(c) Supplemental Information. The
Commission may require the licensee to
submit specific additional information
beyond that required by paragraph (b) of
this section if the Commission finds that
supplemental material is necessary for
complete understanding of an unusually
complex or significant event. These
requests for supplemental information
will be made in writing and the licensee
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must submit, as specified in § 53.040,
the requested information as a
supplement to the initial LER.
(d) Submission of Reports. Licensee
Event Reports must be prepared on
Form NRC 366 and submitted to the
NRC, as specified in § 53.040.
(e) Report Legibility. The reports and
copies that licensees are required to
submit to the Commission under the
provisions of this section must be of
sufficient quality to permit legible
reproduction and micrographic
processing.
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53.1645 Reports of radiation exposure to
members of the public.
(a) Each holder of an OL, and each
holder of a COL after the Commission
has made the finding under
§ 53.1452(g), must submit radiological
reports as required by 10 CFR part 20,
as well as an Annual Radioactive
Effluent Release Report and an Annual
Radiological Environmental Operating
Report. The Annual Radioactive
Effluent Release Report must specify the
quantity of each of the principal
radionuclides released to unrestricted
areas in liquid and in gaseous effluents
and an estimate of the dose received by
the maximally exposed member of the
public in an unrestricted area from
effluents and direct radiation from
contained sources during the previous
calendar year. The Annual Radiological
Environmental Operating Report must
provide data on measurable levels of
radiation and radioactive materials in
the environment, must include an
evaluation of the relationship between
quantities of radioactive material
released in effluents and resultant
radiation doses to individuals from
principal pathways of exposure, and
must include the results of
environmental monitoring during the
previous calendar year. These reports
must also include any other information
as may be required by the Commission
to estimate maximum potential annual
radiation doses to the public. The
reports must be submitted as specified
in § 53.040 by May 15 of each
successive year. If the total effective
dose equivalent to members of the
public in unrestricted areas during the
reporting period is greater than the as
low as is reasonably achievable
(ALARA) design objectives established
under § 53.425, the report must specify
the causes for exceeding the ALARA
design objective and describe any
corrective actions. On the basis of these
reports and any additional information
the Commission may obtain from the
licensee or others, the Commission may
require the licensee to take action as the
Commission deems appropriate.
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(b) If during any calendar quarter the
radiation exposure to a member of the
public in the unrestricted areas,
calculated on the same basis as the
respective ALARA design objective
exposure, exceeds one-half of the
annual ALARA design objective
exposure, the licensee must submit a
report as specified in § 53.040. The
report shall specify the causes for
exceeding one-half the annual ALARA
design objective exposure in a quarter
and describe corrective actions that the
licensee will take to maintain radiation
exposure to levels within the ALARA
design objectives for the remainder of
the year. The report shall be submitted
within 30 days from the end of the
quarter when one-half of the annual
ALARA design objective exposure was
exceeded.
§ 53.1650 Facility information and
verification.
(a) In response to a written request by
the Commission, each applicant for a CP
or license and each recipient of a CP or
a license must submit facility
information, as described in § 75.10 of
this chapter, on International Atomic
Energy Agency (IAEA) Design
Information Questionnaire forms and
site information on DOC/NRC Form AP–
A and associated forms;
(b) As required by the Additional
Protocol, must submit location
information described in § 75.11 of this
chapter on DOC/NRC Form AP–1 and
associated forms; and
(c) Must permit verification thereof by
the IAEA and take other action as
necessary to implement the US/IAEA
Safeguards Agreement, as described in
part 75 of this chapter.
§ 53.1660
Financial requirements.
Sections 53.1670 through 53.1700 set
out the requirements and procedures
related to financial qualifications and
related reporting requirements.
§ 53.1670
Financial qualifications.
Except for an electric utility applicant
for a license to operate a commercial
nuclear plant, an applicant for a CP, OL,
or COL under this part must possess or
have reasonable assurance of obtaining
the funds necessary for the activities for
which the permit or license is sought.
§ 53.1680
Annual financial reports.
With respect to any commercial
nuclear plant of a type described in
§ 53.020, each licensee and each holder
of a CP must submit its annual financial
report, including the certified financial
statements, to the Commission, as
specified in § 53.040, upon issuance of
the report. However, licensees and
holders of a CP who submit a Form 10–
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Q with the Securities and Exchange
Commission or a Form 1 with the
Federal Energy Regulatory Commission
need not submit the annual financial
report or the certified financial
statement under this section.
§ 53.1690 Licensee’s change of status;
financial qualifications.
(a) An electric utility licensee holding
an OL or COL (including a renewed
license) for a commercial nuclear plant,
no later than seventy-five (75) days prior
to ceasing to be an electric utility in any
manner not involving a license transfer
under § 53.1399 or § 53.1456 must
provide the NRC with the financial
qualifications information that would be
required for obtaining an initial OL or
COL under this part. The financial
qualifications information must address
the first full 5 years of operation after
the date the licensee ceases to be an
electric utility.
(b)(1) Any holder of a license issued
under this part must notify the
appropriate NRC Regional
Administrator, in writing, immediately
following the filing of a voluntary or
involuntary petition for bankruptcy
under any chapter of title 11
(Bankruptcy) of the United States Code
by or against—
(i) The licensee;
(ii) An entity (as 11 U.S.C. 101(14)
defines that term) controlling the
licensee or listing the license or licensee
as property of the estate; or
(iii) An affiliate (as 11 U.S.C. 101(2)
defines that term) of the licensee.
(2) This notification must indicate—
(i) The bankruptcy court in which the
petition for bankruptcy was filed; and
(ii) The date of the filing of the
petition.
§ 53.1700
Creditor regulations.
(a) Pursuant to section 184 of the Act,
the Commission consents, without
individual application, to the creation of
any mortgage, pledge, or other lien upon
any facility not owned by the United
States which is the subject of a license
or upon any leasehold or other interest
in such facility; provided—
(1) That the rights of any creditor so
secured may be exercised only in
compliance with and subject to the
same requirements and restrictions as
would apply to the licensee pursuant to
the provisions of the license, the Act,
and regulations issued by the
Commission under the Act; and
(2) That no creditor so secured may
take possession of the facility pursuant
to the provisions of this section prior to
either the issuance of a license from the
Commission authorizing such
possession or the transfer of the license.
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(b) Any creditor so secured may apply
for transfer of the license covering such
facility by filing an application for
transfer of the license under § 53.1570.
The Commission will act upon such
application under subpart I of this part.
(c) Nothing contained in this
regulation shall be deemed to affect the
means of acquiring, or the priority of,
any tax lien or other lien provided by
law.
(d) As used in this section—
License includes any license under
this part, which may be issued by the
Commission with regard to a facility.
Creditor includes, without implied
limitation, the trustee under any
mortgage, pledge or lien on a facility
made to secure any creditor, any trustee
or receiver of the facility appointed by
a court of competent jurisdiction in any
action brought for the benefit of any
creditor secured by such mortgage,
pledge or lien, any purchaser of such
facility at the sale thereof upon
foreclosure of such mortgage, pledge, or
lien or upon exercise of any power of
sale contained therein, or any assignee
of any such purchaser.
Facility includes, but is not limited to,
a site which is the subject of an early
site permit under this part, and a reactor
manufactured under an ML under this
part.
§ 53.1710
Financial protection.
Sections 53.1720 and 53.1730 set out
the requirements and procedures related
to licensees obtaining and maintaining
insurance to cover stabilization and
decontamination activities in the event
of an accident and financial protection
in accordance with part 140, ‘‘Financial
Protection Requirements and Indemnity
Agreements,’’ of this chapter.
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§ 53.1720 Insurance required to stabilize
and decontaminate plant following an
accident.
Each commercial nuclear plant
licensee under this part must take
reasonable steps to obtain insurance
available at reasonable costs and on
reasonable terms from private sources or
to demonstrate that it possesses an
equivalent amount of protection
covering the licensee’s obligation, in the
event of an accident at the licensee’s
commercial nuclear reactor, to stabilize
and decontaminate the plant and the
plant site at which such an accident
may occur, provided that—
(a) The insurance required by this
section must have a minimum coverage
limit for each commercial nuclear plant
site of $1.06 billion, an amount based on
plant-specific estimates of costs to
stabilize and decontaminate a plant, or
whatever amount of insurance is
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generally available from private sources,
whichever is less. The required
insurance must clearly state that, as and
to the extent provided in paragraph
(d)(1) of this section, any proceeds must
be payable first for stabilization of the
plant and next for decontamination of
the plant and the plant site. If a
licensee’s coverage falls below the
required minimum, the licensee must
within 60 days take all reasonable steps
to restore its coverage to the required
minimum. The required insurance may,
at the option of the licensee, be
included within policies that also
provide coverage for other risks,
including, but not limited to, the risk of
direct physical damage.
(b)(1) With respect to policies issued
or annually renewed, the proceeds of
such required insurance must be
dedicated, as and to the extent provided
in this paragraph, to reimbursement or
payment on behalf of the insured of
reasonable expenses incurred or
estimated to be incurred by the licensee
in taking action to fulfill the licensee’s
obligation, in the event of an accident at
the licensee’s plant, to ensure that the
plant is in, or is returned to, and
maintained in, a safe and stable
condition and that radioactive
contamination is removed or controlled
such that personnel exposures are
consistent with the occupational
exposure limits in 10 CFR part 20.
These actions must be consistent with
any other obligation the licensee may
have under this chapter and must be
subject to paragraph (d) of this section.
As used in this section, an ‘‘accident’’
means an event that involves the release
of radioactive material from its intended
place of confinement within the
commercial nuclear plant such that
there is a present danger of release off
site in amounts that would pose a threat
to the public health and safety.
(2) The stabilization and
decontamination requirements set forth
in paragraph (d) of this section must
apply uniformly to all insurance
policies required under this section.
(c) The licensee shall report to the
NRC on April 1 of each year the current
levels of this insurance or financial
security it maintains and the sources of
this insurance or financial security.
(d)(1) In the event of an accident at
the licensee’s plant, whenever the
estimated costs of stabilizing the
licensed plant and of decontaminating
the plant and the plant site exceed one
tenth of the minimum insurance under
paragraph (a) of this section, the
proceeds of the insurance required by
this section must be dedicated to and
used, first, to ensure that the licensed
plant is in, or is returned to, and can be
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maintained in, a safe and stable
condition so as to prevent any
significant risk to the public health and
safety and, second, to decontaminate the
plant and the plant site in accordance
with the licensee’s cleanup plan as
approved by order of the Director, Office
of Nuclear Reactor Regulation. This
priority on insurance proceeds must
remain in effect for 60 days or, upon
order of the Director, for such longer
periods, in increments not to exceed 60
days except as provided for activities
under the cleanup plan required in
paragraphs (d)(3) and (d)(4) of this
section, as the Director may find
necessary to protect the public health
and safety. Actions needed to bring the
plant to and maintain the plant in a safe
and stable condition may include one or
more of the following, as appropriate:
(i) Shutdown of the reactor(s) and
other processes at the plant;
(ii) Establishment and maintenance of
long-term cooling with stable decay heat
removal;
(iii) Maintenance of sub-criticality;
(iv) Control of radioactive releases;
and
(v) Securing of structures, systems, or
components to minimize radiation
exposure to onsite personnel or to the
offsite public or to facilitate later
decontamination or both.
(2) The licensee must inform the
Director, Office of Nuclear Reactor
Regulation in writing when the plant is
and can be maintained in a safe and
stable condition so as to prevent any
significant risk to the public health and
safety. Within 30 days after the licensee
informs the Director that the plant is in
this condition, or at such earlier time as
the licensee may elect or the Director
may for good cause direct, the licensee
must prepare and submit a cleanup plan
for the Director’s approval. The cleanup
plan must identify and contain an
estimate of the cost of each cleanup
operation that will be required to
decontaminate the reactor sufficiently to
permit the licensee either to resume
operation of the reactor or to apply to
the Commission under subpart G of this
part for authority to decommission the
reactor and to surrender the license
voluntarily. Cleanup operations may
include one or more of the following, as
appropriate:
(i) Processing any contaminated
materials generated by the accident and
by decontamination operations to
remove radioactive materials;
(ii) Decontamination of surfaces
inside the plant buildings to levels
consistent with the Commission’s
occupational exposure limits in 10 CFR
part 20, and decontamination or
disposal of equipment;
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(iii) Decontamination or removal and
disposal of internal parts, damaged fuel
from the reactor coolant or fuel systems,
or related process or waste systems; and
(iv) Cleanup of the reactor coolant or
fuel systems or related process or waste
systems.
(3) Following review of the licensee’s
cleanup plan, the Director will order the
licensee to complete all operations that
the Director finds are necessary to
decontaminate the reactor sufficiently to
permit the licensee either to resume
operation of the reactor or to apply to
the Commission under subpart G of this
part for authority to decommission the
reactor and to surrender the license
voluntarily. The Director must approve
or disapprove, in whole or in part for
stated reasons, the licensee’s estimate of
cleanup costs for such operations. Such
order may not be effective for more than
one year, at which time it may be
renewed. Each subsequent renewal
order, if imposed, may be effective for
not more than 6 months.
(4) Of the balance of the proceeds of
the required insurance not already
expended to place the plant in a safe
and stable condition under paragraph
(b)(1) of this section, an amount
sufficient to cover the expenses of
completion of those decontamination
operations that are the subject of the
Director’s order must be dedicated to
such use, provided that, upon
certification to the Director of the
amounts expended previously and from
time to time for stabilization and
decontamination and upon further
certification to the Director as to the
sufficiency of the dedicated amount
remaining, policies of insurance may
provide for payment to the licensee or
other loss payees of amounts not so
dedicated, and the licensee may proceed
to use in parallel (and not in preference
thereto) any insurance proceeds not so
dedicated for other purposes.
§ 53.1730 Financial protection
requirements.
Commercial nuclear plant licensees
must satisfy the applicable provisions of
part 140, ‘‘Financial Protection
Requirements and Indemnity
Agreements,’’ of this chapter.
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Subparts K and L [Reserved]
Subpart M—Enforcement
§ 53.9000
Violations.
(a) The Commission may obtain an
injunction or other court order to
prevent a violation of the provisions
of—
(1) The Atomic Energy Act of 1954, as
amended (the Act);
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(2) Title II of the Energy
Reorganization Act of 1974, as
amended; or
(3) A regulation or order issued under
those Acts.
(b) The Commission may obtain a
court order for the payment of a civil
penalty imposed under Section 234 of
the Act:
(1) For violations of—
(i) Sections 53, 57, 62, 63, 81, 82, 101,
103, 104, 107, or 109 of the Act;
(ii) Section 206 of the Energy
Reorganization Act of 1974, as
amended;
(iii) Any rule, regulation, or order
issued under the sections specified in
paragraph (b)(1)(i) of this section;
(iv) Any term, condition, or limitation
of any license issued under the sections
specified in paragraph (b)(1)(i) of this
section.
(2) For any violation for which a
license may be revoked under section
186 of the Act.
§ 53.9010
Criminal penalties.
(a) Section 223 of the Act provides for
criminal sanctions for willful violation
of, attempted violation of, or conspiracy
to violate, any regulation issued under
sections 161b, 161i, or 161o of the Act.
For purposes of section 223, all the
regulations in part 53 are issued under
one or more of sections 161b, 161i, or
161o, except for the sections listed in
paragraph (b) of this section.
(b) The regulations in 10 CFR part 53
that are not issued under sections 161b,
161i, or 161o for the purposes of section
223 are as follows: §§ 53.000, 53.015,
53.020, 53.040, 53.080, 53.090, 53.100,
53.110, 53.120, 53.600, 53.725, 53.726,
53.735, 53.760, 53.775, 53.790, 53.795,
53.820, 53.910, 53.1000, 53.1050,
53.1100, 53.1103, 53.1106, 53.1109,
53.1112, 53.1115, 53.1118, 53.1120,
53.1121, 53.1124, 53.1140, 53.1143,
53.1144, 53.1146, 53.1149, 53.1155,
53.1158, 53.1164, 53.1170, 53.1173,
53.1176, 53.1179, 53.1188, 53.1200,
53.1203, 53.1206, 53.1209, 53.1210,
53.1212, 53.1215, 53.1218, 53.1221,
53.1230, 53.1236, 53.1239, 53.1241,
53.1242, 53.1245, 53.1248, 53.1251,
53.1254, 52.1257, 52.1260, 53.1263,
53.1270, 53.1273, 53.1276, 53.1279,
53.1282, 53.1285, 53.1286, 53.1287,
53.1288, 53.1291, 53.1293, 53.125,
53.1300, 53.1306, 53.1309, 53.1312,
53.1315, 53.1318, 53.1324, 53.1330,
53.1333, 53.1336, 53.1348, 53.1360,
53.1366, 53.1369, 53.1372, 53.1375,
53.1381, 53.1384, 53.1387, 53.1390,
53.1396, 53.1401, 53.1405, 53.1410,
53.1416, 53.1419, 53.1422, 53.1425,
53.1431, 53.1437, 53.1440, 53.1443,
53.1452, 53.1455, 53.1456, 53.1458,
53.1461, 53.1470, 53.1500, 53.1510,
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53.1515, 53.1520, 53.1525, 53.1530,
53.1535, 53.1540, 53.1560, 53.1585,
53.1590, 53.1595, 53.1600, 53.1660,
53.1670, 53.1700, 53.1710, 53.1730,
53.9000, 53.9010.
PART 70—DOMESTIC LICENSING OF
SPECIAL NUCLEAR MATERIAL
129. The authority citation for 10 CFR
part 70 continues to read as follows:
■
Authority: Atomic Energy Act of 1954,
secs. 51, 53, 57(d), 108, 122, 161, 182, 183,
184, 186, 187, 193, 223, 234, 274, 1701 (42
U.S.C. 2071, 2073, 2077(d), 2138, 2152, 2201,
2232, 2233, 2234, 2236, 2237, 2243, 2273,
2282, 2021, 2297f); Energy Reorganization
Act of 1974, secs. 201, 202, 206, 211 (42
U.S.C. 5841, 5842, 5846, 5851); Nuclear
Waste Policy Act of 1982, secs. 135, 141 (42
U.S.C. 10155, 10161); 44 U.S.C. 3504 note.
§ 70.20a
[Amended]
130. In § 70.20a, in paragraph (b)
remove the phrase ‘‘parts 30 through 36,
39, 40, 50, 72, 110,’’ and add in its place
the phrase ‘‘parts 30 through 36, 39, 40,
50, 53, 72, 110’’.
■
§ 70.22
[Amended]
131. In § 70.22, wherever it appears,
remove the phrase ‘‘part 50’’ and add in
its place the phrase ‘‘parst 50 or 53’’.
■ 132. In § 70.24, revise paragraph (d) to
read as follows:
■
§ 70.24
Criticality accident requirements.
*
*
*
*
*
(d)(1) The requirements in paragraphs
(a) through (c) of this section do not
apply to a holder of a construction
permit or operating license for a nuclear
power reactor issued under part 50 or
part 53 of this chapter or a combined
license issued under part 52 or part 53
of this chapter, if the holder complies
with the requirements of paragraph (b)
of 10 CFR 50.68 or paragraph (m)(2) of
10 CFR 53.440,as applicable.
(2) An exemption from § 70.24 held
by a licensee who thereafter elects to
comply with requirements of paragraph
(b) of 10 CFR 50.68 or paragraph (m)(2)
of 10 CFR 53.440 does not exempt that
licensee from complying with any of the
requirements in § 50.68 or § 53.440(m)
of this chapter but shall be ineffective so
long as the licensee elects to comply
with § 50.68(b) or § 53.440(m)(2) of this
chapter, as applicable.
§ 70.32
[Amended]
133. In § 70.32, in paragraph (c)(1)
introductory text, remove the phrase
‘‘part 50 of this chapter’’ and add in its
place the phrase ‘‘parts 50 or 53 of this
chapter’’; and in paragraph (d) remove
the phrase ‘‘or § 70.34 of this chapter, as
appropriate.’’ and add in its place the
phrase ‘‘, §§ 74.34 or 53.1510 of this
chapter, as appropriate.’’.
■
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134. In § 70.50, revise paragraph (d) to
read as follows:
■
§ 70.50
Reporting requirements.
*
*
*
*
*
(d) The provisions of § 70.50 do not
apply to licensees subject to §§ 50.72 or
53.1630 of this chapter. They do apply
to those 10 CFR parts 50 or 53 licensees
possessing material licensed under 10
CFR part 70 that are not subject to the
notification requirements in §§ 50.72 or
53.1630 of this chapter.
PART 72—LICENSING
REQUIREMENTS FOR THE
INDEPENDENT STORAGE OF SPENT
NUCLEAR FUEL AND HIGH-LEVEL
RADIOACTIVE WASTE, AND
REACTOR-RELATED GREATER THAN
CLASS C WASTE
135. The authority citation for 10 CFR
part 72 continues to read as follows:
■
Authority: Atomic Energy Act of 1954,
secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182,
183, 184, 186, 187, 189, 223, 234, 274 (42
U.S.C. 2071, 2073, 2077, 2092, 2093, 2095,
2099, 2111, 2201, 2210e, 2232, 2233, 2234,
2236, 2237, 2238, 2273, 2282, 2021); Energy
Reorganization Act of 1974, secs. 201, 202,
206, 211 (42 U.S.C. 5841, 5842, 5846, 5851);
National Environmental Policy Act of 1969
(42 U.S.C. 4332); Nuclear Waste Policy Act
of 1982, secs. 117(a), 132, 133, 134, 135, 137,
141, 145(g), 148, 218(a) (42 U.S.C. 10137(a),
10152, 10153, 10154, 10155, 10157, 10161,
10165(g), 10168, 10198(a)); 44 U.S.C. 3504
note.
136. In § 72.3, revise the definition for
‘‘Independent spent fuel storage
installation or ISFSI’’ to read as follows:
■
§ 72.3
Definitions.
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Independent spent fuel storage
installation or ISFSI means a complex
designed and constructed for the
interim storage of spent nuclear fuel,
solid reactor-related GTCC waste, and
other radioactive materials associated
with spent fuel and reactor-related
GTCC waste storage. An ISFSI which is
located on the site of another facility
licensed under this part or a facility
licensed under part 50 or part 53 of this
chapter and which shares common
utilities and services with that facility or
is physically connected with that other
facility may still be considered
independent.
*
*
*
*
*
■ 137. In § 72.30, revise paragraph (e)(5)
to read as follows:
§ 72.30 Financial assurance and
recordkeeping for decommissioning.
*
*
*
*
*
(e) * * *
(5) In the case of licensees who are
issued a power reactor license under
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parts 50 or 53 of this chapter or ISFSI
licensees who are an electric utility, as
defined in parts 50 or 53 of this chapter,
with a specific license issued under this
part, the methods of §§ 50.75(b), (e), and
(h) or 53.1010, 53.1040, 53.1045(b), and
53.1060 of this chapter, as applicable. In
the event that funds remaining to be
placed into the licensee’s ISFSI
decommissioning external sinking fund
are no longer approved for recovery in
rates by a competent rate making
authority, the licensee must make
changes to provide financial assurance
using one or more of the methods stated
in paragraphs (a)(1) through (4) of this
section.
*
*
*
*
*
■ 138. In § 72.32, revise paragraph (c)(2)
to read as follows:
§ 72.32
Emergency plan.
*
*
*
*
*
(c) * * *
(2)(i) Located within the exclusion
area as defined in 10 CFR part 100, of
a nuclear power reactor licensed for
operation by the Commission, the
emergency plan that meets either the
requirements in § 50.160 of this chapter
or the requirements in appendix E to
part 50 of this chapter and § 50.47(b) of
this chapter shall be deemed to satisfy
the requirements of this section.
(ii) Located within the exclusion area,
as defined in 10 CFR part 53, of a
commercial nuclear plant licensed for
operation by the Commission, the
emergency plan that meets either the
requirements in § 50.160 of this chapter
or the requirements in appendix E to
part 50 of this chapter and § 50.47(b) of
this chapter shall be deemed to satisfy
the requirements of this section.
*
*
*
*
*
§ 72.40
[Amended]
139. In § 72.40, in paragraph (c)
remove the phrase ‘‘under part 50 of this
chapter,’’ and add in its place the phrase
‘‘under parts 50 or 53 of this chapter,’’.
■ 140. In § 72.75, revise paragraph
(i)(1)(ii) to read as follows:
■
§ 72.75 Reporting requirements for
specific events and conditions.
*
*
*
*
*
(i) * * *
(1) * * *
(ii) Licensees issued a general license
under § 72.210, after the licensee has
placed spent fuel on the ISFSI storage
pad (if the ISFSI is located inside the
collocated protected area, for a reactor
licensed under parts 50 or 53 of this
chapter) or after the licensee has
transferred spent fuel waste outside the
reactor licensee’s protected area to the
ISFSI storage pad (if the ISFSI is located
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87117
outside the collocated protected area,
for a reactor licensed under parts 50 or
53 of this chapter).
*
*
*
*
*
§ 72.184
[Amended]
141. In § 72.184, in paragraph (a)
remove the phrase ‘‘under part 50 of this
chapter’’ and add in its place the phrase
‘‘under parts 50 or 53 of this chapter’’.
■ 142. Revise § 72.210 to read as
follows:
■
§ 72.210
General license issued.
A general license is hereby issued for
the storage of spent fuel in an
independent spent fuel storage
installation at power reactor sites to
persons authorized to possess or operate
nuclear power reactors under 10 CFR
parts 50, 52, or 53.
■ 143. In § 72.212, revise paragraph
(b)(8) to read as follows:
§ 72.212 Conditions of general license
issued under § 72.210.
*
*
*
*
*
(b) * * *
(8) Before use of the general license,
determine whether activities related to
storage of spent fuel under this general
license involve a change in the facility
Technical Specifications or require a
license amendment for the facility
pursuant to §§ 50.59(c) or 53.1550 of
this chapter. Results of this
determination must be documented in
the evaluations made in paragraph (b)(5)
of this section.
*
*
*
*
*
■ 144. In § 72.218, revise paragraphs (a)
and (b) to read as follows:
§ 72.218
Termination of licenses.
(a) The notification regarding the
program for the management of spent
fuel at the reactor required by
§§ 50.54(bb) or 53.1060 of this chapter
must include a plan for removal of the
spent fuel stored under this general
license from the reactor site. The plan
must show how the spent fuel will be
managed before starting to
decommission systems and components
needed for moving, unloading, and
shipping this spent fuel.
(b) An application for termination of
a reactor operating license issued under
10 CFR part 50 and submitted under
§ 50.82 of this chapter, or a combined
license issued under 10 CFR part 52 and
submitted under § 52.110 of this
chapter, or a reactor operating or
combined license under 10 CFR part 53
and submitted under § 53.1070 of this
chapter must contain a description of
how the spent fuel stored under this
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general license will be removed from
the reactor site.
*
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*
*
*
PART 73—PHYSICAL PROTECTION OF
PLANTS AND MATERIALS
145. The authority citation for 10 CFR
part 73 continues to read as follows:
■
Authority: Atomic Energy Act of 1954,
secs. 53, 147, 149, 161, 161A, 170D, 170E,
170H, 170I, 223, 229, 234, 1701 (42 U.S.C.
2073, 2167, 2169, 2201, 2201a, 2210d, 2210e,
2210h, 2210i, 2273, 2278a, 2282, 2297f);
Energy Reorganization Act of 1974, secs. 201,
202 (42 U.S.C. 5841, 5842); Nuclear Waste
Policy Act of 1982, secs. 135, 141 (42 U.S.C.
10155, 10161); 44 U.S.C. 3504 note.
Section 73.37(b)(2) also issued under sec.
301, Pub. L. 96–295, 94 Stat. 789 (42 U.S.C.
5841 note).
146. In § 73.1, revise paragraph
(b)(1)(i) to read as follows:
■
§ 73.1
[Amended]
*
*
*
*
*
(b) * * *
(1) * * *
(i) The physical protection of
production and utilization facilities
licensed under parts 50, 52, or 53 of this
chapter,
*
*
*
*
*
■ 147. In § 73.2, revise the introductory
text and paragraph (a) to read as follows:
§ 73.2
Definitions.
As used in this part:
(a) Terms defined in parts 50, 52, 53,
70, and 95 of this chapter have the same
meaning when used in this part.
*
*
*
*
*
■ 148. In § 73.8, revise paragraph (b) to
read as follows:
§ 73.8 Information collection
requirements: OMB approval.
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*
*
*
(b) The approved information
collection requirements contained in
this part appear in §§ 73.5, 73.15, 73.17,
73.20, 73.21, 73.24, 73.25, 73.26, 73.27,
73.37, 73.40, 73.45, 73.46, 73.50, 73.54,
73.55, 73.56, 73.57, 73.58, 73.60, 73.67,
73.70, 73.72, 73.73, 73.74, 73.77, 73.100,
73.110, 73.120, 73.1200, 73.1205,
73.1210, 73.1215, and appendices B and
C to this part.
*
*
*
*
*
■ 149. In § 73.50, revise the introductory
text to read as follows:
§ 73.50 Requirements for physical
protection of licensed activities.
Each licensee who is not subject to
§ 73.51, but who possesses, uses, or
stores formula quantities of strategic
special nuclear material that are not
readily separable from other radioactive
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material and which have a total external
radiation level in excess of 1 gray (100
rad) per hour at a distance of 1 meter
(3.3 feet) from any accessible surfaces
without intervening shielding other
than at a nuclear reactor facility
licensed under parts 50, 52, or 53 of this
chapter, shall comply with the
following:
*
*
*
*
*
■ 150. In § 73.55, revise paragraphs
(a)(4) and (6), (i)(4)(iii), (l)(1), (l)(7)(ii),
(p)(1)(i), (r)(2) and (r)(4)(iii) to read as
follows:
§ 73.55 Requirements for physical
protection of licensed activities in nuclear
power reactors against radiological
sabotage.
(a) * * *
(4) Applicants for an operating license
under the provisions of part 50 or part
53 of this chapter or holders of a
combined license under the provisions
of part 52 or part 53 of this chapter shall
implement the requirements of this
section before fuel is allowed onsite
(protected area).
*
*
*
*
*
(6) Applicants for an operating license
under the provisions of part 50 or part
53 of this chapter, or holders of a
combined license under the provisions
of part 52 or part 53 of this chapter that
do not reference a standard design
certification or reference a standard
design certification issued after May 26,
2009, shall meet the requirement of
§ 73.55(i)(4)(iii).
*
*
*
*
*
(i) * * *
(4) * * *
(iii) Applicants for an operating
license under the provisions of part 50
of this chapter, or holders of a combined
license under the provisions of part 52
of this chapter, or licensees under part
53 of this chapter that elect to
demonstrate compliance with § 73.55,
consistent with § 53.860(a)(2) of this
chapter, shall construct, locate, protect,
and equip both the central and
secondary alarm stations to the
standards for the central alarm station
contained in this section. Both alarm
stations shall be equal and redundant,
such that all functions needed to satisfy
the requirements of this section can be
performed in both alarm stations.
*
*
*
*
*
(l) * * *
(1) Commercial nuclear power
reactors licensed under 10 CFR parts 50,
52, or 53 and authorized to use special
nuclear material in the form of MOX
fuel assemblies containing up to 20
weight percent PuO2 shall, in addition
to demonstrating compliance with the
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requirements of this section, protect unirradiated MOX fuel assemblies against
theft or diversion as described in this
paragraph.
*
*
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*
(7) * * *
(ii) Additional measures for the
physical protection of un-irradiated
MOX fuel assemblies containing greater
than 20 weight percent PuO2 shall be
determined by the Commission on a
case-by-case basis and documented
through license amendment in
accordance with §§ 50.90 or 53.1510 of
this chapter.
*
*
*
*
*
(p) * * *
(1) * * *
(i) Under §§ 50.54(x) and (y) or
53.740(h) of this chapter, the licensee
may suspend any security measures
under this section in an emergency
when this action is immediately needed
to protect the public health and safety
and no action consistent with license
conditions and technical specifications
that can provide adequate or equivalent
protection is immediately apparent.
This suspension of security measures
must be approved as a minimum by a
licensed senior operator before taking
this action.
*
*
*
*
*
(r) * * *
(2) The licensee shall submit
proposed alternative measure(s) to the
Commission for review and approval
under §§ 50.4 and 50.90, or §§ 53.040
and 53.1510 of this chapter before
implementation.
*
*
*
*
*
(4) * * *
(iii) Based on comparison of the costs
of the alternative measures to the costs
of demonstrating compliance with the
Commission’s requirements using the
essential elements of §§ 50.109 or
53.1590 of this chapter, the costs of fully
demonstrating compliance with the
Commission’s requirements are not
justified by the protection that would be
provided.
■ 151. In § 73.56, revise paragraph (a)(3)
to read as follows:
§ 73.56 Personnel access authorization
requirements for nuclear power plants.
(a) * * *
(3) Each applicant for an operating
license under the provisions of part 50
of this chapter, each holder of a
combined license under the provisions
of part 52 of this chapter, and applicants
for an operating license or holders of a
combined license under part 53 of this
chapter that do not meet the
requirements of § 53.860(a)(2) of this
chapter, shall implement the
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requirements of this section before fuel
is allowed on site (protected area).
*
*
*
*
*
■ 152. In § 73.57, revise paragraph (a)(3)
to read as follows:
§ 73.57 Requirements for criminal history
records checks of individuals granted
unescorted access to a nuclear power
facility, a non-power reactor, or access to
Safeguards Information.
(a) * * *
(3) Before receiving its operating
license under 10 CFR parts 50 or 53 or
before the Commission makes its
finding under §§ 52.103(g) or 53.1452(g)
of this chapter, each applicant for a
license to operate a nuclear power
reactor (including an applicant for a
combined license) or a non-power
reactor may submit fingerprints for
those individuals who will require
unescorted access to the nuclear power
facility or non-power reactor facility.
*
*
*
*
*
■ 153. In § 73.58, revise paragraph (a) to
read as follows:
§ 73.58 Safety/security interface
requirements for nuclear power reactors.
(a) Each operating nuclear power
reactor licensee with a license issued
under parts 50, 52, or 53 of this chapter
shall comply with the requirements of
this section.
*
*
*
*
*
■ 154. In § 73.67, revise paragraphs (d)
introductory text and (f) introductory
text to read as follows:
§ 73.67 Licensee fixed site and in-transit
requirements for the physical protection of
special nuclear material of moderate and
low strategic significance.
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(d) Fixed site requirements for special
nuclear material of moderate strategic
significance. Each licensee who
possesses, stores, or uses quantities and
types of special nuclear material of
moderate strategic significance at a fixed
site or contiguous sites, except as
allowed by paragraph (b)(2) of this
section and except those who are
licensed to operate a nuclear power
reactor pursuant to part 50 or part 53,
provided that the special nuclear
material is located within a protected
area and protected under § 73.55 or
§ 73.100, shall:
*
*
*
*
*
(f) Fixed site requirements for special
nuclear material of low strategic
significance. Each licensee who
possesses, stores, or uses special nuclear
material of low strategic significance at
a fixed site or contiguous sites, except
those who are licensed to operate a
nuclear power reactor pursuant to part
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50 or part 53, provided that the special
nuclear material is located within a
protected area and protected under
§ 73.55 or § 73.100, shall:
*
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*
*
*
■ 155. In § 73.77, revise paragraphs (a),
(b)(1), (c)(6) and (7) to read as follows:
§ 73.77
Cybersecurity event notifications.
(a) Each licensee subject to the
provisions of §§ 73.54 or 73.110 shall
notify the NRC Headquarters Operations
Center via the Emergency Notification
System (ENS), under paragraph (c) of
this section:
(1) Within one hour after discovery of
a cyberattack that adversely impacted:
(i) Safety-related or important-tosafety functions, security functions, or
emergency preparedness functions
(including offsite communications); or
that compromised support systems and
equipment resulting in adverse impacts
to safety, security, or emergency
preparedness functions within the scope
of § 73.54; or,
(ii) Functions performed by digital
assets that would prevent a postulated
fission product release resulting in
offsite doses exceeding the values in
§ 53.210 of this chapter, or functions
performed by digital assets used by the
licensee for implementing the physical
security requirements in § 53.860(a) of
this chapter.
(2) Within 4 hours:
(i) After discovery of a cyberattack
that could have caused an adverse
impact to:
(A) Safety-related or important-tosafety functions, security functions, or
emergency preparedness functions
(including offsite communications); or
that could have compromised support
systems and equipment, which if
compromised, could have adversely
impacted safety, security, or emergency
preparedness functions within the scope
of § 73.54; or,
(B) Functions performed by digital
assets that would prevent a postulated
fission product release resulting in
offsite doses exceeding the values in
§ 53.210 of this chapter, or functions
performed by digital assets used by the
licensee for implementing the physical
security requirements in § 53.860(a) of
this chapter.
(ii) After discovery of a suspected or
actual cyberattack initiated by personnel
with physical or electronic access to
digital computer and communication
systems and networks within the scope
of §§ 73.54 or 73.110.
(iii) After notification of a local, State,
or other Federal agency (e.g., law
enforcement, Federal Bureau of
Investigation (FBI), etc.) of an event
related to the licensee’s implementation
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87119
of their cybersecurity program for digital
computer and communication systems
and networks within the scope of
§§ 73.54 or 73.110 that does not
otherwise require a notification under
paragraph (a) of this section.
(3) Within 8 hours after receipt or
collection of information regarding
observed behavior, activities, or
statements that may indicate
intelligence gathering or pre-operational
planning related to a cyberattack against
digital computer and communication
systems and networks within the scope
of §§ 73.54 or 73.110.
(b) Twenty-four hour recordable
events. (1) The licensee shall use the site
corrective action program to record
vulnerabilities, weaknesses, failures and
deficiencies in their § 73.54 or § 73.110
cybersecurity program within 24 hours
of their discovery.
*
*
*
*
*
(c) * * *
(6) Declaration of emergencies.
Notifications made to the NRC for the
declaration of an emergency class shall
be performed in accordance with
§§ 50.72 or 53.1630 of this chapter, as
applicable.
(7) Elimination of duplication.
Separate notifications and reports are
not required for events that are also
reportable under §§ 50.72 and 50.73 or
§§ 53.1630 and 53.1640 of this chapter.
However, these notifications should also
indicate the applicable § 73.77 reporting
criteria.
*
*
*
*
*
■ 156. Add Subpart J consisting of
§§ 73.100 through 73.120 to read as
follows:
Subpart J—Security Requirements at
Commercial Nuclear Plants
Sec.
73.100 Technology-inclusive requirements
for physical protection of licensed
activities at commercial nuclear plants
against radiological sabotage.
73.110 Technology-inclusive requirements
for protection of digital computer and
communication systems and networks.
73.120 Access authorization program for
commercial nuclear plants.
Subpart J—Security Requirements at
Commercial Nuclear Plants
§ 73.100 Technology-inclusive
requirements for physical protection of
licensed activities at commercial nuclear
plants against radiological sabotage.
(a) Introduction. (1) Each licensee that
is licensed to operate a commercial
nuclear plant under 10 CFR part 53 and
elects to implement the requirements of
this section must do so through its
physical security plan, training and
qualification plan, safeguards
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contingency plan, and cybersecurity
plan, referred to collectively hereafter as
‘‘security plans,’’ before initial fuel load
into the reactor (or, for a fueled
manufactured reactor, before initiating
the physical removal of any one of the
independent physical mechanisms to
prevent criticality required under
§ 53.620(d)(1) of this chapter).
(2) The security plans must identify,
describe, and account for site-specific
conditions that affect the licensee’s
capability to satisfy the requirements of
this section.
(b) General performance objective and
requirements. (1) The licensee must
establish, implement, and maintain a
physical protection program and a
security organization, which will have
as their objective to provide reasonable
assurance that activities involving
special nuclear material are not inimical
to the common defense and security and
do not constitute an unreasonable risk
to the public health and safety.
(2) To satisfy the general performance
objective of paragraph (b)(1) of this
section, the physical protection program
must protect against the design basis
threat of radiological sabotage as stated
in § 73.1. Specifically, the licensee
must—
(i) Ensure that the physical protection
program capabilities to protect against
the design basis threat of radiological
sabotage are maintained at all times; and
(ii) Provide defense in depth in
achieving performance requirements
through the integration of engineered
systems, administrative controls, and
management measures.
(3) The physical protection program
must be designed and implemented to
achieve and maintain the reliability and
availability of structures, systems, and
components (SSCs) required for
demonstrating compliance with the
following performance requirements at
all times:
(i) Intrusion detection. The licensee
must be capable of detecting attempted
and actual unauthorized access to
interior and exterior areas containing
SSCs needed to implement safety and
security functions.
(ii) Intrusion assessment. The licensee
must be capable of timely assessment
for determining the cause of a detected
intrusion.
(iii) Security communication. The
licensee must be capable of continuous
security communications.
Communication systems must account
for design basis threats that can
interrupt or interfere with continuity or
integrity of communications.
(iv) Security response. The physical
protection program must be designed to
provide timely security response to
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interdict and neutralize adversary
attacks up to and including the design
basis threat of radiological sabotage. The
physical protection program must be
designed to provide layers of security
response, with each layer assuring that
a single failure does not result in the
loss of capability to neutralize the
design basis threat adversary.
Structures, systems, and components
relied on for delay functions must be
designed to allow for timely security
responses to adversary attacks with
adequate defense in depth.
(A) The security response may rely on
the use of onsite responders, law
enforcement or other offsite armed
responders, or a combination thereof, to
fulfill the interdiction and
neutralization functions required by
paragraph (b)(3)(iv) of this section. A
licensee relying entirely or partially on
law enforcement or other offsite armed
responders must—
(1) Maintain the capability to detect,
assess, interdict, and neutralize threats
as required by paragraphs (b)(3)(i),
(b)(3)(ii), and (b)(3)(iv) of this section;
(2) Provide adequate delay to enable
law enforcement or other offsite armed
responders to fulfill the interdiction and
neutralization functions for threats up to
and including the design basis threat of
radiological sabotage;
(3) Provide necessary information
about the facility and make available
periodic training to law enforcement or
other offsite armed responders who will
fulfill the interdiction and
neutralization functions for threats up to
and including the design basis threat of
radiological sabotage;
(4) Fully describe in the safeguards
contingency plan the role that law
enforcement or other offsite armed
responders will play in the licensee’s
protective strategy. The description
must provide sufficient detail to enable
the NRC to determine that the licensee’s
physical protection program provides
reasonable assurance of adequate
protection against threats up to and
including the design basis threat of
radiological sabotage; and
(5) Identify criteria and measures to
compensate for the degradation or
absence of law enforcement or other
offsite armed responders and propose
suitable compensatory measures that
meet the requirements of paragraph
(h)(3) of this section to address this
degradation.
(B) For licensees relying entirely or
partially on law enforcement responders
to fulfill the interdiction and
neutralization functions required by
paragraph (b)(3)(iv) of this section, the
training and qualification requirements
related to armed response personnel in
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paragraphs (c) and (e) of this section do
not apply to law enforcement
responders. The licensee shall continue
to satisfy the performance evaluation
requirements in paragraph (g) of this
section for all armed response
personnel, including law enforcement.
(v) Protecting against land and
waterborne vehicle bomb assaults. The
licensee must be capable of protecting
the plant against the design basis threat
vehicle bomb assault. The methods that
are relied on to protect against a design
basis threat land vehicle and waterborne
vehicle bomb assault must be designed
to protect the reactor building and
structures containing safety- or securityrelated systems, and components from
explosive effects.
(vi) Access control portals. The
licensee must be capable of detecting
and denying unauthorized access to
persons and pass-through of contraband
materials (e.g., weapons, incendiaries,
explosives) to protected areas.
(4) The licensee must meet the
requirements related to target sets in
§ 73.55(f).
(5) The licensee must identify and
analyze site-specific conditions,
including target sets, that may affect the
physical protection program needed to
implement the requirements of this
section. The licensee must account for
these conditions in demonstrating
compliance with the requirements of
this section.
(6) The licensee must establish,
implement, and maintain a performance
evaluation program to assess the
effectiveness of the licensee’s
implementation of the physical
protection program to protect against
the design basis threat of radiological
sabotage.
(7) The licensee must establish,
implement, and maintain an access
authorization program under § 73.56
and must describe the program in the
physical security plan.
(8) The licensee must establish,
implement, and maintain a
cybersecurity program under §§ 73.54 or
73.110 and must describe the program
in the cybersecurity plan.
(9) The licensee must establish,
implement, and maintain an insider
mitigation program and must describe
the program in the physical security
plan.
(i) The insider mitigation program
must monitor the initial and continuing
trustworthiness and reliability of
individuals granted or retaining
unescorted access or unescorted access
authorization to a protected or vital
area, and implement defense-in-depth
methodologies to minimize the potential
for an insider (active, passive, or both)
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to adversely affect, either directly or
indirectly, the licensee’s capability to
protect against radiological sabotage.
(ii) The insider mitigation program
must integrate elements of—
(A) The access authorization program
under § 73.56;
(B) The fitness-for-duty program
under 10 CFR part 26;
(C) The cybersecurity program under
§§ 73.54 or 73.110; and
(D) The physical protection program
under this section.
(10) The licensee must have the
capability to track, trend, correct, and
prevent recurrence of failures and
deficiencies in the implementation of
the requirements of this section.
(11) Implementation of security plans
and associated procedures must be
coordinated with other onsite plans and
procedures to preclude conflict during
both normal and emergency conditions
and ensure the adequate management of
the safety and security interface.
(12)(i) The licensee must ensure that
the firearms background check
requirements of § 73.17 of this part are
met for all members of the security
organization whose official duties
require access to covered weapons or
who inventory enhanced weapons.
(ii) The provisions of this paragraph
are only applicable to licensees subject
to this section that are also subject to the
firearms background check provisions of
§ 73.17 of this part.
(c) Security organization. The licensee
must establish and maintain a security
organization that is staffed, trained,
qualified, and equipped to implement
the physical protection program under
the requirements of this section.
(1) The licensee must establish a
management system for maintaining and
implementing security policies and
procedures to implement the
requirements of this section and the
security plans.
(2) Implementing procedures must
document the conduct of security
operations, security design and
configuration controls, maintenance,
training and qualification, and
contingency responses.
(3) The licensee must—
(i) Establish a process for the approval
of designs, policies, processes, and
procedures and changes by the
individual with overall responsibility
for the physical protection program; and
(ii) Ensure that revisions and changes
to the physical protection program and
implementing policies, processes, and
procedures satisfy the requirements of
this section.
(4) The licensee must retain, in
accordance with § 73.70, all analyses,
assessments, calculations, and
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descriptions of the technical basis for
demonstrating compliance with the
performance requirements of
§ 73.100(b). The licensee must protect
these records in accordance with the
requirements for protecting safeguards
information in §§ 73.21 and 73.22.
(5) The licensee may not permit any
individual to implement any part of the
physical protection program unless the
individual has been trained, equipped,
and qualified to perform their assigned
duties and responsibilities in
accordance with the training and
qualification plan.
(d) Search requirements. The licensee
must establish and implement searches
of individuals, vehicles, and materials
to detect and prevent the introduction
into the protected area of firearms,
explosives, incendiary devices, or other
items and material which could be used
to commit radiological sabotage.
(e) Training and qualification
program. The licensee must establish
and maintain a training and
qualification program that ensures
personnel who are responsible for the
physical protection of the facility
against radiological sabotage are able to
effectively perform their assigned
security-related job duties for
implementing the requirements of this
section and must describe the program
in the training and qualification plan.
(f) Security reviews. The licensee must
establish and implement security
reviews to assess the effectiveness of the
implementation of the physical
protection program. Security reviews
must be performed by individuals
independent of those personnel
responsible for program management
and any individual who has direct
responsibility for implementing the
onsite physical protection program.
(1) The licensee must review each
element of the physical protection
program at a frequency commensurate
with the importance or significance to
safety of plant operations to ensure
timely identification and documentation
of vulnerabilities, improvements, and
corrective actions. The objective of these
reviews must be maintaining effective
implementation of the engineered and
administrative controls required to
achieve the physical protection program
functions and the management system
required to implement programs and
requirements in this section.
(2) The licensee must establish and
perform self-assessments to ensure the
effective implementation of the physical
protection program functions of
detection, assessment, communication,
delay, and interdiction and
neutralization to protect against the
design basis threat of radiological
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sabotage. The licensee must perform
design verification and assessments of
the capabilities of active and passive
engineering systems relied on to protect
against the design basis threat.
(3) Reviews of the security program
must include, but are not limited to, an
audit of the effectiveness of the physical
protection program, security plans,
implementing procedures, cybersecurity
programs, safety/security interface
activities, the testing, maintenance, and
calibration program, and response
commitments by local, State, and
Federal law enforcement authorities.
(4) The results and recommendations
of the onsite physical protection
program reviews, management’s
findings regarding program
effectiveness, and any actions taken as
a result of recommendations from prior
program reviews, must be documented
in a report and must be maintained in
an auditable form and available for
inspection.
(g) Performance evaluation. Licensee
performance evaluations must include
methods appropriate and necessary to
assess, test, and challenge the
integration of the physical protection
program’s functions to protect against
the design basis threat, including
measures to protect against cyberattack
and engineered systems designed to
protect against the design basis threat
standalone ground vehicle bomb attack.
(1) The licensee must establish the
frequencies for performance evaluations
commensurate with the security
significance of the physical protection
program.
(2) The licensee must document
processes and procedures for
implementing the performance
evaluations. The licensee must maintain
records, including results, findings, and
corrective actions identified during the
performance evaluations.
(h) Maintenance, testing, and
calibration and corrective actions. (1)
The licensee must ensure that security
SSCs, including supporting systems, are
inspected, tested, and calibrated for
operability and performance at intervals
necessary and sufficient to meet the
requirements of this section.
(2) The licensee must implement
corrective actions to ensure resolution
of identified vulnerabilities and
deficiencies to meet the requirements of
this section.
(3) The licensee must establish and
implement timely compensatory
measures for degraded or inoperable
security SSCs to meet the requirements
of this section. Compensatory measures
must provide a level of protection that
is equivalent to the protection that was
provided prior to the degradation or
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inoperability of the security structures,
systems, or components.
(4) The licensee must document
processes and procedures and maintain
records for implementing the corrective
actions, compensatory measures, and
maintenance, inspection, testing, and
calibration of security SSCs.
(i) Suspension of security measures.
(1) The licensee may suspend
implementation of affected
requirements of this section in
accordance with § 53.740(h) of this
chapter under the following conditions:
(i) In an emergency, when action is
immediately needed to protect the
public health and safety; and
(ii) During severe weather, when the
suspension of affected security
measures is immediately needed to
protect the personal health and safety of
personnel.
(2) Suspended security measures must
be reinstated as soon as conditions
permit.
(3) The suspension of security
measures must be reported and
documented in accordance with the
provisions of §§ 73.1200 and 73.1205.
(j) Records. (1) The Commission may
inspect, copy, retain, and remove all
reports, records, and documents
required to be kept by Commission
regulations, orders, or license
conditions, whether the reports, records,
and documents are kept by the licensee
or a contractor.
(2) The licensee must maintain all
records required to be kept by
Commission regulations, orders, or
license conditions, until the
Commission terminates the license for
which the records were developed and
must maintain superseded portions of
these records for at least 3 years after the
record is superseded, unless otherwise
specified by the Commission.
(3) If a contracted security force is
used to implement the onsite physical
protection program, the licensee’s
written agreement with the contractor
must be retained by the licensee as a
record for the duration of the contract.
(4) Review and audit reports must be
available for inspection, for a period of
3 years.
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§ 73.110 Technology-inclusive
requirements for protection of digital
computer and communication systems and
networks.
(a) Each licensee that is licensed to
operate a commercial nuclear plant
under 10 CFR part 53 and elects to
implement the requirements of this
section must establish, implement, and
maintain a cybersecurity program that is
commensurate with the potential
consequences resulting from
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cyberattacks, up to and including the
design basis threat as described in
§ 73.1. The cybersecurity program must
provide reasonable assurance that
digital computer and communication
systems and networks are adequately
protected against cyberattacks that are
capable of causing the following
consequences:
(1) Adversely impacting the functions
performed by digital assets that would
prevent a postulated fission product
release resulting in offsite doses
exceeding the values in § 53.210 of this
chapter.
(2) Adversely impacting the functions
performed by digital assets used by the
licensee for implementing the physical
security requirements in § 53.860(a) of
this chapter.
(b) To protect digital computer and
communication systems and networks
associated with the functions described
in paragraphs (a)(1) and (2), the licensee
must—
(1) Analyze the potential
consequences resulting from
cyberattacks on digital computer and
communication systems and networks
and identify those assets that must be
protected to demonstrate compliance
with paragraph (a) of this section; and
(2) Implement the cybersecurity
program in accordance with paragraph
(d) of this section.
(c) The licensee must comply with the
requirements in § 73.54(a)(2) for the
systems and networks identified in
paragraph (b)(1) of this section in a
manner that is commensurate with the
potential consequences resulting from
cyberattacks.
(d) The cybersecurity program must
be designed in a manner that is
commensurate with the potential
consequences resulting from
cyberattacks through the following
steps:
(1) Implement security controls to
protect the assets identified under
paragraph (b)(1) of this section from
cyberattacks, commensurate with their
safety and security significance;
(2) Apply and maintain defense-indepth protective strategies to ensure the
capability to detect, delay, respond to,
and recover from cyberattacks capable
of causing the consequences identified
in paragraph (a) of this section;
(3) Mitigate the adverse effects of
cyberattacks capable of causing the
consequences identified in paragraph (a)
of this section; and
(4) Ensure that the functions of
protected assets identified under
paragraph (b)(1) of this section are not
adversely impacted due to cyberattacks.
(e) The licensee must implement the
following requirements in a manner that
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is commensurate with the potential
consequences resulting from
cyberattacks:
(1) As part of the cybersecurity
program, the licensee must comply with
the requirements in § 73.54(d)(1), (2),
and (4), and must ensure that
modifications to assets, identified under
paragraph (b)(1) of this section are
evaluated before implementation to
ensure that the cybersecurity
performance objectives identified in
paragraph (a) of this section are
maintained.
(2) The licensee must establish,
implement, and maintain a
cybersecurity plan that implements the
cybersecurity program requirements of
this section.
(i) The cybersecurity plan must
describe how the requirements of this
section will be implemented and must
account for the site-specific conditions
that affect implementation.
(ii) The cybersecurity plan must
include measures for incident response
and recovery for cyberattacks. The
cybersecurity plan must include the
analysis identified under paragraph
(b)(1) of this section and describe how
the licensee will—
(A) Apply and maintain defense-indepth protective strategies as required
in paragraph (d)(2) of this section;
(B) Maintain the capability for timely
detection and response to cyberattacks;
(C) Mitigate the consequences of
cyberattacks;
(D) Correct exploited vulnerabilities;
and
(E) Restore affected systems,
networks, and/or equipment affected by
cyberattacks.
(3) The licensee must develop and
maintain written policies and
implementing procedures to implement
the cybersecurity plan. Policies,
implementing procedures, and other
supporting technical information used
by the licensee need not be submitted
for Commission review and approval as
part of the cybersecurity plan but are
subject to inspection by NRC staff on a
periodic basis.
(4) The licensee must establish and
implement cybersecurity reviews to
assess the effectiveness of the
implementation of the cybersecurity
program.
(i) The licensee must review each
element of the cybersecurity program at
a frequency commensurate with the
importance or significance to safety of
plant operations to ensure timely
identification and documentation of
vulnerabilities, improvements, and
corrective actions.
(ii) Cybersecurity reviews must be
performed by individuals independent
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of those personnel responsible for
program management and any
individual who has direct responsibility
for implementing the cybersecurity
program.
(iii) The licensee must establish and
perform self-assessments to ensure the
effective implementation of the
cybersecurity program.
(iv) The results and recommendations
of the cybersecurity program reviews,
management’s findings regarding
program effectiveness, and any actions
taken as a result of recommendations
from prior program reviews, must be
documented in a report and must be
maintained in an auditable form and
available for inspection.
(5) The licensee must retain all
records and supporting technical
documentation required to demonstrate
compliance with the requirements of
this section as a record until the
Commission terminates the license for
which the records were developed and
must maintain superseded portions of
these records for at least three (3) years
after the record is superseded, unless
otherwise specified by the Commission.
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§ 73.120 Access authorization program for
commercial nuclear plants.
(a) Introduction and scope. Each
applicant for an operating license or a
holder of a combined license under 10
CFR part 53 must establish, maintain,
and implement an access authorization
program before initial fuel load into the
reactor (or, for a fueled manufactured
reactor, before initiating the physical
removal of any one of the independent
physical mechanisms to prevent
criticality required under § 53.620(d)(1)
of this chapter). The requirements in
this section apply to licensees satisfying
the criterion in § 53.860(a)(2)(i) of this
chapter.
(b) Applicability. (1) The following
individuals must be subject to an access
authorization program under this
section:
(i) Any individual to whom a licensee
intends to grant unescorted access to a
commercial nuclear plant protected
area, vital area, or controlled access area
where licensed material is used or
stored;
(ii) Any individual whose duties and
responsibilities permit the individual to
take actions by electronic means, either
on site or remotely, that could adversely
impact the licensee’s or applicant’s
operational safety, security, or
emergency preparedness;
(iii) Any individual who has
responsibilities for implementing a
licensee’s or applicant’s protective
strategy, including armed security force
officers, alarm station operators, and
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tactical response team leaders but not
including Federal, State, or local law
enforcement personnel; and
(iv) The licensee or applicant access
authorization program reviewing official
or contractor or vendor access
authorization program reviewers.
(2) The licensee or applicant may
subject other individuals, including
employees of a contractor or a vendor
who are designated in access
authorization program procedures, to an
access authorization program that
demonstrates compliance with the
requirements of this section.
(c) General performance objectives
and requirements. Each licensee’s or
applicant’s access authorization
program under this section must
demonstrate that the individuals who
are specified in paragraph (b) of this
section are trustworthy and reliable,
such that they do not constitute an
unreasonable risk to public health and
safety or the common defense and
security. The licensee’s access
authorization program must maintain
the capabilities for demonstrating
compliance with the following
performance requirements:
(1) Background investigation. (i)(A)
Licensees and applicants must ensure
that any individual seeking initial
unescorted access or to maintain
unescorted access is subject to a
background investigation.
(B) Background investigations must
include the program elements contained
under § 37.25 of this chapter and must
also include a credit history evaluation.
(ii) Background investigations must
include fingerprinting and an FBI
identification and criminal history
records check in accordance with
§ 37.27 of this chapter.
(iii) Licensees must have the informed
and signed consent of the subject
individual to initiate a background
investigation. This consent must
include authorization to share personal
information with other individuals or
organizations as necessary to complete
the background investigation. A signed
consent must be obtained prior to any
reinvestigation. The subject individual
may withdraw his or her consent at any
time. Licensees must inform the
individual that—
(A) If an individual withdraws his or
her consent, the licensee may not
initiate any elements of the background
investigation that were not in progress
at the time the individual withdrew his
or her consent; and
(B) The withdrawal of consent for the
background investigation is sufficient
cause for denial or termination of
unescorted access authorization.
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(2) Behavioral observation. Licensees,
applicants, contractors, and vendors
must ensure the access authorization
program includes provisions that the
individuals specified in paragraph (b) of
this section are subject to behavioral
observation.
(i) Each person subject to behavioral
observation must communicate to the
licensee or applicant observed behaviors
or activities of individuals that may
constitute an unreasonable risk to the
health and safety of the public and
common defense and security.
(ii) Behavioral observation must
include visual observation, in person or
remotely by video, to detect and
promptly report to plant supervision
any concerns arising from behavioral
observation, including, but not limited
to, concerns related to any questionable
behavior patterns or activities of others.
(3) Self-reporting of legal actions.
Licensees or applicants must inform
personnel who are granted and who
maintain unescorted access of their
responsibilities to self-report to plant
supervision legal actions taken by a law
enforcement authority or court of law
against the individual that could result
in incarceration or a court order or that
requires a court appearance, including
but not limited to an arrest, an
indictment, the filing of charges, or a
conviction, but excluding minor civil
actions or misdemeanors such as
parking violations or speeding tickets,
for any individual who has applied for
unescorted access or who maintains
unescorted access.
(4) Unescorted access. Licensees or
applicants must grant unescorted access
only after the licensee has verified an
individual is trustworthy and reliable. A
list of persons currently approved for
unescorted access to a protected area,
vital area, or controlled access area must
be maintained at all times. Unescorted
access determinations must be reviewed
annually by the reviewing official.
Licensees and applicants must complete
an FBI criminal history record check
update for each individual maintaining
unescorted access, within 10 years of
the last review.
(5) Termination of unescorted access.
Licensees and applicants must promptly
terminate unescorted access when this
access is no longer required or a
reviewing official determines an
individual is no longer trustworthy and
reliable in accordance with this section.
(6) Determination basis for access. (i)
The licensee’s or applicant’s reviewing
official must determine whether to
permit, deny, unfavorably terminate,
maintain, or administratively withdraw
an individual’s unescorted access based
on an evaluation of all of the
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information collected to demonstrate
compliance with the requirements of
this section.
(ii) Licensees and applicants must
provide individuals subject to this
section, prior to any final adverse
determination, the right to complete,
correct, and explain information
obtained as a result of the licensee’s
background investigation pursuant to
§ 37.23(g) of this chapter.
(iii) The licensee’s or applicant’s
reviewing officials are the only
individuals authorized to make
unescorted access determination
decisions. Each licensee or applicant
must name one or more individuals to
be reviewing officials pursuant to the
requirements of § 37.23(b)(2) of this
chapter.
(7) Review procedures. Review
procedures must be established in
accordance with § 37.23(f) of this
chapter, to include provisions for the
notification in writing of individuals
who are denied unescorted access or
who are unfavorably terminated.
(8) Protection of information.
Licensees, applicants, contractors, or
vendors must establish and maintain a
system of files and procedures in
accordance with § 37.31 of this chapter,
to ensure personal information is not
disclosed to unauthorized persons.
(9) Access authorization reviews and
corrective action. Licensees and
applicants must develop, implement,
and maintain procedures for conduct of
access authorization reviews and
corrective actions in accordance with
§ 37.33 of this chapter to ensure the
continuing effectiveness of the access
authorization program and to ensure
that the access authorization program
and program elements are in
compliance with the requirements of
this section. Each licensee and applicant
must be responsible for the continuing
effectiveness of the access authorization
program, including access authorization
program elements that are provided by
the contractors or vendors, and the
access authorization programs of any of
the contractors or vendors that are
accepted by the licensee or applicant.
(10) Records. Licensees, applicants,
and contractors or vendors must
document the processes and procedures
for maintaining records used or created
to establish an individual’s
trustworthiness and reliability or to
document access determinations.
Licensees, applicants, and contractor or
vendors must—
(i) Retain documentation regarding
the trustworthiness and reliability of
individual employees for 3 years from
the date the individual no longer
requires unescorted access;
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(ii) Retain a copy of the current access
authorization program procedures as a
record for 3 years after the procedure is
no longer needed. If any portion of the
procedure is superseded, retain the
superseded material for 3 years after the
record is superseded; and
(iii) Retain the list of persons
approved for unescorted access for 3
years after the list is superseded or
replaced. Records maintained in any
database(s) must be available for NRC
review.
■ 157. In § 73.1200, revise paragraphs
(a) introductory text, (c)(1) introductory
text, (e)(1) introductory text, (e)(3) and
(4), (g)(1) introductory text, (o)(5)(i) and
(o)(6)(i), (r) and (s) to read as follows:
§ 73.1200
events.
Notification of physical security
(a) 15-minute notifications—facilities.
Each licensee subject to the provisions
of § 73.20, § 73.45, § 73.46, § 73.51,
§ 73.55, or § 73.100 must notify the NRC
Headquarters Operations Center, as soon
as possible but within 15 minutes
after—
*
*
*
*
*
(c) * * *
(1) Each licensee subject to the
provisions of §§ 73.20, 73.45, 73.46,
73.50, 73.51, 73.55, 73.60, 73.67, or
73.100 must notify the NRC
Headquarters Operations Center as soon
as possible but no later than 1 hour after
the time of discovery of the following
significant facility security events
involving—
*
*
*
*
*
(e) * * *
(1) Each licensee subject to the
provisions of §§ 73.20, 73.45, 73.46,
73.50, 73.51, 73.55, 73.60, 73.67, or
73.100 must notify the NRC
Headquarters Operations Center within
4 hours after time of discovery of the
following facility security events
involving—
*
*
*
*
*
(3)(i) An event involving a law
enforcement response to the facility that
could reasonably be expected to result
in public or media inquiries and that
does not otherwise require a notification
under paragraphs (a) through (h) of this
section, or in other NRC regulations
such as § 50.72(b)(2)(xi) or
§ 53.1630(b)(2)(v) of this chapter.
(ii) As an exemption, licensees need
not report law enforcement responses to
minor incidents, such as traffic
accidents.
(4) For licensees subject to the
provisions of § 73.55 or § 73.100 of this
part, an event involving the licensee’s
suspension of security measures.
*
*
*
*
*
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(g) * * *
(1) Each licensee subject to the
provisions of § 73.20, § 73.45, § 73.46,
§ 73.50, § 73.51, § 73.55, § 73.60, § 73.67,
or § 73.100 must notify the NRC
Headquarters Operations Center within
8 hours after time of discovery of the
following facility security program
failures involving—
*
*
*
*
*
(o) * * *
(5) * * *
(i) Licensees must establish the
requested continuous communications
channel once the licensee has
completed other required notifications
under this section, § 50.72 of this
chapter, appendix E to part 50 of this
chapter, § 53.1630 of this chapter,
§ 70.50 of this chapter; or § 72.75 of this
chapter; as appropriate.
*
*
*
*
*
(6) * * *
(i) Licensees must establish the
requested continuous communications
channel once the licensee or the
movement control center has completed
other required notifications under this
section, § 50.72 of this chapter,
appendix E to part 50 of this chapter,
§ 53.1630 of this chapter, § 70.50 of this
chapter; § 72.75 of this chapter; or
requested assistance from the LLEA, as
appropriate.
*
*
*
*
*
(r) Declaration of emergencies.
Licensees notifying the NRC of the
declaration of an emergency class must
do so in accordance with §§ 50.72,
53.1630, 63.73, 70.50, and 72.75 of this
chapter, as applicable.
(s) Elimination of duplication.
Licensees with notification obligations
under paragraphs (a) through (h), (m),
and (n) of this section and §§ 50.72,
53.1630, 63.73, 70.50, and 72.75 of this
chapter may notify the NRC of events in
a single communication. This
communication must identify each
regulation under which the licensee is
reporting.
*
*
*
*
*
■ 158. In § 73.1205, revise paragraph
(b)(2) to read as follows:
§ 73.1205 Written follow-up reports of
physical security events.
*
*
*
*
*
(b) * * *
(2)(i) Licensees subject to § 50.73 or
§ 53.1640 of this chapter must prepare
the written follow-up report on NRC
Form 366.
(ii) Licensees not subject to § 50.73 or
§ 53.1640 of this chapter must prepare
the written follow-up report in a letter
format.
*
*
*
*
*
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■
159. In § 73.1210, revise paragraphs
(a)(1) and (b)(3)(i) to read as follows:
§ 74.31 Nuclear material control and
accounting for special nuclear material of
low strategic significance.
§ 73.1210 Recordkeeping of physical
security events.
(a) General performance objectives.
Each licensee who is authorized to
possess and use more than one effective
kilogram of special nuclear material of
low strategic significance, excluding
sealed sources, at any site or contiguous
sites subject to control by the licensee,
other than a production or utilization
facility licensed pursuant to parts 50,
53, or 70 of this chapter, or operations
involved in waste disposal, shall
implement and maintain a Commission
approved material control and
accounting system that will achieve the
following objectives:
*
*
*
*
*
■ 164. In § 74.41, revise paragraph (a)
introductory text to read as follows:
(a) * * *
(1) Licensees with facilities or
shipment activities subject to the
provisions of § 73.20, § 73.25, § 73.26,
§ 73.27, § 73.37, § 73.45, § 73.46, § 73.50,
§ 73.51, § 73.55, § 73.60, § 73.67, or
§ 73.100, must record the physical
security events and conditions adverse
to security that are specified in
paragraphs (c) through (f) of this section.
*
*
*
*
*
(b) * * *
(3)(i) Licensees must record these
physical security events and conditions
adverse to security in either a standalone safeguards event log or as part of
the licensee’s corrective action program,
as specified under the applicable quality
assurance program provisions of parts
50, 52, 53, 60, 63, 70, and 72 of this
chapter, or both.
*
*
*
*
*
■ 160. In § 73.1215, revise paragraph
(d)(1) introductory text to read as
follows:
§ 73.1215
Suspicious activity reports.
*
*
*
*
*
(d) * * *
(1) For licensees subject to the
provisions of §§ 73.20, 73.45, 73.46,
73.50, 73.51, 73.55, 73.60, 73.67, or
73.100, the licensees must report
activities they assess are suspicious.
Examples include, but are not limited
to, the following:
*
*
*
*
*
■ 161. In appendix B to part 73, revise
Definitions introductory text to read as
follows:
Appendix B to Part 73—General
Criteria for Security Personnel
*
*
*
*
*
Definitions
Terms defined in parts 50, 53, 70, and 73
of this chapter have the same meaning when
used in this appendix.
*
*
*
*
*
PART 74—MATERIAL CONTROL AND
ACCOUNTING OF SPECIAL NUCLEAR
MATERIAL
162. The authority citation for 10 CFR
part 74 continues to read as follows:
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■
Authority: Atomic Energy Act of 1954,
secs. 53, 57, 161, 182, 223, 234, 1701 (42
U.S.C. 2073, 2077, 2201, 2232, 2273, 2282,
2297f); Energy Reorganization Act of 1974,
secs. 201, 202 (42 U.S.C. 5841, 5842); 44
U.S.C. 3504 note.
163. In § 74.31, revise paragraph (a)
introductory text to read as follows:
■
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§ 74.41 Nuclear material control and
accounting for special nuclear material of
moderate strategic significance.
(a) General performance objectives.
Each licensee who is authorized to
possess special nuclear material (SNM)
of moderate strategic significance or
SNM in a quantity exceeding one
effective kilogram of strategic special
nuclear material in irradiated fuel
reprocessing operations other than as
sealed sources and to use this material
at any site other than a nuclear reactor
licensed pursuant to parts 50 or 53 of
this chapter; or as reactor irradiated
fuels involved in research,
development, and evaluation programs
in facilities other than irradiated fuel
reprocessing plants; or an operation
involved with waste disposal, shall
establish, implement, and maintain a
Commission-approved material control
and accounting (MC&A) system that will
achieve the following performance
objectives:
*
*
*
*
*
■ 165. In § 74.51, revise paragraph (a)
introductory text to read as follows:
§ 74.51 Nuclear material control and
accounting for strategic special nuclear
material.
(a) General performance objectives.
Each licensee who is authorized to
possess five or more formula kilograms
of strategic special nuclear material
(SSNM) and to use such material at any
site, other than a nuclear reactor
licensed pursuant to parts 50 or 53 of
this chapter, an irradiated fuel
reprocessing plant, an operation
involved with waste disposal, or an
independent spent fuel storage facility
licensed pursuant to part 72 of this
chapter shall establish, implement, and
maintain a Commission-approved
material control and accounting (MC&A)
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87125
system that will achieve the following
objectives:
*
*
*
*
*
PART 75—SAFEGUARDS ON
NUCLEAR MATERIAL—
IMPLEMENTATION OF SAFEGUARDS
AGREEMENTS BETWEEN THE UNITED
STATES AND THE INTERNATIONAL
ATOMIC ENERGY AGENCY
166. The authority citation for 10 CFR
part 75 continues to read as follows:
■
Authority: Atomic Energy Act of 1954,
secs. 53, 63, 103, 104, 122, 161, 223, 234,
1701 (42 U.S.C. 2073, 2093, 2133, 2134, 2152,
2201, 2273, 2282, 2297f); Energy
Reorganization Act of 1974, sec. 201 (42
U.S.C. 5841); Nuclear Waste Policy Act of
1982, secs. 135, 141 (42 U.S.C. 10155, 10161);
44 U.S.C. 3504 note.
167. In § 75.4, revise the introductory
text and paragraph (6) of the definition
for ‘‘Facility’’ to read as follows:
■
§ 75.4
Definitions.
*
*
*
*
*
Unless otherwise defined in this
section, the terms defined in §§ 40.4,
50.2, 53.020, and 70.4 of this chapter
have the same meaning when used in
this part.
*
*
*
*
*
Facility means:
*
*
*
*
*
(6) Any plant or location where the
possession of more than 1 effective
kilogram of nuclear material is licensed
pursuant to 10 CFR part 40, 50, 53, 60,
61, 63, 70, 72, 76, or 150 of this chapter
or an Agreement State license.
*
*
*
*
*
PART 95—FACILITY SECURITY
CLEARANCE AND SAFEGUARDING
OF NATIONAL SECURITY
INFORMATION AND RESTRICTED
DATA
168. The authority citation for 10 CFR
part 95 continues to read as follows:
■
Authority: Atomic Energy Act of 1954,
secs. 145, 161, 223, 234 (42 U.S.C. 2165,
2201, 2273, 2282); Energy Reorganization Act
of 1974, sec. 201 (42 U.S.C. 5841); 44 U.S.C.
3504 note; E.O. 10865, as amended, 25 FR
1583, 3 CFR, 1959–1963 Comp., p. 398; E.O.
12829, 58 FR 3479, 3 CFR, 1993 Comp., p.
570; E.O. 12968, 60 FR 40245, 3 CFR, 1995
Comp., p. 391; E.O. 13526, 75 FR 707, 3 CFR,
2009 Comp., p. 298.
169. In § 95.5, revise the definition for
‘‘License’’ to read as follows:
■
§ 95.5
Definitions.
*
*
*
*
*
License means a license issued under
10 CFR part 50, 52, 53, 54, 60, 63, 70,
or 72.
*
*
*
*
*
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[Amended]
170. In § 95.39(a), remove ‘‘part 52’’
and add in its place ‘‘parts 52 or 53.’’
■
PART 140—FINANCIAL PROTECTION
REQUIREMENTS AND INDEMNITY
AGREEMENTS
171. The authority citation for 10 CFR
part 140 continues to read as follows:
■
Authority: Atomic Energy Act of 1954,
secs. 161, 170, 223, 234 (42 U.S.C. 2201,
2210, 2273, 2282); Energy Reorganization Act
of 1974, secs. 201, 202 (42 U.S.C. 5841,
5842); 44 U.S.C. 3504 note.
172. In § 140.2, revise paragraphs
(a)(1) and (2) to read as follows:
■
§ 140.2
Scope.
(a) * * *
(1) To each person who is an
applicant for or holder of a license
issued under 10 CFR part 50, 52, 53, or
54 to operate a nuclear reactor, and
(2) With respect to an extraordinary
nuclear occurrence, to each person who
is an applicant for or holder of a license
to operate a production facility or a
utilization facility (including an
operating license issued under part 50
or part 53 of this chapter and a
combined license under part 52 or part
53 of this chapter), and to other persons
indemnified with respect to the
involved facilities.
*
*
*
*
*
■ 173. Revise § 140.10 to read as
follows:
§ 140.10
Scope.
This subpart applies to each person
who is an applicant for or holder of a
license issued under 10 CFR parts 50, 53
or 54 to operate a nuclear reactor, or is
the applicant for or holder of a
combined license issued under 10 CFR
parts 52, 53, or 54, except licenses held
by persons found by the Commission to
be Federal agencies or nonprofit
educational institutions licensed to
conduct educational activities. This
subpart also applies to persons licensed
to possess and use plutonium in a
plutonium processing and fuel
fabrication plant.
■ 174. In § 140.11, revise paragraph (b)
to read as follows:
§ 140.11 Amounts of financial protection
for certain reactors.
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*
(b) In any case where a person is
authorized under 10 CFR parts 50, 52,
53, or 54 to operate two or more nuclear
reactors at the same location, the total
primary financial protection required of
the licensee for all such reactors is the
highest amount which would otherwise
be required for any one of those
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reactors; provided, that such primary
financial protection covers all reactors
at the location.
■ 175. In § 140.12, revise paragraph (c)
to read as follows:
PART 150—EXEMPTIONS AND
CONTINUED REGULATORY
AUTHORITY IN AGREEMENT STATES
AND IN OFFSHORE WATERS UNDER
SECTION 274
§ 140.12 Amount of financial protection
required for other reactors.
■
*
*
*
*
*
(c) In any case where a person is
authorized under 10 CFR parts 50, 52,
53, or 54 to operate two or more nuclear
reactors at the same location, the total
financial protection required of the
licensee for all such reactors is the
highest amount which would otherwise
be required for any one of those
reactors; provided, that such financial
protection covers all reactors at the
location.
*
*
*
*
*
■ 176. Revise § 140.13 to read as
follows:
§ 140.13 Amount of financial protection
required of certain holders of construction
permits and combined licenses under 10
CFR part 52.
Each holder of a 10 CFR part 50 or 10
CFR part 53 construction permit, or a
holder of a combined license under
parts 52 or 53 of this chapter before the
date that the Commission had made the
finding under §§ 52.103(g) or 53.1452(g)
of this chapter, who also holds a license
under part 70 of this chapter authorizing
ownership, possession and storage only
of special nuclear material at the site of
the nuclear reactor for use as fuel in
operation of the nuclear reactor after
issuance of either an operating license
under 10 CFR part 50 or 53, or a
combined license under 10 CFR part 52
or 53, shall, during the period before
issuance of a license authorizing
operation under 10 CFR part 50 or 53,
or the period before the Commission
makes the finding under § 52.103(g) or
§ 53.1452(g) of this chapter, as
applicable, have and maintain financial
protection in the amount of $1,000,000.
Proof of financial protection shall be
filed with the Commission in the
manner specified in § 140.15 before
issuance of the license under part 70 of
this chapter.
■ 177. In § 140.20, revise paragraphs
(a)(1)(i) and (ii) to read as follows:
§ 140.20
Indemnity agreements and liens.
(a) * * *
(1)(i) The effective date of the license
(issued under part 50 or part 53 of this
chapter) authorizing the licensee to
operate the nuclear reactor involved; or
(ii) The date that the Commission
makes the finding under §§ 52.103(g) or
53.1452(g) of this chapter; or
*
*
*
*
*
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178. The authority citation for 10 CFR
part 150 continues to read as follows:
Authority: Atomic Energy Act of 1954,
secs. 11, 53, 81, 83, 84, 122, 161, 181, 223,
234, 274 (42 U.S.C. 2014, 2201, 2231, 2273,
2282, 2021); Energy Reorganization Act of
1974, sec. 201 (42 U.S.C. 5841); Nuclear
Waste Policy Act of 1982, secs. 135, 141 (42
U.S.C. 10155, 10161); 44 U.S.C. 3504 note.
179. In § 150.15, revise paragraphs
(a)(7)(iii) and (a)(8) to read as follows:
■
§ 150.15
Persons not exempt.
(a) * * *
(7) * * *
(iii) Greater than Class C (GTCC)
waste, as defined in part 72 of this
chapter, in an ISFSI or an MRS licensed
under part 72 of this chapter; the GTCC
waste must originate in, or be used by,
a facility licensed under parts 50, 52, or
53 of this chapter.
(8) Greater than Class C waste, as
defined in part 72 of this chapter, that
originates in, or is used by, a facility
licensed under parts 50, 52, or 53 of this
chapter and is licensed under part 30
and/or part 70 of this chapter.
*
*
*
*
*
PART 170—FEES FOR FACILITIES,
MATERIALS, IMPORT AND EXPORT
LICENSES, AND OTHER
REGULATORY SERVICES UNDER THE
ATOMIC ENERGY ACT OF 1954, AS
AMENDED
180. The authority citation for 10 CFR
part 170 continues to read as follows:
■
Authority: Atomic Energy Act of 1954,
secs. 11, 161(w) (42 U.S.C. 2014, 2201(w));
Energy Reorganization Act of 1974, sec. 201
(42 U.S.C. 5841); 42 U.S.C. 2215; 31 U.S.C.
901, 902, 9701; 44 U.S.C. 3504 note.
181. In § 170.3, revise the definitions
for ‘‘Manufacturing License,’’ ‘‘Part 55
Reviews,’’ ‘‘Power reactor,’’ and
‘‘Special projects’’ to read as follows:
■
§ 170.3
Definitions.
*
*
*
*
*
Manufacturing license means a
license under subpart F of part 52 of this
chapter or subpart H of part 53 of this
chapter to manufacture a nuclear power
reactor(s) to be operated at sites not
identified in the license application.
*
*
*
*
*
Part 55 Reviews as used in this part
means those services provided by the
Commission to administer
requalification and replacement
examinations and tests for reactor
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operators licensed under 10 CFR part 55
or 53 of the Commission’s regulations
and employed by part 50 or 53
licensees. These services also include
related items such as the preparation,
review, and grading of the examinations
and tests.
*
*
*
*
*
Power reactor means a nuclear reactor
designed to produce electrical or heat
energy licensed by the Commission
under the authority of section 103 or
subsection 104b of the Act, and under
the provisions of §§ 50.21(b), 50.22, or
part 53 of this chapter.
*
*
*
*
*
Special projects means specific
services provided by the Commission
for which fees are not otherwise
specified in this chapter. This includes,
but is not limited to, contested hearings
on licensing actions directly related to
U.S. Government national security
initiatives (as determined by the NRC),
topical report reviews, early site
reviews, waste solidification activities,
activities related to the tracking and
monitoring of shipment of classified
matter, services provided to certify
licensee, vendor, or other private
industry personnel as instructors for 10
CFR part 55 or 53 reactor operators,
reviews of financial assurance
submittals that do not require a license
amendment, reviews of responses to
Confirmatory Action Letters, reviews of
uranium recovery licensees’ land-use
survey reports, and reviews of §§ 50.71
or 53.1545 of this chapter Final Safety
Analysis Reports. Special projects does
not include activities otherwise exempt
from fees under this part. It also does
not include those contested hearings for
which a fee exemption is granted in
§ 170.11(a)(2), including those related to
individual plant security modifications.
*
*
*
*
*
■ 182. In § 170.12, revise paragraph
(d)(1)(v) to read as follows:
§ 170.12
Payment of fees.
*
*
*
*
(d) * * *
(1) * * *
(v) 10 CFR 50.71 or 53.1545 final
safety analysis reports;
*
*
*
*
*
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*
§ 170.21
[Amended]
183. In § 170.21, in footnote 1 remove
the phrase ‘‘(e.g., 10 CFR 50.12, 10 CFR
73.5)’’ and add in its place the phrase
‘‘(e.g., 10 CFR 50.12, 10 CFR 53.080, 10
CFR 73.5)’’.
■ 184. Revise § 170.41to read as follows:
■
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§ 170.41 Failure by an applicant or
licensee to pay prescribed fees.
If the Commission determines that an
applicant or a licensee has failed to pay
a prescribed fee required in this part,
the Commission will not process any
application and may suspend or revoke
any license or approval issued to the
applicant or licensee. The Commission
may issue an order with respect to
licensed activities that the Commission
determines to be appropriate or
necessary to carry out the provisions of
this part, parts 30, 31, 32 through 35, 40,
50, 53, 61, 70, 71, 72, 73, and 76 of this
chapter, and of the Act.
PART 171—ANNUAL FEES FOR
REACTOR LICENSES AND FUEL
CYCLE LICENSES AND MATERIALS
LICENSES, INCLUDING HOLDERS OF
CERTIFICATES OF COMPLIANCE,
REGISTRATIONS, AND QUALITY
ASSURANCE PROGRAM APPROVALS
AND GOVERNMENT AGENCIES
LICENSED BY THE NRC
185. The authority citation for 10 CFR
part 171 continues to read as follows:
■
Authority: Atomic Energy Act of 1954,
secs. 11, 161(w), 223, 234 (42 U.S.C. 2014,
2201(w), 2273, 2282); Energy Reorganization
Act of 1974, sec. 201 (42 U.S.C. 5841); 42
U.S.C. 2215; 44 U.S.C. 3504 note.
■
186. Revise § 171.3 to read as follows:
§ 171.3
Scope.
The regulations in this part apply to
any person holding an operating license
for a test reactor or research reactor
issued under part 50 of this chapter, and
to any person holding an operating
license for a power reactor licensed
under 10 CFR part 50 or 53, or a
combined license issued under 10 CFR
part 52 or 53, that has provided
notification to the U.S. Nuclear
Regulatory Commission (NRC) that the
licensee has successfully completed
power ascension testing. The
regulations in this part also apply to any
person holding a materials license as
defined in this part, a Certificate of
Compliance, a sealed source or device
registration, a quality assurance program
approval, and to a Government agency
as defined in this part. Notwithstanding
the other provisions in this section, the
regulations in this part do not apply to
uranium recovery and fuel facility
licensees until after the Commission
verifies through inspection that the
facility has been constructed in
accordance with the requirements of the
license.
■ 187. In § 171.5, revise the definitions
for ‘‘Operating license,’’ and ‘‘Power
reactor’’ to read as follows:
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§ 171.5
87127
Definitions.
*
*
*
*
*
Operating license means having a
license issued under §§ 50.57 or 53.1387
of this chapter. It does not include
licenses that only authorize possession
of special nuclear material after the
Commission has received a request from
the licensee to amend its licensee to
permanently withdraw its authority to
operate or the Commission has
permanently revoked such authority.
*
*
*
*
*
Power reactor means a nuclear reactor
designed to produce electrical or heat
energy and licensed by the Commission
under the authority of section 103 or
subsection 104b of the Atomic Energy
Act of 1954, as amended, and under the
provisions of §§ 50.21(b) or 50.22, or
part 53 of this chapter.
*
*
*
*
*
■ 188. In § 171.15, revise paragraphs (a),
(b)(2)(iii), (c)(1), and (d)(1) to read as
follows:
§ 171.15 Annual fees: Non-power
production or utilization licenses, reactor
licenses, and independent spent fuel
storage licenses.
(a) Each person holding an operating
license for one or more non-power
production or utilization facilities under
10 CFR part 50 that has provided
notification to the NRC of the successful
completion of startup testing; each
person holding an operating license for
a power reactor licensed under 10 CFR
part 50 or a combined license under 10
CFR part 52, or an operating license or
combined license for a commercial
nuclear plant under 10 CFR part 53, that
has provided notification to the NRC of
the successful completion of power
ascension testing; each person holding a
10 CFR part 50 or 52, power reactor
license, or a 10 CFR part 53 commercial
nuclear plant license that is in
decommissioning or possession only
status, except those that have no spent
fuel onsite; and each person holding a
10 CFR part 72 license who does not
hold a 10 CFR part 50, 52, or 53 license
and provides notification under
§ 72.80(g) of this chapter, shall pay the
annual fee for each license held during
the Federal fiscal year in which the fee
is due. This paragraph (a) does not
apply to test or research reactors
exempted under § 171.11(b).
(b) * * *
(2) * * *
(iii) Generic activities required largely
for NRC to regulate power reactors (e.g.,
updating part 50, part 52, or part 53 of
this chapter, operating the Incident
Response Center, new reactor regulatory
infrastructure). The base annual fee for
operating power reactors does not
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include generic activities specifically
related to reactor decommissioning.
(c)(1) The FY 2022 annual fee for each
power reactor holding a 10 CFR part 50
operating license or combined license
issued under 10 CFR part 52 or part 53
that is in a decommissioning or
possession-only status and has spent
fuel onsite, and for each independent
spent fuel storage 10 CFR part 72
licensee who does not hold a 10 CFR
part 50 or part 53 operating license, or
a 10 CFR part 52 or part 53 combined
license, is $227,000.
*
*
*
*
*
(d)(1) Each person holding an
operating license for an SMR issued
under 10 CFR part 50 or part 53, or a
combined license issued under 10 CFR
part 52 or part 53, that has provided
notification to the NRC of the successful
completion startup testing, shall pay the
annual fee for all licenses held for an
SMR site. The annual fee will be
determined using the cumulative
licensed thermal power rating of all
SMR units and the bundled unit
concept, during the fiscal year in which
the fee is due. For a given site, the use
of the bundled unit concept is
independent of the number of SMR
plants, the number of SMR licenses
issued, or the sequencing of the SMR
licenses that have been issued.
*
*
*
*
*
■ 189. In § 171.17, revise paragraphs (a)
introductory text, (a)(1)(ii), and (a)(2) to
read as follows:
§ 171.17
Proration.
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who do not hold 10 CFR part 50, 52, or
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53 licenses, and materials licenses with
annual fees of $100,000 or greater for a
single fee category. The NRC will base
the proration of annual fees for
terminated and downgraded licenses on
the fee rule in effect at the time the
action is official. The NRC will base the
determinations on the proration
requirements under paragraphs (a)(2)
and (3) of this section.
(1) * * *
(ii) The annual fees for new licenses
for non-power production or utilization
facilities, 10 CFR part 72 licensees who
do not hold 10 CFR part 50, 52, or 53
licenses, and materials licenses with
annual fees of $100,000 or greater for a
single fee category for the current FY,
that are subject to fees under this part
and are granted a license to operate on
or after October 1 of a FY, are prorated
on the basis of the number of days
remaining in the FY. Thereafter, the full
annual fee is due and payable each
subsequent FY.
(2) Terminations. The base operating
power reactor annual fee for operating
reactor licensees or the annual fee for
small modular reactor licensees, who
have requested amendment to withdraw
operating authority permanently during
the FY will be prorated based on the
number of days during the FY the
license was in effect before docketing of
the certifications for permanent
cessation of operations and permanent
removal of fuel from the reactor vessel
or when a final legally effective order to
permanently cease operations has come
into effect. The spent fuel storage/
reactor decommissioning annual fee for
reactor licensees who permanently
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cease operations and have permanently
removed fuel from the site during the
FY will be prorated on the basis of the
number of days remaining in the FY
after docketing of both the certifications
of permanent cessation of operations
and permanent removal of fuel from the
site. The spent fuel storage/reactor
decommissioning annual fee will be
prorated for those 10 CFR part 72
licensees who do not hold a 10 CFR part
50, 52, or 53 license who request
termination of the 10 CFR part 72
license and permanently cease activities
authorized by the license during the FY
based on the number of days the license
was in effect before receipt of the
termination request. The annual fee for
materials licenses with annual fees of
$100,000 or greater for a single fee
category for the current FY will be
prorated based on the number of days
remaining in the FY when a termination
request or a request for a possessiononly license is received by the NRC,
provided the licensee permanently
ceased licensed activities during the
specified period. The annual fee for
non-power production or utilization
facilities will be prorated based on the
number of days remaining in the FY
when the authorization to operate the
facility has been permanently removed
from the license during the FY.
*
*
*
*
*
Dated: October 7, 2024.
For the Nuclear Regulatory Commission.
Carrie Safford,
Secretary of the Commission.
[FR Doc. 2024–23434 Filed 10–23–24; 8:45 am]
BILLING CODE 7590–01–P
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[Federal Register Volume 89, Number 211 (Thursday, October 31, 2024)]
[Proposed Rules]
[Pages 86918-87128]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2024-23434]
[[Page 86917]]
Vol. 89
Thursday,
No. 211
October 31, 2024
Part II
Nuclear Regulatory Commission
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10 CFR Parts 1, 2, 10, et al.
Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced
Reactors; Proposed Rule
Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 /
Proposed Rules
[[Page 86918]]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
10 CFR Parts 1, 2, 10, 11, 19, 20, 21, 25, 26, 30, 40, 50, 51, 53,
70, 72, 73, 74, 75, 95, 140, 150, 170, and 171
[NRC-2019-0062]
RIN 3150-AK31
Risk-Informed, Technology-Inclusive Regulatory Framework for
Advanced Reactors
AGENCY: Nuclear Regulatory Commission.
ACTION: Proposed rule.
-----------------------------------------------------------------------
SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to
revise the NRC's regulations by adding a risk-informed, performance-
based, and technology-inclusive regulatory framework for commercial
nuclear plants in response to the Nuclear Energy Innovation and
Modernization Act (NEIMA). The NRC plans to hold a public meeting to
promote full understanding of the proposed rule and facilitate public
comments.
DATES: Submit comments by December 30, 2024. Comments received after
this date will be considered if it is practical to do so, but the NRC
is able to ensure consideration only for comments received before this
date.
ADDRESSES: You may submit comments by any of the following methods
however, the NRC encourages electronic comment submission through the
Federal rulemaking website:
Federal Rulemaking website: Go to https://www.regulations.gov and search for Docket ID NRC-2019-0062. Address
questions about NRC dockets to Helen Chang; telephone: 301-415-3228;
email: [email protected]. For technical questions contact the
individuals listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Email comments to: [email protected]. If you do
not receive an automatic email reply confirming receipt, then contact
us at 301-415-1677.
Fax comments to: Secretary, U.S. Nuclear Regulatory
Commission at 301-415-1101.
Mail comments to: Secretary, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, ATTN: Rulemakings and
Adjudications Staff.
Hand deliver comments to: 11555 Rockville Pike, Rockville,
Maryland 20852, between 7:30 a.m. and 4:15 p.m. eastern time, Federal
workdays; telephone: 301-415-1677.
You can read a plain language description of this proposed rule at
https://www.regulations.gov/docket/NRC-2019-0062. For additional
direction on obtaining information and submitting comments, see
``Obtaining Information and Submitting Comments'' in the SUPPLEMENTARY
INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Robert Beall, Office of Nuclear
Material Safety and Safeguards, telephone: 301-415-3874; email:
[email protected]; or Anders Gilbertson, Office of Nuclear Reactor
Regulation, telephone: 301-415-1541; email: [email protected].
Both are staff of the U.S. NRC, Washington, DC 20555-0001.
SUPPLEMENTARY INFORMATION:
Executive Summary
A. Need for the Regulatory Action
On January 14, 2019, the President signed the Nuclear Energy
Innovation and Modernization Act (NEIMA) into law (Pub. L. 115-439).
NEIMA section 103(a)(4) directs the NRC to ``complete a rulemaking to
establish a technology-inclusive, regulatory framework for optional use
by commercial advanced nuclear reactor applicants for new reactor
license applications.'' NEIMA defines a ``technology-inclusive
regulatory framework'' as one that is ``developed using methods of
evaluation that are flexible and practicable for application to a
variety of reactor technologies, including, where appropriate, the use
of risk-informed and performance-based techniques.'' NEIMA, as further
amended by the Accelerating Deployment of Versatile, Advanced Nuclear
for Clean Energy Act of 2024 (ADVANCE Act), defines the term ``advanced
nuclear reactor'' as ``a nuclear fission reactor or fusion machine,
including a prototype plant (as defined in sections 50.2 and 52.1 of
title 10, Code of Federal Regulations (as in effect on the date of
enactment of [NEIMA])), with significant improvements compared to
commercial nuclear reactors under construction as of the date of
enactment of [NEIMA].''
The NRC initially considered establishing the scope of proposed
part 53, ``Risk-Informed, Technology-Inclusive Regulatory Framework for
Commercial Nuclear Plants,'' of title 10 of the Code of Federal
Regulations (10 CFR) as being for ``advanced nuclear plants''
consisting of one or more ``advanced nuclear reactors'' as defined in
NEIMA. Based on public discussions on the use of the term, the NRC
determined that the NEIMA definition, although broad, did not define
``significant improvements'' with enough specificity to implement in
NRC regulations. Additionally, a number of stakeholders suggested that
the descriptor, ``advanced,'' implied enhanced safety, while the NEIMA
definition includes ``significant improvements'' in areas other than
safety enhancements. In response to this feedback, and to be technology
inclusive, the NRC determined that the broader term ``commercial
nuclear plant'' would be preferable.
The current application and licensing requirements in 10 CFR part
50, ``Domestic Licensing of Production and Utilization Facilities,''
and 10 CFR part 52, ``Licenses, Certifications, and Approvals for
Nuclear Power Plants,'' were primarily developed to address license
requests concerning water-cooled reactors, and to address operational
requirements for those types of reactors. This proposed rule responds
to NEIMA by creating an alternative regulatory framework for licensing
future commercial nuclear plants. The new alternative requirements and
implementing guidance would adopt technology-inclusive approaches and
use risk-informed and performance-based techniques to ensure an
equivalent level of safety to that of operating commercial nuclear
plants while providing flexibility for licensing and regulating a
variety of technologies and designs for commercial nuclear reactors.
B. Major Provisions
Major provisions of this proposed rule, supported by accompanying
guidance, include the following:
A new alternative technology-inclusive, risk-informed,
performance-based framework that includes requirements for licensing
and regulating nuclear plants during the various stages of their life
cycles.
A new alternative technology-inclusive, risk-informed, and
performance-based framework in 10 CFR part 26, ``Fitness for Duty
Programs,'' developed from existing requirements in subpart K, ``FFD
Programs for Construction,'' of part 26.
A new alternative technology-inclusive and performance-
based security framework in 10 CFR part 73, ``Physical Protection of
Plants and Materials,'' that includes requirements for protection of
licensed activities at commercial nuclear plants.
C. Costs and Benefits
The NRC prepared a draft regulatory analysis to determine the
expected quantitative costs and benefits of this proposed rule and
associated guidance as well as qualitative factors to be considered in
the NRC's rulemaking decision. The conclusion from the
[[Page 86919]]
analysis is that this proposed rule and associated guidance would
result in net averted costs to the industry and the NRC ranging from
$53.6 million using a 7-percent discount rate to $68.2 million using a
3-percent discount rate, using an assumption of one applicant under 10
CFR part 53. As the number of applicants increases, so do the estimated
averted costs.
The draft regulatory analysis also considers qualitative factors,
such as greater regulatory stability, predictability, and clarity to
the licensing process. These benefits would result from incorporating
advances in probabilistic risk assessment (PRA) and other risk-informed
analyses and codifying regulatory enhancements that currently exist in
regulatory guides (RGs). Another qualitative factor is promoting a
performance-based regulatory framework that specifies requirements to
be met and provides flexibility to an applicant or licensee regarding
the information or approach needed to satisfy those requirements.
For more information, please see the draft regulatory analysis
(available in the NRC's Agencywide Documents Access and Management
System (ADAMS) Accession No. ML21165A112).
Table of Contents
I. Obtaining Information and Submitting Comments
A. Obtaining Information
B. Submitting Comments
II. Background
A. NRC Advanced Reactor Readiness
B. Stakeholder Views on Part 53 Preliminary Proposed Rule
Language
III. Discussion
A. Objective and Applicability
B. Need for Changes to the Existing Regulatory Framework
C. 10 CFR Part 53: Framework
IV. Part 53: Framework
Subpart A--General Provisions
A. Discussion of Definitions in Proposed Part 53
B. Other General Provisions
Subpart B--Technology-Inclusive Safety Requirements
Subpart C--Design and Analysis Requirements
Subpart D--Siting Requirements
Subpart E--Construction and Manufacturing Requirements
Subpart F--Requirements for Operation
Subpart G--Decommissioning Requirements
Subpart H--Licenses, Certifications, and Approvals
Subpart I--Maintaining and Revising Licensing Basis Information
Subpart J--Reporting and Other Administrative Requirements
Subpart M--Enforcement
V. Changes to Other Parts of 10 CFR Chapter I
10 CFR Part 26
A. Introduction
B. Proposed Changes to Part 26, Subparts A Through E and I
C. Proposed Requirements for Part 26, Subpart M
D. Proposed Changes to Part 26, Subpart N
E. Proposed Changes to Part 26, Subpart O
10 CFR Part 50
A. Section 50.160: Emergency Preparedness for Small Modular
Reactors, Non-Light-Water Reactors, and Non-Power Production or
Utilization Facilities
B. Appendix B to Part 50: Quality Assurance Criteria for Nuclear
Power Plants and Fuel Reprocessing Plants
10 CFR Part 73
A. Section 73.100: Technology-Inclusive Requirements for
Physical Protection of Licensed Activities at Commercial Nuclear
Plants Against Radiological Sabotage
B. Section 73.110: Technology-Inclusive Requirements for
Protection of Digital Computer and Communication Systems and
Networks
C. Section 73.120: Access Authorization Program for Commercial
Nuclear Plants
VI. Specific Requests for Comments
VII. Section-by-Section Analysis
VIII. Regulatory Flexibility Certification
IX. Regulatory Analysis
X. Backfitting and Issue Finality
XI. Cumulative Effects of Regulation
XII. Plain Writing
XIII. Environmental Assessment and Proposed Finding of No
Significant Environmental Impact
XIV. Paperwork Reduction Act
XV. Criminal Penalties
XVI. Voluntary Consensus Standards
XVII. Availability of Guidance
XVIII. Public Meeting
XIX. Availability of Documents
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2019-0062 when contacting the NRC
about the availability of information for this action. You may obtain
publicly available information related to this action by any of the
following methods:
Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2019-0062.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly available documents online in the
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``Begin Web-based ADAMS
Search.'' For problems with ADAMS, please contact the NRC's Public
Document Room (PDR) reference staff at 1-800-397-4209, at 301-415-4737,
or by email to [email protected]. For the convenience of the reader,
instructions about obtaining materials referenced in this document are
provided in the ``Availability of Documents'' section.
NRC's PDR: The PDR, where you may examine and order copies
of publicly available documents, is open by appointment. To make an
appointment to visit the PDR, please send an email to
[email protected] or call 1-800-397-4209 or 301-415-4737, between 8
a.m. and 4 p.m. eastern time, Monday through Friday, except Federal
holidays.
B. Submitting Comments
The NRC encourages electronic comment submission through the
Federal rulemaking website (https://www.regulations.gov). Please
include Docket ID NRC-2019-0062 in your comment submission. To
facilitate NRC review, please distinguish between comments on the
proposed rule and comments on the proposed guidance.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC will post all comment submissions at
https://www.regulations.gov as well as enter the comment submissions
into ADAMS. The NRC does not routinely edit comment submissions to
remove identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.
II. Background
A. NRC Advanced Reactor Readiness
In its ``Policy Statement on the Regulation of Advanced Nuclear
Power Plants,'' dated July 8, 1986, the Commission stated that it
considered the term ``advanced'' to apply to reactors that are
significantly different from current (i.e., current in 1986) generation
light-water reactors (LWRs) then under construction or in operation,
and that ``advanced'' includes reactors that provide enhanced margins
of safety or utilize simplified inherent or other innovative means to
accomplish their safety functions. At the time, certain high
temperature gas-cooled reactors, liquid metal reactors, and LWRs of
innovative design were considered to be ``advanced.'' The 1986 policy
statement
[[Page 86920]]
provided the Commission's policy regarding the review of, and desired
characteristics associated with, advanced reactors. The NRC updated
this statement in the ``Policy Statement on the Regulation of Advanced
Reactors,'' dated October 14, 2008 (Advanced Reactor Policy Statement).
The agency has undertaken many activities related to advanced
reactors, including issuing an advance notice of proposed rulemaking
titled, ``Approaches to Risk-Informed and Performance-Based
Requirements for Nuclear Power Reactors,'' dated May 4, 2006 (71 FR
26267). These efforts were often done in parallel, and sometimes
interwoven, with the NRC's efforts to improve risk-informed and
performance-based approaches within the agency (e.g., the Commission's
policy statement, ``Use of Probabilistic Risk Assessment Methods in
Nuclear Regulatory Activities,'' dated August 16, 1995 (PRA Policy
Statement)).
In 2016, the NRC issued ``NRC Vision and Strategy: Safely Achieving
Effective and Efficient Non-Light-Water Mission Readiness'' (Advanced
Reactor Vision and Strategy Document), in response to increasing
interest in advanced reactor designs. The NRC considered the Department
of Energy's (DOE's) advanced reactor deployment goals in developing the
Advanced Reactor Vision and Strategy Document. Since publication of the
document, the NRC continues to manage its activities to support the
DOE's deployment goals. The Advanced Reactor Vision and Strategy
Document identified initiating and developing a new risk-informed and
performance-based regulatory framework as a possible long-term goal.
However, the NRC staff's initial efforts were focused on resolving
policy issues and developing guidance for licensing non-LWR
technologies under the existing regulatory frameworks (parts 50 and
52). The NRC staff issues annual Commission papers on the status and
progress of the NRC staff's activities related to advanced reactors
(e.g., SECY-24-0020, ``Advanced Reactor Program Status,'' dated
February 27, 2024). These Commission papers provide status updates for
advanced reactor activities undertaken both prior to and after
initiation of this rulemaking.
In 2017, the NRC staff prioritized activities to support the
development of technology-inclusive, risk-informed, and performance-
based licensing approaches that could be implemented under the existing
regulatory framework in parts 50 and 52. One key element of these
efforts was the Licensing Modernization Project (LMP), a cost-shared
initiative led by nuclear utilities and supported by DOE. The LMP is a
technology-inclusive, risk-informed, and performance-based methodology
developed for non-LWR designs. The LMP provides a systematic and
reproducible process for licensing-basis event (LBE) selection and
evaluation; classification of structures, systems, and components
(SSCs); and assessment of defense in depth. The LMP refined the DOE's
Next Generation Nuclear Plant Program methodologies to reflect
interactions with the NRC, to address feedback from industry, and to
broaden the scope of the approach to ensure applicability to various
non-LWR technologies. The LMP activities led to the publication and
submittal of Nuclear Energy Institute (NEI) 18-04, Revision 1, ``Risk-
Informed Performance-Based Technology Inclusive Guidance for Non-Light
Water Reactor Licensing Basis Development,'' issued August 2019. The
document indicates that controlling the frequencies and potential
consequences of a wide spectrum of events is the primary focus of the
LMP approach.
The NRC endorsed the principles and methodology in NEI 18-04, with
clarifications, in RG 1.233, ``Guidance for a Technology-Inclusive,
Risk-Informed, and Performance-Based Methodology to Inform the
Licensing Basis and Content of Applications for Licenses,
Certifications, and Approvals for Non-Light-Water Reactors.'' The NRC
staff sought Commission approval of the use of LMP and NEI-18-04 in
SECY-19-0117, ``Technology-Inclusive, Risk-Informed, and Performance-
Based Methodology to Inform the Licensing Basis and Content of
Applications for Licenses, Certifications, and Approvals for Non-Light-
Water Reactors,'' dated December 2, 2019. In that paper, the staff
described the relationship between the LMP and NEI-18-04 and previous
relevant Commission decisions, including those described in SECY-93-
092, ``Issues Pertaining to the Advanced Reactor (PRISM, MHTGR, and
PIUS) and CANDU 3 Designs and their Relationship to Current Regulatory
Requirements,'' dated April 8, 1993. The Commission approved the use of
the LMP methodology and NEI-18-04 as a reasonable approach for
establishing key parts of the licensing basis and content of
applications for licenses, certifications, and approvals for non-LWRs
in Staff Requirements Memorandum (SRM) SRM-SECY-19-0117, dated May 26,
2020. Although the LMP approach is technology- inclusive, the industry
and NRC staff initially focused the LMP's applicability on non-LWRs,
both for efficiency and to support near-term non-LWR applications under
the existing regulatory framework, such as the Advanced Reactor
Demonstration Projects supported by DOE.
As stated in the part 53 rulemaking plan, SECY-20-0032, the NRC
staff developed part 53 by building upon recent and ongoing activities
such as the LMP approach described in SECY-19-0117. Such an approach
supports implementing the NEIMA requirement to use, where appropriate,
risk-informed and performance-based techniques, and it also capitalizes
on previous initiatives by the industry, DOE, and the NRC, including
the LMP. This approach highlights the role of PRA in risk-informed and
performance-based approaches to identifying enhanced safety margins
that can be used to justify operational flexibilities. The proposed
framework is largely based on the methodology described in SECY-19-0117
and includes a prominent role for PRA.
As discussed in section II.B, ``Stakeholder Views on Part 53
Preliminary Proposed Rule Language,'' of this document, the NRC
conducted extensive public outreach on early versions of the proposed
rule text. Early versions of the draft proposed rule included two
alternative regulatory frameworks. One framework (called ``Framework
A'') offered a licensing approach centered largely on risk analysis and
the other framework (called ``Framework B'') largely replicated the
existing licensing approach in parts 50 and 52 but modified it to be
technology neutral. In its SRM to SECY-23-0021, ``Proposed Rule: Risk-
Informed, Technology-Inclusive Regulatory Framework for Advanced
Reactors (RIN 3150-AK31),'' the Commission disapproved the inclusion of
Framework B in this proposed rule and directed the staff to provide
them within one year an options paper for possible future use of the
Framework B methodology.
B. Stakeholder Views on Part 53 Preliminary Proposed Rule Language
In SRM-SECY-20-0032, the Commission directed the NRC staff to
prepare and release preliminary proposed rule language, followed by
public outreach and dialogue, and then further revise the language
until the NRC staff had established the rudiments of its proposed rule
for Commission consideration. To implement the Commission's direction,
the NRC staff undertook an unprecedented program of stakeholder
engagement, recognizing the importance of this rulemaking to the
advanced reactor community and
[[Page 86921]]
interested stakeholders from a broad range of backgrounds and
organizations.
On November 6, 2020, the NRC published a notification in the
Federal Register (85 FR 71002) describing plans for the periodic
release of preliminary proposed rule language, meetings with
stakeholders, and the ability of stakeholders to provide input during
the development of this proposed rule. Sections of the preliminary
proposed rule language were subsequently released, and the NRC held
numerous public meetings to discuss the preliminary proposed rule
language and obtain input from stakeholders. On December 10, 2021, the
NRC published a second notification in the Federal Register (86 FR
70423) announcing that the development of the proposed rule and related
interactions with stakeholders were being extended until August 31,
2022.
By the close of the public stakeholder interactions on August 31,
2022, the NRC staff had held 24 public meetings since September 2020.
The NRC staff also met with the Advisory Committee on Reactor
Safeguards (ACRS) in 16 public meetings during this period. By the
close of the public engagement period on the preliminary proposed rule
language, 126 letters were received on the preliminary proposed rule
language. Of these 126 letters, 21 were from non-governmental
organizations, 31 were from the public, one was from Congress, and the
remaining 73 letters were from NRC licensees, the NEI, and other
industry groups. In addition, the ACRS wrote four interim letter
reports to the Chair on this rulemaking and issued its final letter
report on November 22, 2022. The letters from stakeholders provided
various points of view and suggestions for clarifications, additions,
and deletions to the preliminary proposed rule language. Copies of
these letters may be viewed and downloaded from the Federal rulemaking
website https://www.regulations.gov, under docket number NRC-2019-0062.
The inputs received were considered in the development of this proposed
rule. However, as described during the various public interactions
related to this rulemaking and in supporting documents, the NRC will
not formally disposition the questions and suggestions related to the
preliminary proposed rule language as it will for public comments
received following the publication of this proposed rule.
III. Discussion
A. Objective and Applicability
The NRC is proposing to add a new, alternative part to its
regulations that would set out a risk-informed, technology-inclusive
framework for the licensing and regulation of commercial nuclear
plants. This new approach would achieve the following: (1) continue to
provide reasonable assurance of adequate protection of public health
and safety and the common defense and security; (2) promote regulatory
stability, predictability, and clarity; (3) reduce requests for
exemptions from the current requirements in parts 50 and 52; (4)
establish new requirements to address non-LWR technologies; (5)
recognize technological advancements in reactor design; and (6) credit
the possible response of some designs of commercial nuclear plants to
postulated accidents, including slower transient response times and
relatively small and slow release of fission products. This proposed
rule would add 10 CFR part 53; subpart M, ``Fitness for Duty Programs
for Facilities Licensed Under 10 CFR Part 53,'' to Part 26; Sec.
73.100, ``Technology-inclusive requirements for physical protection of
licensed activities at commercial nuclear plants against radiological
sabotage,'' Sec. 73.110, ``Technology-inclusive requirements for
protection of digital computer and communication systems and
networks,'' and Sec. 73.120, ``Access authorization program for
commercial nuclear plants,'' as well as make conforming changes
throughout 10 CFR chapter I, ``Nuclear Regulatory Commission.''
B. Need for Changes to the Existing Regulatory Framework
The NRC has long recognized that the licensing and regulation of a
variety of nuclear reactor technologies would present challenges
because the existing regulatory framework has evolved primarily to
address the LWR designs that compose the current operating fleet
(widely referred to as Generation II reactors). The NRC has had many
interactions with designers of various reactor technologies under
development, sometimes collectively referred to as advanced reactors
(widely referred to as Generation III/III+ (i.e., evolutionary light-
water) and Generation IV (i.e., non-light-water) reactors). The
interactions have informed the development of policies and guidance to
support the potential licensing of new and different types of reactor
facilities, some of which may not utilize LWR designs. The NRC issued
its Advanced Reactor Policy Statement to provide all interested
parties, including the public, with the Commission's views concerning
the desired characteristics of advanced reactor designs. The NRC
further described its early efforts to establish a technology-inclusive
approach to the regulation of nuclear reactors in the advance notice of
proposed rulemaking published in 2006. The NRC acknowledged in its
``Report to Congress: Advanced Reactor Licensing,'' issued August 2012,
that while the safety philosophy inherent in the current regulations
applies to all reactor technologies, the specific and prescriptive
aspects of those regulations clearly focus on the current fleet of LWR
facilities.
Congress similarly recognized the potential benefits of developing
a regulatory infrastructure to support the development and
commercialization of advanced nuclear reactors. Consequently, Congress
passed NEIMA in late 2018, and the President signed it into law in
January 2019. NEIMA directed the NRC to undertake a rulemaking to
establish a technology-inclusive regulatory framework for optional use
by applicants for new commercial advanced nuclear reactor licenses. In
addition, on July 9, 2024, the President signed into law the
Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy
Act of 2024, also referred to as the ADVANCE Act. The NRC is evaluating
its plans for implementing the ADVANCE Act, including how its
regulations, as well as the proposed part 53 or future revisions to it,
could be used to address provisions in the ADVANCE Act. The ADVANCE Act
contains provisions on a variety of nuclear-related topics, such as
micro reactors, nuclear reactor license application reviews, and
nuclear fuel. In Section VI, ``Specific Requests for Comments,'' the
NRC is requesting public input on how part 53 could be revised to
better enable its potential use to implement the ADVANCE Act.
The requirements in part 53 would support a wide variety of
potential commercial nuclear reactor technologies. As noted in this
discussion, the current regulatory framework in parts 50 and 52 evolved
in the context of the current operating reactor fleet dominated by LWRs
and as a result includes provisions specific to LWR technologies. While
the NRC can license other reactor technologies under the current
framework by using existing regulatory flexibilities and the exemption
process, there is significant interest in developing a regulatory
framework that is flexible enough to accommodate multiple technologies
and robust enough to ensure a level of safety equivalent to parts 50
and 52, consistent with the Commission's Advanced Reactor Policy
Statement. The Commission reiterated its safety
[[Page 86922]]
expectations for new reactors in the SRM for SECY-10-0121, ``Modifying
the Risk-Informed Regulatory Guidance for New Reactors,'' dated March
2, 2011:
Because new plant designs incorporate operating experience from
current generation reactors, severe accident research, and risk
insights from design probabilistic risk assessments, the Commission
expects that the advanced technologies incorporated in new reactors
will result in enhanced margins of safety. However, the Commission
continues to expect (consistent with the 2008 Advanced Reactor
Policy Statement), as a minimum, at least the same degree of
protection of the public and the environment that is required for
current-generation light-water reactors. New reactors with these
enhanced margins and safety features should have greater operational
flexibility than current reactors.
However, developing a regulatory framework that can accommodate a
wide range of technologies while maintaining an acceptable level of
safety presents significant regulatory challenges. The existing
regulations have been developed over the course of decades and reflect
changes to address events discovered through operating experience. In
contrast, part 53 is being developed to accommodate technologies that,
in some cases, lack significant operating experience. To address these
challenges, the NRC drew on well-developed approaches to licensing to
produce a technology-neutral and robust regulatory framework. The
proposed regulatory framework would use PRAs to assess risks, help
establish technical requirements, and manage operations. The framework
builds on the LMP, which is a technology-inclusive approach to
licensing that leverages insights from a detailed PRA to provide
applicants with significant design and operation flexibilities.
C. 10 CFR Part 53: Framework
This proposed rule consists of several major components, including
a new part 53, to be added to 10 CFR chapter I, revisions for part 26,
part 50, and part 73, and conforming changes throughout 10 CFR chapter
I.
Part 53 is comprised of subparts A through M. These provisions are
organized to provide high-level performance criteria and to specify
requirements to demonstrate compliance with those performance criteria
throughout major stages of the life cycle of commercial nuclear plants.
This organization reflects a systems-engineering style approach to the
design, licensing, operation, and ultimately decommissioning of future
commercial nuclear plants. Organizing requirements in this manner also
supports performance-based approaches. Required programs (e.g.,
radiation protection) and monitoring (e.g., technical specification
(TS) surveillance) during the operations phase that are similar to
those required by part 50 would complement the design and analysis
requirements in subpart C. The performance-based approach proposed in
part 53 also includes regulatory requirements that would allow
applicants to use a flexible and graded approach to the performance of
safety functions based on the role of a particular SSC, human action,
or program in limiting the overall risks to the public below accepted
standards through balanced measures to prevent and mitigate possible
events.
Proposed subpart M of part 26 would be new and would be largely
consistent with the objective-based fitness for duty (FFD) requirements
in current subpart K, ``FFD Programs for Construction,'' of part 26
supplemented by select requirements from subparts A through I, N, and O
of part 26. These requirements are designed to ensure program
effectiveness, maintain protections afforded to individuals subject to
the FFD program, and align with FFD program implementation by parts 50
and 52 licensees. The proposed requirements are not entirely equivalent
because current subpart K of part 26 only applies during construction
of the commercial nuclear plant, whereas proposed subpart M of part 26
would apply during construction, operation, and decommissioning.
Furthermore, proposed subpart M of part 26 would allow the use of a
variety of biological specimens for drug testing as well as innovative
technologies for drug and alcohol screening and testing that are not
described or allowed by the requirements in subparts A through K, N,
and O of part 26, except under limited conditions.
Proposed revisions to part 73 would establish a new technology-
inclusive consequence-based approach for a range of security areas,
including physical security, cybersecurity, and access authorization
(AA) for commercial nuclear reactors. The NRC used operating experience
to include additional regulatory flexibility for a part 53 licensee's
implementation of security requirements.
In addition, this proposed rule would make conforming changes
throughout 10 CFR chapter I, by adding ``and part 53'' where
appropriate to account for the addition of the proposed part 53.
IV. Part 53: Framework
Subpart A--General Provisions
Subpart A would provide the general provisions applicable to all
applicants and licensees that would be established in part 53 for the
issuance, amendment, and termination of licenses, permits,
certifications, and approvals for commercial nuclear plants licensed
under Section 103 of the Atomic Energy Act of 1954, as amended (the
Act) and title II of the Energy Reorganization Act of 1974 (88 Stat.
1242). Subpart A would include purpose, scope, definitions, written
communications, employee protections, completeness and accuracy of
information, exemptions, standards for review, jurisdictional limits,
consideration of attacks and destructive acts by enemies of the United
States, and information collection requirements.
The requirements in subpart A would be largely equivalent to the
general requirements in part 50 that are applicable to all part 50
applicants and licensees (specifically, Sec. Sec. 50.1 through 50.13)
but would reference the corresponding regulations in part 53 in place
of references to part 50.
A. Discussion of Definitions in Proposed Part 53
This proposed rule would include a definition section in Sec.
53.020. The definitions of most terms in Sec. 53.020 would be
equivalent to the corresponding terms defined in: (1) Sec. Sec. 50.2,
52.1, and other NRC regulations; (2) NEI 18-04, as endorsed by RG
1.233; or (3) American Society of Mechanical Engineers (ASME)/American
Nuclear Society Risk Assessment Standard (RA-S)-1.4-2021, as endorsed
for trial use by RG 1.247, ``Acceptability of Probabilistic Risk
Assessment Results for Non-Light-Water Reactor Risk-Informed
Activities.'' This is intended to provide clarity and consistency in
terminology where possible and to utilize past and ongoing NRC
initiatives to support the licensing of new reactors. Specific
deviations from existing definitions are further explained in the
following paragraphs.
Regarding the definition of ``Commercial nuclear plant'' and
``Commercial nuclear reactor'' in proposed Sec. 53.020, as noted
previously, the NRC initially considered establishing the scope of part
53 as being for ``advanced nuclear plants.'' The preliminary proposed
rule language defined ``advanced nuclear plant'' as ``a utilization
facility consisting of one or more advanced nuclear reactors'' as
defined in NEIMA. NEIMA defines the term ``advanced nuclear reactor''
as ``a
[[Page 86923]]
nuclear fission reactor or fusion machine, including a prototype plant
(as defined in sections 50.2 and 52.1 of title 10, Code of Federal
Regulations (as in effect on the date of enactment of this Act)), with
significant improvements compared to commercial nuclear reactors under
construction as of the date of enactment of this Act, including
improvements such as--(A) additional inherent safety features; (B)
significantly lower levelized cost of electricity; (C) lower waste
yields; (D) greater fuel utilization; (E) enhanced reliability; (F)
increased proliferation resistance; (G) increased thermal efficiency;
or (H) ability to integrate into electric and nonelectric
applications.''
Based on public discussions on the use of the term, the NRC
determined that the NEIMA definition, although broad, did not define
``significant improvements'' with enough specificity to implement in
NRC regulations. Additionally, a number of stakeholders suggested that
the descriptor, ``advanced,'' implied enhanced safety, while the NEIMA
definition includes ``significant improvements'' in areas other than
safety enhancements. In response to this feedback, and to be technology
inclusive, the NRC determined that the broader term ``commercial
nuclear plant'' would be preferable. The NEIMA definition of advanced
nuclear reactor also includes fusion technologies. Fusion energy
systems have not been included in the scope of part 53 but are the
subject of a separate rulemaking activity, ``Regulatory Framework for
Fusion Systems.'' See NRC docket ID NRC-2023-0017 on the Federal
rulemaking website https://www.regulations.gov.
The NRC proposes to allow use of part 53 by any ``commercial
nuclear plant.'' The use of the term ``plant'' versus ``reactor,'' as
used in existing regulations (i.e., Sec. 50.2), recognizes that co-
located support facilities and radionuclide sources need to be
considered in the licensing of a facility. The phrase ``commercial
purposes,'' as used in the definition of ``commercial nuclear plant,''
includes purposes such as providing process heat for a variety of
industrial applications (e.g., desalination, oil refining, hydrogen
production). The NRC has not compiled a complete list of such
commercial purposes. The definition of ``Commercial nuclear plant''
refers to a ``Commercial nuclear reactor,'' which is defined based on
the definition of ``Nuclear reactor'' in Sec. 50.2. However, the
phrase ``in a self-supporting chain reaction'' was removed from the
definition to enable applying part 53 to accelerator driven systems
that use special nuclear material (SNM) but that do not involve self-
sustaining chain reactions. Relatedly, ``Utilization facility'' is also
defined in Sec. 53.020 based on the definition of that term in Sec.
50.2 but is also revised to refer to a ``Commercial nuclear plant'' as
defined in Sec. 53.020.
The NRC proposes to include a definition of ``Consensus code or
standard'' in part 53 that is based on the use of these terms in the
National Technology Transfer and Advancement Act of 1995 (NTTAA) (Pub.
L. 104-113) and the Office of Management and Budget (OMB) Circular No.
A-119, ``Federal Participation in the Development and Use of Voluntary
Consensus Standards and in Conformity Assessment Activities.'' As
required by NTTAA, the NRC undertakes the following activities: (i)
consults with voluntary consensus standards bodies; (ii) participates
with voluntary consensus bodies in the development of consensus
standards; and (iii) uses consensus standards as a means to carry out
the NRC's policy objectives. In part 53, the NRC is not proposing to
incorporate by reference specific codes and standards as is done under
the existing regulations in Sec. 50.55a, ``Codes and standards,''
because some codes and standards are LWR-specific. Part 53 would
require that design features must be designed using generally accepted
consensus codes and standards but would not incorporate the specific
code or standard into the NRC's regulations. During public meetings,
significant discussions with stakeholders indicated that future reactor
designers were interested in the use of international consensus
standards that have not yet been endorsed by the NRC. The definition
proposed in part 53 would allow for the use of international codes and
standards not previously used in NRC licensing but recognizes that the
use of any consensus code or standard would ultimately need to be found
acceptable by the NRC, either through generic efforts to endorse a code
or standard or on an application-specific basis during an individual
licensing review.
The proposed definition of ``Construction'' is slightly different
than the definition in Sec. 50.10--it would cover the same concept but
be applied to a slightly different scope of activities based on how
SSCs are classified under part 53. In part 53, the definition of
``Construction'' is based on the definition in Sec. 50.10 but modified
to apply to safety-related (SR) and non-safety-related but safety-
significant (NSRSS) SSCs identified by the design and analysis
requirements in subparts B and C to ensure the safety criteria are met.
Section 53.020 would also add definitions for terms related to
event selection (LBEs, design-basis accidents (DBAs), anticipated event
sequences, unlikely event sequences, and very unlikely event
sequences); equipment classifications (SR, NSRSS, and non-safety-
significant SSCs); performance metrics (e.g., safety criteria and
functional design criteria); and special treatment.
The regulation would define ``Safety criteria'' in terms of the
plant-level performance-based metrics that would be provided in
Sec. Sec. 53.210 and 53.220. The term ``Functional design criteria''
would be defined as metrics for the performance of specific SSCs that
are determined from the role of the SSC in meeting the safety criteria.
These are new terms that have not previously been defined or used in
NRC regulation.
The term ``Safety-related SSCs'' would refer to those SSCs needed
to meet the safety criteria in Sec. 53.210. The term ``Non-safety-
related but safety-significant SSCs'' would mean those SSCs that are
not SR because they are not relied upon to perform any function
necessary to demonstrate compliance with Sec. 53.210 but warrant
special treatment because they are relied on to achieve adequate
defense in depth or perform risk-significant functions. The term
``Special treatment'' would be defined as requirements, such as quality
assurance and programmatic controls, identified for each design feature
to ensure that the safety criteria are satisfied and the safety
functions are fulfilled. These requirements would also ensure that SR
and NSRSS SSCs will provide defense in depth, or perform risk-
significant functions, under service conditions and with SSC
reliabilities that are consistent with the analysis required in
proposed subpart C. Structures, systems, and components designated as
SR would also contribute to defense in depth and risk-significant
functions and may warrant special treatments beyond those defined for
the SR functions needed for compliance with Sec. 53.210. The term
``Non-safety-significant SSCs'' would mean those SSCs that are not SR
or NSRSS.
The terms ``Design-basis accidents,'' ``Anticipated event
sequences,'' ``Unlikely event sequences,'' and ``Very unlikely event
sequences'' would be defined to be different types of ``Licensing-basis
events'' and would also be largely equivalent to the LMP's definitions
of DBAs, anticipated operational occurrences (AOOs), design-basis
events (DBEs), and beyond-design-basis events, respectively. The term
[[Page 86924]]
``Design-basis accidents'' would be defined as postulated event
sequences that are used to set functional design criteria and
performance objectives for the design of SR SSCs through deterministic
analyses. Design-basis accidents would be derived from the unlikely
event sequences from the PRA and then analyzed in a conservative
approach by prescriptively assuming that only SR SSCs are available to
mitigate postulated accident scenarios. Within the LMP methodology,
event sequences with mean frequencies of 1 x 10-2/plant-year
and greater would be classified as anticipated event sequences. Within
the LMP methodology, infrequent event sequences with mean frequencies
of 1 x 10-4/plant-year to 1 x 10-2/plant-year
would be classified as unlikely event sequences. ``Very unlikely event
sequences'' would be less likely to occur than unlikely event
sequences. Within the LMP methodology, rare event sequences with
frequencies of 5 x 10-7/plant-year to 1 x 10-4/
plant-year would be classified as very unlikely event sequences. While
the proposed terminology for these event sequences would create some
differences between part 53 and the LMP, part 53 would use new terms
for these event sequences specifically to avoid conflicts with terms
already used within part 50 and part 52 to represent different
concepts. Further, because some stakeholder comments demonstrated
confusion related to the history of beyond-design-basis accidents
terminology, these definitions seek to clarify the event categories in
part 53. The sections of this preamble related to subparts B and C
provide additional discussion of LBEs.
B. Other General Provisions
Section 53.040 would govern written communications and how
applications and other required information must be submitted to the
NRC. These requirements would be equivalent to those in Sec. 50.4.
Section 53.050 would establish requirements for enforcement action
to which a licensee, an applicant, or a licensee's or applicant's
contractor or subcontractor, or an employee of any of them may be
subject for engaging in deliberate misconduct. These requirements would
be equivalent to those in Sec. 50.5.
Section 53.060 would prohibit discrimination against an employee of
a holder or applicant for an NRC license, permit, design certification
(DC), or design approval, or a contractor or subcontractor of a holder
or applicant for an NRC license, permit, DC, or design approval for
engaging in certain protected activities. Section 53.060 also would
prescribe a procedure for seeking a remedy for employees who believe
they have been discriminated against for engaging in such protected
activities. These requirements would be equivalent to those in
Sec. Sec. 50.7 and 52.5.
Section 53.070 would govern the completeness and accuracy of
information provided to the NRC. These requirements would be equivalent
to those in Sec. Sec. 50.9 and 52.6.
Section 53.080 would govern exemptions from the requirements of the
regulations in part 53. These requirements would be equivalent to those
in Sec. Sec. 50.12 and 52.7.
Paragraphs (a) through (d) of Sec. 50.90 would establish
requirements for standards that the NRC would consider in determining
whether a construction permit (CP), operating license (OL), early site
permit (ESP), combined license, or manufacturing license (ML) under
part 53 would be issued to an applicant. These requirements would be
equivalent to those in Sec. Sec. 50.40, 50.42, 50.43 and 50.22,
respectively. Requirements equivalent to those in Sec. Sec. 50.41 and
50.21 would not be included in part 53 because they apply to Class 104
licenses, and part 53 would not apply to those licenses.
Section 53.100 would require that no license issued under part 53
would cover activities which are not under or within the jurisdiction
of the United States. These requirements would be equivalent to those
in Sec. 50.53.
Section 53.110 would state that licensees and applicants would not
be required to provide design features or other measures for the
specific purpose of protection against the effects of attacks and
destructive acts by enemies of the United States directed against the
facility or deployment of weapons incident to U.S. defense activities.
These requirements would be equivalent to those in Sec. 50.13.
Section 53.115 would establish requirements for rights related to
SNM. These requirements would be equivalent to those in Sec. 50.54(b)
and (c).
Section 53.117 would establish requirements for license suspension
and rights of recapture of the material or control of the facility in a
state of war or national emergency declared by Congress. These
requirements would be equivalent to those in Sec. 50.54(d).
Section 53.120 would establish requirements for information
collection requirements and OMB approval. These requirements would be
equivalent to those in Sec. 50.8.
Subpart B--Technology-Inclusive Safety Requirements
Proposed subpart B, ``Technology-Inclusive Safety Requirements,''
would provide technology-inclusive safety criteria that would serve as
performance standards for the subsequent performance-based requirements
used throughout part 53. Subsequent subparts would define how specific
activities during various stages of the life cycle of a commercial
nuclear plant contribute to satisfying these high-level performance
standards. The performance standards in subpart B would also establish
a means to determine appropriate regulatory controls for SSCs, human
actions, and programs in the following subparts. For example, the
classification of SR SSCs would be built upon the proposed safety
criteria in Sec. 53.210, ``Safety criteria for design-basis
accidents.'' The more detailed requirements for those SSCs would then
be further defined in the design and analysis requirements in subpart
C, ``Design and Analysis Requirements.'' The activities for
manufacturing, constructing, and maintaining the SR SSCs would be
governed by subpart E, ``Construction and Manufacturing Requirements,''
and subpart F, ``Requirements for Operation.''
Requirements for NSRSS SSCs warranting special treatment would
likewise be determined under Sec. 53.220, ``Safety criteria for
licensing-basis events other than design-basis accidents,'' in subpart
B and Sec. 53.460, ``Safety categorization and special treatment,'' in
subpart C. Regulatory requirements related to the NSRSS SSCs would be
distinguished from the regulatory requirements for SR SSCs throughout
part 53. Part 53 would afford more flexibility to applicants and
licensees regarding how NSRSS SSCs would be used in the design and
maintained during plant operations, as compared to SR SSCs.
The collective set of performance-based requirements in part 53
would be sufficient, if met, for the NRC to make the findings required
to grant an application for a utilization facility under Section 182 of
the Act that the utilization of SNM will be in accord with the common
defense and security and will provide adequate protection to the health
and safety of the public. This construct would be similar to existing
NRC regulations, which the Commission has said on many occasions do not
specifically define ``adequate protection.'' However, compliance with
NRC regulations may be presumed to assure adequate protection at a
[[Page 86925]]
minimum. The requirements throughout part 53 that support demonstrating
compliance with Sec. 53.220 would be similar to current regulations
that both contribute to assuring adequate protection of public health
and safety and are desirable to promote the common defense and security
or to protect health or to minimize danger to life or property under
Section 161 of the Act.
Consistent with historical practice, Sections 182 and 161 of the
Act are cited as authorizing legislation within this proposed rule.
However, specific language from the Act would not be incorporated into
the safety objectives or safety criteria in part 53. This is because,
again consistent with historical practice, the NRC would not be
defining ``adequate protection'' through the individual safety
requirements in part 53. Rather, part 53 would enable the NRC to make
its required findings under the Act by providing sufficient performance
standards, safety criteria, and related requirements on how applicants
must demonstrate compliance with subpart B and other subparts.
Section 53.210 would provide safety criteria for DBAs that would be
required to be identified under Sec. 53.240 and analyzed under Sec.
53.450(f) in subpart C of part 53. Subsequent sections in part 53 would
require that the SSCs relied upon to demonstrate compliance with the
criteria in Sec. 53.210 be classified as SR. The use of SR SSCs and
the 25 rem reference values for potential radiological consequences
would align with traditional deterministic approaches for LWRs from
Sec. Sec. 50.34, 52.79, and 100.11 for evaluating the effectiveness of
plant design features with respect to postulated reactor accidents. A
footnote similar to that included in Sec. 50.34(a)(1)(ii)(D)(1) and
Sec. 52.79(a)(1)(vi)(A) would be included in Sec. 53.210 to explain
that the use of the 25 rem value would not be intended to imply that
this number constitutes an acceptable limit for an emergency dose to
the public under accident conditions. Rather, this dose value has been
set forth in this proposed section as a reference value that would be
used in the evaluation of plant design features with respect to DBAs to
verify that the proposed designs would provide assurance of low risk of
public exposure to radiation in the event of an accident. The inclusion
of the safety criteria for DBAs in subpart B would provide a logical
structure supporting the identification and treatment of SR SSCs and
establishing the corresponding functional design criteria for those
SSCs.
Section 53.220 would provide safety criteria for LBEs other than
DBAs that would be required to be identified under Sec. 53.240 and
analyzed under Sec. 53.450(e) in subpart C. Whereas Sec. 53.210 and
the related requirements for SR SSCs would provide that a defined
success path exists for DBAs, the safety criteria for LBEs other than
DBAs would establish the connections between SSC design, human actions,
and programmatic controls and a broader set of potential internal and
external hazards. These safety criteria would also address defense-in-
depth matters such as a balanced consideration of prevention and
mitigation.
The safety criterion in Sec. 53.220(b) would include a requirement
to use a comprehensive risk metric or set of metrics and associated
risk performance objectives against which calculated values of the risk
metrics are compared. The comprehensive risk metrics or set of metrics
and associated risk performance objectives would support a performance-
based approach to developing an appropriate combination of design
features and programmatic controls to prevent or mitigate LBEs other
than DBAs. The applicant must propose the comprehensive risk metric or
set of metrics and associated risk performance objectives, and the
comprehensive risk metric or set of metrics and associated risk
performance objectives must provide an appropriate level of safety.
Comprehensive risk metrics should consist of a proposed plant risk
metric or set of proposed risk metrics that approximate the total,
overall risk from the facility and that address the range of possible
plant configurations and associated internal and external hazards to
the extent practicable. The associated risk performance objectives are
preestablished, indicative values of the comprehensive risk metrics
that are used as part of risk-informed decision-making. The methodology
for developing and using proposed comprehensive risk metrics and
associated risk performance objectives is defined by the proposed
requirements for analyses in Sec. 53.450. Therefore, the application
must include a description of that methodology and, among other things,
should explain the initial conditions, boundary conditions, and key
assumptions used to develop and calculate the risk metrics. Screening
tools and bounding or simplified methods may be used for any mode or
hazard, provided that the applicant provides an acceptable technical
basis. As with all risk-informed methodologies, treatment of
uncertainties must be addressed.
The risk performance objectives established under this methodology
are likely to involve assessing and averaging the risks over a period
of time (e.g., plant year) and would not constitute a real-time
requirement that must be continuously demonstrated by the licensee. The
use of a comprehensive risk metric or set of risk metrics and risk
performance objectives that reflect an average risk to establish
performance goals for SR and NSRSS SSCs is consistent with current
practices that use other risk assessment techniques to address short-
term plant configurations during plant maintenance activities.
It is worth noting that the evaluation of plant risks, as
represented by a comparison of analysis results to acceptable risk
performance objectives for comprehensive risk metrics, would be one of
several performance standards used in subpart B. The proposed use of
multiple performance standards, including deterministic criteria and
defense-in-depth measures, reflects an integrated decision-making
process similar to that described in RG 1.174, ``An Approach for Using
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis,'' Revision 3. The NRC's
approval of using a comprehensive risk metric or set of metrics with
associated risk performance objectives is not, by itself, an indicator
of adequate protection. Rather, the comparison of comprehensive risk
metrics to associated risk performance objectives that are acceptable
to the NRC is part of a suite of regulatory requirements that, when
considered holistically, form the basis for the NRC's decision-making.
This is analogous to the approach used for plants licensed under part
50 and part 52, where no single regulatory requirement governs whether
a plant is ``safe enough.''
The RG 1.233, ``Guidance for a Technology-Inclusive, Risk-Informed,
and Performance-Based Methodology to Inform the Licensing Basis and
Content of Applications for Licenses, Certifications, and Approvals for
Non-Light-Water Reactors,'' describes an example of an acceptable
approach for identifying and analyzing LBEs under part 50 and part 52,
including the use of the quantitative health objectives (QHOs) stated
in the NRC's policy statement, ``Safety Goals for Nuclear Power Plant
Operation,'' dated August 4, 1986 (51 FR 28044), as corrected and
republished August 21, 1986 (51 FR 30028) (Safety Goals Policy
Statement), as acceptable performance objectives for
[[Page 86926]]
comprehensive risk metrics. The use of comprehensive risk metrics, such
as the individual early fatality risk (IEFR) and the individual latent
cancer fatality risk (ILCFR), and associated risk performance
objectives, such as the QHOs, from the Safety Goals Policy Statement,
could form the basis for one approach to meet Sec. 53.220(b). The
requirement for comprehensive risk metrics, in combination with the
other proposed requirements in subparts B and C, would bring the
approach endorsed in RG 1.233 for parts 50 and 52 into part 53.
Additionally, the use of comprehensive risk metrics and associated risk
performance objectives would provide a logical performance objective to
support the risk management approaches in the various subparts
comprising proposed part 53.
The Commission stated in the introduction of the Safety Goals
Policy Statement that improvements to then-current regulatory practices
could lead to a more coherent and consistent regulation of nuclear
power plants, a more predictable regulatory process, a better public
understanding of the regulatory criteria that the NRC applies, and
public confidence in the safety of operating plants. Accordingly, the
Commission announced the safety goals with a focus on the risks to the
public from nuclear power plant operation. Following the issuance of
the Safety Goals Policy Statement, the NRC has used the comprehensive
risk metrics and performance objectives provided in the safety goals
within the criteria for many decisions involving safety judgments
during the licensing and regulation of operating reactors and proposed
nuclear reactor designs. Consistent with NUREG-0880, the proposed
comprehensive risk metrics and associated risk performance objectives
required under Sec. 53.220(b) could be expressed in terms of a
biologically average individual in terms of age and other risk factors.
Although some comprehensive risk objectives such as the IEFR and ILCFR
are defined in terms of fatality risks, the Commission continues to
make clear that no death attributable to nuclear power plant operation
will ever be ``acceptable'' in the sense that the Commission would
regard it as a routine or permissible event. Comprehensive risk metrics
and associated risk performance objectives as used in this proposed
rule would establish acceptable risks, not acceptable deaths.
Applicants under the proposed part 53 may choose to develop and
seek NRC approval of comprehensive risk metrics or sets of risk metrics
and associated risk performance objectives beyond those discussed
above, including the use of surrogate measures for use in specific
analyses to satisfy the proposed requirements in Sec. 53.220(b). Such
surrogate measures for comprehensive risk metrics and associated risk
performance objectives could be used in a manner similar to the use of
core damage frequency and conditional containment failure probability
for LWRs within the safety goal evaluation process in NUREG/BR-0058,
``Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory
Commission,'' and other assessments of LWRs using the NRC's safety
goals. The NRC would, as appropriate, review novel approaches for
comprehensive metrics and associated risk performance goals proposed by
applicants, industry organizations, or standard development
organizations and would engage stakeholders during the development of
the related regulatory guidance or specific licensing actions.
Section 53.230 would require safety functions needed to ensure that
the safety criteria under Sec. Sec. 53.210 and 53.220 can be met if an
assumed LBE were to occur at a commercial nuclear plant. Section 53.230
would specify that limiting the release of radioactive materials from
the facility is the primary safety function, and therefore, limiting
potential offsite consequences (i.e., dose to a hypothetical
individual) would be used as the primary performance metric throughout
part 53. The additional or subsidiary safety functions needed to limit
the release of radionuclides may include, without limitation,
controlling processes related to reactivity, heat generation, heat
removal, and chemical interactions. This proposed rule provides
flexibility to applicants and licensees in identifying, implementing,
and maintaining the safety functions supporting retention of
radionuclides for commercial nuclear plants of varying sizes and
technologies.
Proposed Sec. 53.240 would require applicants to identify and
address LBEs. LBEs are unplanned events, resulting from both internal
and external hazards, that are used in the design and analyses required
under part 53 for licensing commercial nuclear plants. This ensures
estimates of offsite consequences from analyses performed under
proposed Sec. 53.450 are below the safety criteria identified under
proposed Sec. Sec. 53.210 and 53.220 and that SSCs, personnel, and
programs address the safety functions from proposed Sec. 53.230.
Including a high-level performance requirement related to the
identification and analysis of LBEs in subpart B would reflect the
historical and continuing importance of evaluating unplanned events as
part of the licensing of commercial nuclear plants. Proposed Sec.
53.240 would require identification and analysis of LBEs under Sec.
53.450, which would require a PRA. Examples of acceptable methods of
using PRAs to identify and assess LBEs would be the methodology in RG
1.233, as discussed in Draft Regulatory Guide (DG)-1413, ``Technology-
Inclusive Identification of Licensing Events for Commercial Nuclear
Plants.''
Section 53.250 would establish defense-in-depth requirements based
on the longstanding philosophy of providing defense in depth to address
uncertainties about the design, operation, and performance of
commercial nuclear plants. For example, parts 50 and 52 address defense
in depth through layered prescriptive technical requirements (e.g.,
fuel performance, cladding integrity, reactor coolant system integrity,
containment performance) for LWRs. In contrast, the flexibility
afforded to applicants in how they propose to demonstrate compliance
with the high-level safety criteria within part 53 would necessitate
this specific requirement to ensure defense in depth is provided. The
requirements in this section would state that no single engineered
design feature, human action, or programmatic control, no matter how
robust, should be exclusively relied upon to address LBEs other than
DBAs. The phrase ``engineered design feature'' would not preclude the
possible crediting of inherent characteristics within the design and
analysis for commercial nuclear reactors. While defense in depth would
only be assessed for LBEs other than DBAs, the need to ensure dedicated
success paths for DBAs would contribute to the overall defense in depth
for each commercial nuclear plant under part 53.
Section 53.260 would govern normal operations and would establish a
level of safety based on current requirements in 10 CFR part 20,
``Standards for Protection Against Radiation,'' which limits doses to
members of the public and dose rates in unrestricted areas.
Section 53.270 would provide for the protection of plant workers
and would establish a level of safety based on current requirements in
10 CFR part 20 which limits occupational dose.
Subpart C--Design and Analysis Requirements
This subpart would provide requirements for the design of
commercial nuclear plants and the supporting analyses, including the
analyses of LBEs, to demonstrate that the performance standards in
proposed
[[Page 86927]]
subpart B can be satisfied. The sections within subpart C would reflect
the overall hierarchy throughout part 53, which would cover: (1) plant-
level safety criteria (Sec. Sec. 53.210, 53.220, and 53.470); (2)
safety functions (Sec. 53.230) needed to demonstrate compliance with
the safety criteria; (3) design features (Sec. 53.400), human actions,
and programmatic controls needed to fulfill the safety functions; and
(4) functional design criteria (Sec. Sec. 53.410 and 53.420) that must
be defined for each design feature relied on to demonstrate the safety
criteria (Sec. Sec. 53.210, 53.220, and 53.470) are met. Subpart C
would also contribute to the logic and structure of part 53 by
distinguishing between SR SSCs and NSRSS SSCs and licensee-controlled
programs that address LBEs other than DBAs. Specifically, SR SSCs,
human actions, and programmatic controls needed to protect against DBAs
are used to satisfy the safety criteria in Sec. 53.210. Non-safety-
related but safety-significant SSCs, human actions, and licensee-
controlled programs that address LBEs other than DBAs generally
contribute to the appropriate measures considering potential risks to
public health and safety.
Section 53.400 would establish a requirement that design features
be provided for each commercial nuclear plant to satisfy the safety
criteria and fulfill safety functions from proposed subpart B during
LBEs. Other sections in subpart C would, in turn, further address the
necessary capabilities and reliabilities for SSCs by establishing
functional design criteria, fulfilling design requirements, performing
analyses of LBEs, performing other supporting analyses, and
categorizing SSCs based on their roles in preventing or mitigating
LBEs.
Section 53.410 would require that functional design criteria be
defined for design features relied upon to demonstrate that the
consequences from DBAs would be below the criteria in Sec. 53.210
through analyses performed under Sec. 53.450(f), which includes
insights from both PRAs and deterministic analyses. Other sections
within part 53 would establish appropriate controls on these design
features (e.g., safety classification, protection from external
hazards, quality assurance, and TS) to ensure the functional design
criteria are satisfied. The performance requirements for the SSCs
needed to address DBAs and the corresponding human actions and
programmatic controls would contribute to ensuring that a commercial
nuclear plant licensed under part 53 would meet the safety criteria in
Sec. 53.210.
Section 53.415 would require that SR SSCs be protected against or
designed to withstand the effects of natural phenomena (e.g.,
earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches) and
constructed hazards (e.g., from dams, transportation routes, and
military or industrial facilities). Specifically, Sec. 53.415 would
require that SR SSCs remain capable of performing the safety functions
stated in Sec. 53.230 for which they are credited up to the design-
basis external hazard levels as determined under Sec. 53.510. As used
in Sec. 53.415 and subpart D of part 53, a hazard level would refer to
such things as the magnitude and recurrence rate of an earthquake and
the resultant ground motions, the height of a flood, the force of
hurricane winds, or the concentrations of chemicals resulting from a
release from a nearby facility. These requirements would support either
traditional deterministic approaches for determining and protecting
against external hazards or probabilistic approaches that are being
developed for seismic and some other external hazards.
Section 53.420 would require that functional design criteria be
defined for design features that play a significant role in
demonstrating that the safety criteria for LBEs other than DBAs are
satisfied. The analyses required for this demonstration would be
described in proposed Sec. 53.450(e), which would require that those
events be identified and assessed using a PRA methodology in
combination with other generally accepted approaches for systematically
evaluating engineered systems. The SSCs determined to be safety
significant (i.e., either SR or NSRSS) would have associated special
treatment requirements as specified in Sec. 53.460. Special treatment
would be defined in subpart A of part 53 and generally refers to
measures (e.g., quality assurance, testing, monitoring) taken beyond
the procurement and installation of commercial grade products to
provide confidence that the SSC will comply with the applicable
functional design criteria. The inclusion of a systematic approach to
identifying the functional design criteria for SSCs and tailoring the
special treatments to specific LBEs and safety functions is an
important contributor to satisfy the proposed safety criteria in
subpart B. Therefore, designers and licensees for commercial nuclear
plants would be provided flexibility on how LBEs other than DBAs are
either prevented or mitigated and how the calculated comprehensive
plant risks satisfy the safety criterion established under Sec.
53.220(b).
Section 53.425 would establish requirements for design features and
related functional design criteria limiting doses to members of the
public during normal operations to satisfy the criteria in part 20.
Section 53.430 would provide similar requirements for design features
and related functional design criteria for protection of plant workers
to meet the safety criteria in part 20. Similar to existing
regulations, the NRC considers that licensees would generally comply
with the requirements of part 20 to keep doses as low as reasonably
achievable by meeting a design objective of keeping doses to the public
from routine plant effluents less than 10 millirem per year. This goal
is similar to that provided by appendix I to part 50 and would assist
designers, applicants, and licensees in performing the evaluations of
possible reductions in public dose from routine effluents when
considering costs and other factors. As emphasized in existing
regulations in part 50, the design objective of keeping doses to the
public from routine plant effluents less than 10 millirem per year
should not be construed as a radiation protection standard. The NRC
anticipates that future guidance will continue to reflect this
performance goal.
The proposed requirements in Sec. Sec. 53.425 and 53.430 for
design features and functional design criteria to support radiation
protection activities have parallels in existing regulations such as
Sec. 50.34(a) and (b)(3), which require in part that the means be
provided for meeting the requirements of part 20 and General Design
Criterion 60, 61, 63, and 64 in appendix A to part 50, which provide
radiation protection related design criteria.
Section 53.440 would address various design requirements that
warrant specific mention to ensure that the design features required by
Sec. 53.400 comply with the functional design criteria required by
Sec. Sec. 53.410 and 53.420. These requirements would be met through
design practices, consideration of testing and operating experience,
and various assessments of LBEs and other potential challenges to
commercial nuclear plants. Discussions of some of the key design
requirements included in this section follow.
Sec. 53.440(a): An essential element to ensuring a
proposed design can comply with the performance criteria in proposed
part 53 would be that the abilities of design features to fulfill their
safety functions are demonstrated by a combination of analyses, test
programs, prototype testing, and operating experience. This requirement
closely aligns with the language in Sec. 50.43(e)
[[Page 86928]]
and is proposed in part 53 as the same foundational requirement. In
addition, the proposed Sec. 53.440(a) would require the design
processes for SSCs under this section to include administrative
procedures for evaluating operating, design, and construction
experience for considering applicable important industry experiences in
the design of those SSCs. This proposed requirement corresponds to the
existing requirement under Sec. 50.34(f)(3)(i) that was developed in
response to the 1979 accident at Three Mile Island Nuclear Generating
Station.
Sec. 53.440(b): The design and licensing of commercial
nuclear plants should use generally accepted consensus codes and
standards. Such codes and standards ensure sufficient testing and
qualification of materials and equipment and provide defined processes,
specifications, and acceptance criteria for use by designers and
suppliers. The NRC would indicate acceptance of consensus codes and
standards used in the design and licensing of a specific commercial
nuclear plant either through the NRC's generic endorsement of a code or
standard (i.e., through regulatory guidance), including any limitations
or conditions, that can be referenced within an application, or through
the review of a referenced code or standard as part of the review of a
specific application.
Sec. 53.440(c): The design requirements in subpart C
would require the materials used for SR and NSRSS SSCs to be qualified
for their service conditions over the design life of the SSC.
Sec. 53.440(d): The requirements in Sec. 53.440 would
include the need to consider possible degradation mechanisms for
materials and equipment to inform both the design process and the
development of integrity assessment programs to be executed during
plant operations in accordance with subpart F of part 53. The inclusion
of requirements related to designing and monitoring for possible
degradation mechanisms reflects important lessons learned from the
history of LWRs as well as operating experience with structures and
systems in countless other engineering endeavors.
Sec. 53.440(e) and (f): The design requirements in
subpart C would state specific design requirements similar to existing
requirements in parts 50, 52, and 73 for protections against fires and
explosions and consideration of safety and security together in the
design process.
Sec. 53.440(g) and (h): Specific design requirements are
proposed to ensure that commercial nuclear reactors under part 53 have
the capability to achieve and maintain subcriticality and long-term
cooling. The requirements would be included to address the potential
that some reactor designs may be able to achieve a stable end state for
the purpose of event analyses but might need further actions to
completely shut down and service the facility.
Sec. 53.440(i): The design, analysis, and development of
programmatic controls under part 53 would consider the number of
reactor units and other significant inventories of radioactive
materials contributing to the risks to public health and safety. This
would reflect the definition of ``Commercial nuclear plant'' in subpart
A and reinforce that the evaluation of LBEs is performed on a plant-
wide basis. This aspect of part 53 would be different from parts 50 and
52, which generally define safety requirements on the assumption of
events involving only individual reactor units.
Sec. 53.440(j): A design requirement is proposed to
provide a technology-inclusive requirement that would be equivalent to
the requirements in Sec. 50.150 to address the possible impact of a
large commercial aircraft.
Sec. 53.440(k): The inclusion of a specific proposed
requirement to address the risks to public health from potential
chemical hazards of licensed material is appropriate given the
diversity of reactor technologies and designs that might be licensed
under part 53. The requirement in part 53 would be similar to the
existing requirements in 10 CFR part 70, ``Domestic Licensing of
Special Nuclear Material,'' that address both potential radiological
and chemical hazards for licensed materials at fuel cycle facilities.
Sec. 53.440(l): Provisions are proposed to require that
measures be taken during the design of commercial nuclear plants to
minimize contamination of the facility and the environment, facilitate
eventual decommissioning, and minimize the generation of radioactive
waste in accordance with Sec. 20.1406.
Sec. 53.440(m): A design requirement is proposed to
provide a technology-inclusive equivalent to the requirements in Sec.
50.68 by including options for commercial nuclear plants to either have
a monitoring system capable of detecting a criticality as described in
Sec. 70.24 or to have restrictions on SNM handling and storage that
would prevent inadvertent criticality events.
Sec. 53.440(n): The design would need to reflect state-
of-the-art human factors principles for safe and reliable performance
in all settings that human activities are expected for performing or
supporting the continued availability of plant safety or emergency
response functions.
Section 53.450 would establish analysis requirements and would
center upon the use of a PRA in combination with other generally
accepted approaches for systematically evaluating engineered systems.
The reliance on PRAs as a key component in the proposed analysis
requirements for part 53 would reflect the decades of improvements in
PRA methodologies and the increasing use of PRA techniques in the
design, licensing, and oversight of both operating and future nuclear
reactors. Part of the Commission's PRA Policy Statement is that the use
of PRA technology should be increased in all regulatory matters to the
extent supported by the state of the art in PRA methods and data and in
a manner that complements the NRC's deterministic approach and supports
the NRC's traditional defense-in-depth philosophy. The need to
supplement PRA insights with other engineering approaches and judgments
reflects the NRC's longstanding policy described in the SRM to SECY-98-
144, ``Staff Requirements--SECY-98-144--White Paper on Risk-Informed
and Performance-Based Regulations,'' dated February 24, 1999, for
regulatory decision-making to be risk-informed but not solely based on
numerical results of a risk assessment (i.e., not a risk-based
approach). Part 53 would maintain a role for NRC's traditional
deterministic approaches (particularly for DBAs) and defense-in-depth
philosophy by including specific requirements utilizing these
regulatory tools in subparts B and C.
PRA would be used in combination with other techniques in part 53
to identify and categorize LBEs, classify SSCs, and evaluate defense in
depth. This increased role for the PRA necessitates that it would be
developed, performed, and maintained in accordance with NRC-approved
standards and practices (see Sec. 53.450(c) and (d)). The computer
codes used to model the plant response and the behavior of the barriers
to the release of radionuclides would need to be qualified for the
range of conditions being simulated across a wide range of unplanned
events. These analyses would need to use realistic approaches and
address uncertainties associated with states of knowledge, modeling,
and performance of SSCs.
While industry consensus PRA standards and peer review processes
endorsed in RGs 1.200 and 1.247 remain
[[Page 86929]]
acceptable for developing a PRA, they are not regulatory requirements
and an application under part 53 need not follow every aspect of the
applicable consensus PRA standard. Existing processes for defining the
scope and capability of a PRA supporting an application offer
flexibility in determining the degree to which the PRA needs to be
developed and may be informed by other factors such as design
complexity and the needed degree of realism and level of detail,
consistent with the use of the PRA and substance of the application.
Such processes are currently available for appropriately defining the
scope of the PRA and determining applicability of supporting
requirements in consensus PRA standards needed to satisfy the proposed
regulatory requirements for the specific uses of analyses under Sec.
53.450(b). Likewise, NRC determinations of the acceptability of such
PRAs would include consideration of the appropriateness of the
applicant-defined scope as part of determining the applicability of and
conformance to consensus PRA standard supporting requirements
consistent with the current state of practice. In addition, these
determinations would include consideration of other aspects of the
development of the PRA, such as PRA peer reviews. An NRC determination
of the acceptability of a PRA includes but is not limited to assessing
the initial and boundary conditions and key assumptions used in the
analysis, treatment of uncertainties, and the use of screening tools
and bounding or simplified methods for any mode or hazard, provided the
use of those tools and methods is justified by an acceptable technical
basis. In that regard, the consensus PRA standards would not be applied
by the NRC as a strict checklist of requirements for part 53 PRA
acceptability determinations.
The proposed Sec. 53.450(c) would require periodic maintenance and
upgrading of the PRA to maintain an alignment between the supporting
analyses and the design and performance of plant equipment, programs
and procedures, and other factors associated with meeting the safety
criteria of the proposed Sec. 53.220 and the evaluation criteria of
proposed Sec. 53.450(e)(2). The periodic maintenance of the PRA would
also be a means to consider new or revised information related to
external hazards, industry operating experience, performance issues
with or degradation of SSCs, and other contributors to the frequency
and potential consequences of various event sequences. The periodic
assessments performed by licensees to support the maintenance of the
PRA and other requirements in the proposed part 53 would be
complemented by NRC inspections and programs to assess new or revised
information related to topics such as natural hazards, operating
experience, and potential generic safety issues.
The categories of LBEs used in part 53 would include anticipated
event sequences, unlikely event sequences, and very unlikely event
sequences. The unlikely event sequences would include those events with
estimated frequencies well below the frequency of events expected to
occur during the lifetime of a commercial nuclear plant. An important
aspect of the analysis requirements is that, under proposed Sec.
53.450(e), the analyses of LBEs other than DBAs would not only be used
to show the performance criteria of Sec. 53.220 are satisfied but to
also show that evaluation criteria defined for each LBE or category of
LBEs would also be satisfied. Such evaluation criteria for specific
LBEs or categories of LBEs would be defined in terms of limits on the
release of radionuclides or maintaining the integrity of one or more
barriers used to limit the release of radionuclides and reflect a
graded approach of allowing lesser potential consequences from more
frequent events. An example of such evaluation criteria for a range of
LBEs that could likely be expanded for part 53 is provided in RG 1.233.
Another proposed requirement for the proposed Sec. 53.450(e) analyses
is that the methodology would need to include a means to identify event
sequences deemed risk-significant such that those event sequences can
be given special attention within other sections of part 53.
Part 53 would maintain an important role for a deterministic
analysis of DBAs in the performance criteria of Sec. 53.210 and the
related analytical requirements in Sec. 53.450(f). The analysis of
DBAs would be required to address event sequences drawn from those with
estimated frequencies below the expected lifetime of a generation of
reactors (e.g., event sequences with frequencies as low as one in ten
thousand years). As proposed in this section, DBAs would need to be
analyzed using deterministic methods and ensure a safe, stable end
state with reliance upon only SR SSCs and human actions, if needed, to
be performed by operators licensed under the provisions of Sec. Sec.
53.760 through 53.795.
While the DBAs analyzed under part 53 would be similar to the
traditional DBAs analyzed under parts 50 and 52, there are important
distinctions between the overall role of DBA analyses in part 50 and
proposed part 53. In part 53, the role of the DBA analysis would be
more narrowly focused on selecting SR SSCs and determining functional
design criteria for those SSCs to ensure the commercial nuclear plant
meets the safety criteria in Sec. 53.210. The overall control of risks
posed by commercial nuclear plants under part 53 would be provided by
the analyses of and measures taken for both DBAs and other LBEs,
including very unlikely event sequences. This would contrast with the
traditional deterministic approach in part 50 wherein the analyses of
DBEs such as DBAs were used to provide bounding assessments,
incorporate standard design rules such as assumptions related to single
failures, and to define conservative performance requirements for SR
SSCs. Limitations related to the traditional deterministic approach
were addressed in part 50 through case-by-case assessments and specific
actions for beyond-design-basis events such as anticipated transients
without scram and station blackout.
Section 53.450 would also include provisions to ensure that
analyses are performed to support the design requirements of Sec.
53.440(e) on fire protection, Sec. 53.440(j) on aircraft impact
assessments, and Sec. 53.425 on using design features and plant
programs to control doses to members of the public from routine
effluents and direct radiation from contained sources. The proposed
analysis requirements related to fire protection would support either a
traditional, deterministic approach or a more risk-informed approach
where the risks from fires are addressed within the identification and
analyses of LBEs.
Section 53.460 would establish criteria for the safety
classification of SSCs and determination of appropriate special
treatments. As noted in subpart A, the term ``Special treatments''
would be defined to mean those items, such as measures taken to satisfy
functional design criteria, quality assurance, and programmatic
controls, which provide assurance that certain SSCs will provide
defense in depth or perform risk-significant functions. These
requirements would also provide confidence that the SSCs will perform
under the service conditions and with the reliability credited in the
analysis performed in accordance with Sec. 53.450 to satisfy the
safety criteria in Sec. Sec. 53.210 and 53.220. The terminology used
in part 53 would include the following categories for SSC
classification: (1) SR; (2) NSRSS; and (3) non-safety significant.
Requirements for SR SSCs would be defined in other sections of
[[Page 86930]]
part 53 and would include using TSs for controls during operation and
the application of quality assurance requirements from appendix B of
part 50.
Requirements for NSRSS SSCs would include the need to identify
necessary special treatments such as performance measures on
reliability. Licensees would generally be afforded flexibility in
maintaining and changing special treatments for SSCs categorized as
NSRSS. Non-safety-significant SSCs would be addressed under normal
licensee programs for commercial grade equipment and typical industry
practices for general plant design and maintenance. Safety-related SSCs
would also contribute to defense in depth and risk-significant
functions and may warrant special treatments beyond those defined for
their SR functions to reflect their role in meeting the safety criteria
in Sec. 53.220 and the evaluation criteria in Sec. 53.450(e).
Section 53.470 would allow an applicant or licensee to seek
operational flexibilities by adopting more restrictive criteria than
those provided in Sec. 53.220 and that might otherwise be used in the
analysis of LBEs under Sec. 53.450(e). Such an approach might be taken
to ensure sufficient safety margins to gain operational flexibilities
in areas such as justifying siting in relation to population centers or
staffing levels. As an example, an applicant or licensee could propose
to justify siting proposals by adopting alternate criteria for very
unlikely event sequences. Such alternate criteria could require
calculated consequences for an individual at the exclusion area
boundary to be less than one rem total effective dose equivalent
(TEDE). This section would establish requirements to ensure that, if
more restrictive evaluation criteria than those required by a
methodology were used to justify operational flexibilities, then the
analysis, design features, and programmatic controls would be
established and maintained accordingly.
Section 53.480 would establish seismic design considerations. This
proposed section would relate to the safety criteria in subpart B, the
analytical requirements related to external hazards in Sec. 53.450,
and subpart D, ``Siting Requirements.'' For licenses issued under part
53, this section in subpart C would support a variety of approaches to
seismic design. For example, a design for a commercial nuclear plant
could show that SSCs are able to withstand the effects of earthquakes
by adopting an approach similar to that in appendix S to part 50.
Alternatively, an applicant could follow the more recent risk-informed
alternatives afforded by standards development organizations (e.g.,
American Society of Civil Engineers (ASCE)/Structural Engineering
Institute (SEI) 43-19, ``Seismic Design Criteria for Structures,
Systems, and Components in Nuclear Facilities.'') Because the agency
has not endorsed ASCE/SEI-43-19, an applicant can propose to use ASCE/
SEI 43-19 on an application specific basis to meet Sec. 53.480 and the
NRC would evaluate the adequacy of the standard as applied in that
application. The design could also be done with the full integration of
seismic PRAs into the design and licensing of a particular commercial
nuclear plant. This section has been developed to accommodate a variety
of potential risk-informed, performance-based seismic design
approaches. The analyses required by Sec. 53.450 would need to address
seismic hazards as well as other external hazards. The expected
responses of SSCs to a range of seismic events would be included in the
analyses when ensuring that the safety criteria defined under Sec.
53.220 would be met. The potential SSC responses to seismic hazards
could be addressed in the analyses using a fragility model (conditional
probability of its failure at a given hazard input level), a high
confidence of low probability of failure value, or other method
endorsed or otherwise found acceptable by the NRC.
Subpart D--Siting Requirements
Proposed subpart D in part 53 would state requirements for the
siting of commercial nuclear plants and would serve the role provided
by 10 CFR part 100, ``Reactor Site Criteria,'' for nuclear reactors
licensed under parts 50 and 52. As reflected in proposed Sec. 53.500,
the reason for establishing siting requirements would remain the same
as it has been historically, which is to ensure that licensees and
applicants assess what impact the site environs may have on a
commercial nuclear plant (e.g., external hazards) and, conversely, what
potential adverse health and safety impacts a commercial nuclear plant
may have on nearby populations in view of the site characteristics.
Proposed Sec. 53.510 would require that design-basis external
hazard levels be identified and characterized based on site-specific
assessments of natural and constructed hazards with the potential to
adversely affect plant functions. The site-specific assessments would
be used in the proposed Sec. 53.415, which would require that SR SSCs
be designed to withstand the effects of natural phenomena and
constructed hazards of levels or severities up to design-basis external
hazard levels. The design-basis levels for external hazards relevant to
a site would need to account for uncertainties and variabilities in
data, models, and methods used to characterize those hazards. Existing
approaches could be used to demonstrate compliance with this
requirement. The historical importance of assessing seismic events as
risks to commercial nuclear plants and the associated development of
risk-informed approaches to address seismic events would be reflected
in proposed Sec. 53.480, ``Earthquake engineering,'' and specific
requirements in subpart C. The NRC is developing a graded approach for
seismic design by grouping SSCs into different seismic design
categories (SDCs) based on their risk significance. While the agency
has not endorsed ASCE/SEI-43-19, an applicant can propose to use ASCE/
SEI 43-19 on an application-specific basis to meet Sec. 53.480 and the
NRC will evaluate the adequacy of the standard as applied in that
application. The NRC staff will continue to review ASCE/SEI-43-19 as
part of its efforts to further develop guidance in this area. The
approach described in RG 1.208, ``A Performance-Based Approach to
Define the Site-Specific Earthquake Ground Motion,'' would be an
acceptable way to develop site-specific ground motion response spectra
for SSCs under appendix S to part 50, which corresponds to SSCs that
are categorized as the highest SDC (SDC-5) in ASCE/SEI 43-19.
The evaluation of seismic hazards under subpart D would need to be
sufficient to inform a site-specific design (e.g., a CP or custom COL)
or confirm the use of a standard design for a commercial nuclear plant
under Sec. 53.480 and other sections of subpart C. A risk-informed
approach could use several design-basis ground motions (DBGMs) to
assess SSCs in various SDCs (i.e., one DBGM per SDC). Section 53.510(d)
would state that geologic and seismic siting factors must also include
related hazards such as seismically induced flooding and volcanic
activity that may affect the design and operation of a proposed
commercial nuclear plant for the proposed site.
Section 53.520 would require applicants to identify and assess site
characteristics related to topics which might include meteorology,
geology, hydrology, or other areas in the design and analyses required
under subpart C.
Proposed section 53.530 would set requirements for population-
related considerations and maintain requirements and definitions
similar to those currently in part 100 for an exclusion area, low
population zone,
[[Page 86931]]
and population center distance. The NRC recognizes that some applicants
may propose to essentially collapse the exclusion area and low
population zone to the site boundary. This approach would rest on a
demonstration that the calculated consequences of DBAs remain below the
proposed dose guidelines used in Sec. 53.210, which are the same as
those in the existing regulations in parts 50, 52, and 100. The
proposed definitions in Sec. 53.020 would allow such configurations,
assuming they were justified by the design and analyses from subpart C.
This approach should provide flexibility to justify alternative
exclusion areas and low population zones without foreclosing the option
for an applicant to define more conventional exclusion areas and low
population zones outside of a defined site boundary. The NRC's long-
standing preference for siting reactors in areas of low population
density would be maintained in part 53 by using the current language
from part 100 in proposed Sec. 53.530(c). The NRC revised guidance
related to population densities surrounding a commercial nuclear plant
in Revision 4 to RG 4.7, ``General Site Suitability Criteria for
Nuclear Power Stations'' to reflect Commission direction in SRM-SECY-
20-0045, ``Population Related Siting Considerations for Advanced
Reactors.'' Site-related requirements in part 20 (restricted area) and
part 73 (protected and owner-controlled areas) would remain applicable
to commercial nuclear plants licensed under part 53.
Proposed section 53.540 would require that site characteristics be
appropriately considered in other activities such as the design and
analysis performed under proposed subpart D and the emergency planning
and security programs under proposed subpart F.
Subpart E--Construction and Manufacturing Requirements
The proposed part 53 language would establish construction and
manufacturing requirements in subpart E. The proposed language for
construction-related activities would largely reflect current
requirements in part 50 without any fundamental changes. Limited
changes would be made in several places, as described in the following
paragraphs, to be technology-neutral and for consistency with the
organization and language of part 53. The proposed language for
requirements for manufacturing activities would largely mirror those
for construction-related activities. However, the proposed
manufacturing requirements have been updated from the current
requirements in subpart F of part 52 to better accommodate the possible
factory fabrication of manufactured reactors. The manufacturing of
specific components outside the scope of an ML would not be addressed
by these proposed subparts.
Section 53.600 would establish the overall construction and
manufacturing requirements for CPs, OLs, COLs, MLs, and limited work
authorizations (LWAs). This section would connect the construction and
manufacturing requirements to the safety criteria, quality assurance
requirements, and other requirements located in other subparts. These
requirements would require that construction and manufacturing
activities be managed and conducted such that when combined with
associated design features and programmatic controls, the constructed
plant would satisfy the relevant requirements in subpart B.
Section 53.605 would establish requirements for the reporting of
defects and instances of noncompliance during construction. This
section would provide equivalent requirements to those in Sec.
50.55(e).
Section 53.610(a) would establish the requirement to have in place
a well-defined command and control structure to manage construction
activities. The requirements would generally reflect current
requirements, with an emphasis on the quality assurance programs for
complying with the requirements in appendix B to part 50. The proposed
Sec. 53.610(a)(6) would require programmatic controls for implementing
special treatment for NSRSS SSCs to align with requirements in other
subparts in part 53. The section would also refer to other NRC
regulations to address matters such as requirements to have a FFD
program, a radiation protection program if radioactive materials are
brought onto the site, and security programs to protect sensitive
information and protect against cyber threats.
Section 53.610(b) would provide requirements governing construction
activities, including the equivalent of the requirement in Sec.
50.10(e) that prohibits starting construction until the NRC has
authorized the activities by issuing a CP, COL, ESP, or LWA. Section
53.610(b)(1)(iii) would require procedures to be in place prior to
beginning construction to ensure that construction-related activities
do not undermine important features such as slope stability and that
construction-related activities such as backfilling of excavated
portions of the site appropriately address potential pre-construction
activities such as the emplacement of retaining walls or drainage
systems. Other requirements in these paragraphs would be equivalent to
requirements in parts 50 and 52 with appropriate references to other
parts for items such as possession of byproduct material or SNM,
protecting operating units from construction activities for commercial
nuclear plants with multiple reactor units, and having a redress plan
in case LWA activities are terminated.
Section 53.610(c) would address inspection and acceptance
activities by including requirements in part 53 equivalent to specific
quality assurance criteria in appendix B to part 50 and inspections,
tests, analyses, and acceptance criteria (ITAAC) in part 52 for COLs.
Section 53.620(a) would include proposed requirements covering the
activities performed under an ML issued under part 53. Provisions
related to MLs were first adopted by the NRC in 1973 through the
addition of appendix M to part 50. The regulation supported the
manufacture of a nuclear power reactor to be incorporated into a
commercial nuclear plant under a CP and operated under an OL at a
different location from the place of manufacture.\1\ The regulations
and processes for MLs were changed substantially in the part 52
rulemaking in 2007 (72 FR 49352). The most important shift in the ML
concept in that rulemaking was that a final reactor design, which would
be equivalent to that required for a standard DC under part 52 or an OL
under part 50, must be submitted and approved before issuance of an ML.
The rationale for that change was that approval of a final design
ensures early consideration and resolution of technical matters before
there is any substantial commitment of resources associated with the
actual manufacture of the reactor, which greatly enhances regulatory
stability and predictability.
---------------------------------------------------------------------------
\1\ On December 17, 1982, the NRC issued ``Manufacturing License
ML-1 to Offshore Power Systems for the manufacture of a maximum of
eight floating nuclear plants,'' dated September 30, 1982, but the
project was subsequently canceled.
---------------------------------------------------------------------------
The proposed part 53 sections in subpart E for manufacturing and in
subpart H for licensing matters would maintain requirements equivalent
to those in part 52 for MLs. The NRC approval of a standard design and
related manufacturing processes, coupled with a stable workforce and
established procedures, has the potential for maintaining and even
improving the quality and consistency of manufacturing, as compared to
the traditional method of constructing
[[Page 86932]]
reactors onsite by a variety of contractors and subcontractors.
Subpart E would include requirements that would apply to portions
of a manufactured reactor in recognition that some activities covered
by an ML may occur at different fabrication facilities. As with the
preceding sections on construction, Sec. 53.620 would establish the
requirements to have in place programs, procedures, and a well-defined
command and control structure to manage manufacturing-related
activities.
Section 53.620(b) in subpart E would propose requirements for
executing the manufacturing activities following receipt of an ML under
part 53. Information about the design and manufacturing processes
should be provided by the applicant. The importance of the ML is
reflected in several of the proposed requirements in Sec. 53.620(b)
that would refer to complying with the ML, including conducting
manufacturing processes within facilities for which the license holder
can control activities. The essential role of post-manufacturing
inspections would also be incorporated into this proposed section by
requiring the holder of the ML to perform inspections and have
acceptance processes for manufactured reactors or portions of a
manufactured reactor.
Section 53.620(c) would provide proposed requirements for the
control of radioactive materials if the holder of an ML plans to
possess and use source, byproduct, or SNM as part of the manufacturing
process. By and large, the proposed subpart E would refer to NRC
regulations in 10 CFR part 30, ``Rules of General Applicability to
Domestic Licensing of Byproduct Material,'' 10 CFR part 40, ``Domestic
Licensing of Source Material,'' and part 70 for the requirements on
controlling radioactive materials. Several specific requirements to
address the potential hazards of radioactive materials are proposed in
areas such as having a fire protection program, an emergency plan,
training programs, and procedures to minimize contamination.
The most significant change proposed for MLs in part 53 as compared
to MLs under part 52 relates to Sec. 53.620(d) in subpart E and the
associated licensing provisions in subpart H. These provisions would
allow and establish requirements for the loading of fuel into a
manufactured reactor at the manufacturing site for subsequent transport
to a commercial nuclear facility that will operate pursuant to a COL.
The first requirement in the proposed Sec. 53.620(d) would establish
limitations on when a license under part 70 would authorize the loading
of fuel into a reactor manufactured under an ML. The proposed
regulation would require the manufactured reactor to include at least
two independent physical mechanisms that will each prevent criticality
should conditions most favorable to critical operation be introduced
(e.g., optimum neutron moderation and reflection). This requirement
would contribute to the NRC's longstanding practice of requiring
defense in depth for preventing accidents in any facility dealing with
SNM, including requirements in Sec. 70.64 for certain part 70
licensees to adhere to the ``double contingency principle.''
The requirements to have in place mechanisms to prevent criticality
could likewise support meeting other provisions in subpart H to part
70, such as those related to having a safety program and integrated
safety assessment. The mechanisms to preclude criticality in the
proposed requirements would reasonably ensure that a manufactured
reactor would not become critical assuming optimum neutron moderation,
and optimum neutron reflection conditions. With the proposed
requirements for mechanisms to prevent criticality and all criticality
safety controls required by 10 CFR part 70 in place, the presence of
fuel in the manufactured reactor would not create a nuclear hazard
different than the hazard from the presence of the same fuel in a
storage location or container licensed under 10 CFR part 70.
Collectively, the proposed measures would reasonably ensure that the
manufactured reactor would not be capable of operations, thereby
obviating the need for a COL under Sec. Sec. 53.1416 and 53.1440 to
authorize fuel loading. Additionally, this approach would focus the ML
application and its review on the design, manufacture, and deployment
of the manufactured reactor.
The activities involving SNM within the manufacturing facility,
including the loading of fuel, would be regulated primarily under the
part 70 license. The reference to the requirements in subpart H of part
70 in section 53.620(d) assures that the activities involving the
receipt, storage, and loading of a variety of possible fuel forms and
enrichments at the manufacturing facility will be analyzed in a
systematic manner and appropriate protection will be provided against
equipment malfunctions, human errors, external hazards, and other
adverse conditions. The regulations in part 51 provide a flexible
approach for environmental review to address the range of regulated
activities under part 70. The flexibility in part 51 will enable the
NRC to determine the appropriate type of environmental review based on
the circumstances associated with the loading of fuel into a specific
manufactured reactor.
The proposed Sec. 53.620(d) cites the requirements in parts 70,
71, and 73 to ensure important features and programs are in place prior
to the receipt of SNM. The features and programs required to be in
place prior to receipt of SNM include (1) radiation monitoring
instrumentation and alarms; (2) measures to detect potential
criticality accidents; (3) appropriate procedures, equipment, and
personnel qualified for the fuel loading; (4) programs for physical
security and cybersecurity; and (5) material control and accounting
(MC&A) programs. Section 53.620(d)(2)(i) proposes requirements to
address security programs for any ML authorizing possession of a
manufactured reactor into which fuel has been loaded at the
manufacturing facility. Currently, for category II SNM, security
measures may be required in addition to requirements included in Sec.
73.67, ``Licensee fixed site and in-transit requirements for the
physical protection of special nuclear material of moderate and low
strategic significance,'' on a case-by-case basis. Including
appropriate security measures in the proposed part 53 regulations will
provide additional openness and transparency for applicants applying
for an ML who seek to load fuel into manufactured reactors at a
manufacturing site.
Currently, Sec. 73.67 only requires a security plan for licensees
who possess, use, transport, or deliver to a carrier for transport SNM
of moderate strategic significance, or 10 kg or more of SNM of low
strategic significance. However, the proposed physical security program
for fueled manufactured reactors would require a security plan for any
ML authorizing possession of a manufactured reactor into which fuel has
been loaded at the manufacturing facility, regardless of fuel type,
enrichment, and quantity. This is consistent with other controls for
MLs, including reactivity and criticality controls.
The proposed requirements would also require a holder of an ML and
part 70 license to address cybersecurity to ensure a cyberattack would
not adversely impact the functions performed by digital assets used by
the licensee for physical security, radiation monitoring, or
criticality prevention.
The proposed regulations in part 53 covering the activities related
to the storage, movement, and loading of fresh
[[Page 86933]]
fuel into a manufactured reactor in the manufacturing facility would
likewise refer to the applicable regulations in part 70. The proposed
Sec. 53.620(d) would also require the loading or unloading of
unirradiated fuel into or from a manufactured reactor and any changes
to the configuration of reactivity-related systems to be performed by a
certified fuel handler meeting the requirements in subpart F. The NRC
is aware of proposals to introduce reprocessing of existing or future
spent nuclear fuel into the fuel cycle for some potential commercial
nuclear plants. This proposed rule does not address the loading of
spent nuclear fuel or fuel resulting from reprocessing of spent nuclear
fuel into a manufactured reactor.
Section 53.620(e) would limit the transport and delivery of a
manufactured reactor or portions of a manufactured reactor only to a
site for which the Commission has issued a COL authorizing the
construction of a commercial nuclear plant using a manufactured reactor
under the specific ML. This proposed requirement is similar to the
limitations in Sec. 52.153, with the difference being that part 53
would allow the installation of a manufactured reactor at the site of a
COL but would not include provisions for installation at a site under a
CP. The possible combination of a manufactured reactor and the
licensing option of CP and OL seems unlikely and would require the
introduction of ITAAC into the licensing provisions for a CP and OL. An
additional proposed paragraph in Sec. 53.620(e) would provide
requirements for protecting fueled manufactured reactors during
transport to the site of the commercial nuclear plant by referencing
the transportation and security requirements in 10 CFR part 71,
``Packaging and Transportation of Radioactive Material,'' and part 73.
Section 53.620(f) would include proposed requirements for the
acceptance and installation of a manufactured reactor at the site of a
commercial nuclear plant. The proposed requirements would reference the
construction requirements in Sec. 53.610 to govern the integration of
the manufactured reactor into the construction of a commercial nuclear
plant. Other proposed requirements in the section would address
required receipt inspections and verification that interface
requirements between the manufactured reactor and the balance of the
commercial nuclear plant have been met.
Subpart F--Requirements for Operation
Proposed subpart F would provide the requirements for the
operations phase of a commercial nuclear plant to ensure that the
safety criteria in subpart B are satisfied throughout the plant's
lifetime and during all modes of normal operation and unplanned events.
Section 53.700 would provide the overall objectives and general
organization of subpart F, which would be to establish requirements
during operations for: (1) plant SSCs; (2) plant personnel; and (3)
plant programs.
Proposed Sec. 53.710 would provide the requirements for
maintaining capabilities, availability, and reliability of SSCs to
demonstrate compliance with the safety criteria and design requirements
for unplanned events that are described in proposed subparts B and C.
The basic structure of this proposed section would be that controls for
SR SSCs are provided by TS and controls for NSRSS SSCs are required to
be addressed with licensee-controlled documents and procedures.
The general content and control of TS under the proposed part 53
would be similar to the requirements in part 50. The proposed
requirements for TS would include limits on the inventories of
radioactive materials, plant operating limits, and specific
requirements for each SR SSC, including limiting conditions for
operation (LCO) and required surveillances. The proposed requirements
for TS would also include a section on important design elements, which
is similar to design features in Sec. 50.36, and a section for
administrative controls. A provision addressing the development and
submittal of TS to address decommissioning activities would also be
included in the proposed subpart G.
The proposed requirements for TS under part 53 would not carry over
safety limits or associated limiting safety system settings from Sec.
50.36, which contains TS requirements for operating reactors under
parts 50 and 52. As discussed in SECY-18-0096, systematic assessments
and more mechanistic approaches to evaluating source terms support an
alternative approach to establishing barrier-based safety limits. An
example provided in that paper is a comparison of: (1) the traditional
specified acceptable fuel design limits (SAFDL) that support protecting
a specific barrier from potential failure mechanisms (e.g., departure
from nucleate boiling to protect fuel cladding); and (2) the specified
acceptable system radionuclide release design limit (SARRDL) concept,
which limits the possible increase in circulating radionuclide
inventory during normal operations or an AOO as part of an integrated
or ``functional containment'' approach. Additional discussion of the
use of SARRDL in the design and licensing of advanced reactors is
provided in RG 1.232. The SARRDL could be addressed as an operating
limit within this proposed construct of requirements for TS. In cases,
such as LWRs, where a SAFDL approach might be used as part of a
mechanistic approach to meeting the design and analysis requirements in
subpart C, the associated functional design criteria proposed in Sec.
53.410 and TS under the proposed Sec. 53.710(a) would define similar
requirements as those provided by the safety limit and limiting safety
system setting requirements in Sec. 50.36.
The proposed requirements for TS under part 53 would not include
specific criteria for identifying when LCOs must be established (i.e.,
would not include an equivalent to Sec. 50.36(c)(2)(ii)). Instead,
consistent with subparts B and C, the TS requirements in subpart F of
part 53 would define TS LCOs as providing limits on SR SSCs. The SR
SSCs protect against DBAs to demonstrate compliance with the safety
criteria in the proposed Sec. 53.210. In the proposed construct for
part 53, risk-significant SSCs would be addressed through a combination
of TS for the SR SSCs and establishment and monitoring of performance
standards for NSRSS SSCs.
In addition to addressing TS for SR SSCs, proposed Sec. 53.710
would require appropriate controls be developed and implemented for
NSRSS SSCs. Examples include appropriate surveillances and controls
established through reliability assurance programs. Configuration
management and other special treatments would provide that the
capabilities, availabilities, and reliabilities of NSRSS SSCs are
maintained consistent with the underlying risk assessments while
providing flexibility to licensees through maintaining the management
functions within licensee-controlled programs. Controls on NSRSS SSCs
are appropriate as part of the overall performance-based approach
within proposed part 53. Special treatments beyond those defined for
their SR functions may also be warranted for SR SSCs to reflect their
role in meeting the safety criteria in Sec. 53.220 and the evaluation
criteria in Sec. 53.450(e). The performance objectives for NSRSS SSCs
would reflect that the comprehensive risk metrics and related risk
performance objectives established under Sec. 53.220 may involve
assessing
[[Page 86934]]
and averaging the risks over a defined period (e.g., plant year) and
would not constitute a real-time requirement that must be continuously
demonstrated by the licensee. The controls under Sec. 53.710(b)
justify proposed changes in part 53 from the traditional or
deterministic approaches in parts 50 and 52 in areas such as replacing
the single-failure criterion with a probabilistic reliability criterion
(see SRM-SECY-03-0047, ``Policy Issues Related to Licensing Non-Light-
Water Reactor Designs,'' dated June 26, 2003). This approach could also
support the incorporation of risk insights and analytical margins to
gain operational flexibilities in areas such as siting and staffing
requirements described in subsequent sections of proposed subpart F.
Proposed Sec. 53.715 would provide the requirements for developing
and implementing a program to do the following: (1) control maintenance
activities; (2) take appropriate corrective action when performance
issues are identified; (3) conduct routine evaluations of
effectiveness; and (4) assess and manage risks resulting from
maintenance activities. These proposed requirements are similar to
those included in Sec. 50.65 (maintenance rule), including the need to
assess and manage the increase in risk that may result from the
proposed maintenance activities. While, for the maintenance rule,
specific criteria must be developed to capture both SR and non-SR but
otherwise important SSCs, the proposed Sec. 53.715 would cover SR SSCs
and NSRSS consistent with other subparts in part 53.
Proposed Sec. 53.720 would provide the requirements for responding
to a seismic event during the operating phase of the life cycle of a
commercial nuclear plant and would be equivalent to the requirements in
paragraph IV(a)(3) of appendix S, ``Earthquake Engineering Criteria for
Nuclear Power Plants,'' to part 50.
The proposed part 53 would include provisions to address staffing,
training, personnel qualifications, and human factors engineering (HFE)
in a manner that is risk informed, technology inclusive, performance
based, and flexible in nature. During the development of part 53, the
staff prepared a draft white paper on ``Risk Informed and Performance
Based Human-System Considerations for Advanced Reactors,'' to support
interactions with stakeholders and the ACRS. Key considerations include
the recognition that staffing, operator qualifications, and HFE are
interconnected areas that must be approached in an integrated manner
and, furthermore, that safety functions, including the means by which
they are fulfilled, provide an effective method for informing
technology-inclusive requirements.
The requirements associated with this approach would be in
Sec. Sec. 53.725 through 53.830. Section 53.725 discusses
applicability and defines specific terms. Some definitions draw from
those in Sec. 55.4. Several new definitions would be introduced for
use within the context of subpart F. These new definitions would be the
following: ``Automation,'' ``Auxiliary operator,'' ``Generally licensed
reactor operator,'' ``Interaction-dependent-mitigation facility,''
``Load following,'' ``Self-reliant-mitigation facility.''
Sections 53.725 through 53.830 would be divided into four portions
that would cover general operational requirements, operator and senior
operator licensing requirements, generally licensed reactor operator
(GLRO) requirements, and general training requirements for plant staff.
The NRC intends to provide guidance addressing the review of operator
staffing plans; the review of operator, senior operator, and GLRO
examination programs; and the implementation of scalable HFE reviews.
Licensees would be required to use GLROs upon demonstrating compliance
with the criteria in Sec. 53.800.
Certain routine communications are necessary to facilitate the
operator licensing process. The NRC is proposing to adapt the
requirements of Sec. Sec. 55.5 and 50.74 to Sec. 53.726 to accomplish
this.
Specific information must be collected in order to facilitate the
initial issuance of operator licenses, as well as to allow for license
renewals and required updates thereafter. Such information collection
activities must also be approved by the OMB. The NRC is proposing to
adapt the requirements of Sec. 55.8, to include any needed updates in
OMB approval information, to Sec. 53.120 to accomplish this.
The information used within the regulatory processes of the NRC
must be free from omissions and inaccuracies to facilitate effective
regulation. Consistent with this, the NRC is proposing to adapt the
requirements of Sec. 55.9 to Sec. 53.728 to require the completeness
and accuracy of material information provided by individual applicants
and license holders.
Section 53.730 would provide performance-based and technology-
inclusive requirements for assessing the role of personnel in facility
safety, applying human-system considerations within facility design,
and incorporating operational approaches that are consistent with
design-specific safety considerations. Most of these requirements would
be adapted from portions of Sec. Sec. 50.34(f) and 50.54 and 10 CFR
part 55, ``Operators' Licenses,'' with considerable modification in
order to reflect the introduction of new technologies and possible
changes in the roles of personnel in preventing and mitigating events.
The NRC is proposing that these technical requirements would, together,
serve as a component of the required content of applications for OLs
and COLs under part 53. Additionally, the NRC proposes that the
specific technical requirements associated with HFE, human-system
interface design, concept of operations, functional requirements
analysis, and function allocation would serve as a component of the
required content of applications for standard DCs, standard design
approvals, MLs, and CPs, as well.
Human factors engineering is essential to facilitate the role of
personnel in facility safety in a manner that is both effective and
reliable. The NRC proposes to adapt Sec. 53.730(a) from the HFE design
requirements of Sec. 50.34(f)(2)(iii). A key difference would be that
the requirement would now be focused on settings where personnel
fulfill their safety or emergency response roles wherever they may
occur. The NRC additionally proposes to include within the scope of
this requirement activities for assuring the continued availability of
plant equipment that is needed for safety, and envisions that this may
encompass relevant maintenance, inspections, and testing as well. The
NRC intends that this requirement would be associated with staff
guidance for conducting scalable reviews of HFE that is planned to
accompany part 53.
Human-system interfaces provide vital information to operators
across a spectrum of operating conditions that can range from normal
operations through severe accident conditions. The specific types of
information that must be available to support operations staff during
such conditions include, in part, those associated with safety function
parameters, safety system status, possible core damage states, barrier
integrity, and radioactive leakage. Due to the importance of such
information, the NRC proposes under Sec. 53.730(b) to require such
human-system interface design features for all facilities, irrespective
of other flexibilities proposed under part 53. Therefore, the NRC
proposes to adapt specific post-Three Mile Island requirements of Sec.
50.34(f) in a technology-inclusive manner as detailed in the following:
[[Page 86935]]
Paragraph (b)(1) would be adapted from Sec.
50.34(f)(2)(iv).
Paragraph (b)(2) would be adapted from Sec.
50.34(f)(2)(v).
Paragraph (b)(3) would be adapted from Sec.
50.34(f)(2)(xi), 50.34(f)(2)(xii), and 50.34(f)(2)(xxi).
Paragraph (b)(4) would be adapted from Sec.
50.34(f)(2)(xvii), 50.34(f)(2)(xviii), 50.34(f)(2)(xix), and
50.34(f)(2)(xxiv).
Paragraph (b)(5) would be adapted from Sec.
50.34(f)(2)(xxvi).
Paragraph (b)(6) would be adapted from Sec.
50.34(f)(2)(xxvii).
In addition to the requirements of Sec. 53.730(b)(1) through (6),
a further set of human-system interface design requirements applicable
only to those facilities that will be staffed by GLROs would be
provided under Sec. 53.730(b)(7). This prescriptive set of design
requirements for those facilities which demonstrate compliance with the
criteria of Sec. 53.800 would recognize that the application of HFE
under Sec. 53.730(a) is anticipated to be significantly reduced at
such facilities in the absence of an expected operator role for the
fulfillment of safety functions. However, it should be noted that the
capability for an immediately initiated, manual reactor shutdown would
be conservatively mandated irrespective of any other design
considerations.
The NRC proposes Sec. 53.730(c) to require the submittal of a
concept of operations that is of sufficient scope and detail to
appropriately inform the staff. The development of a concept of
operations can facilitate a clear understanding on the part of the NRC
for potential novel operating concepts. Additionally, such information
is likely to reduce the degree of resources and interactions needed for
the NRC to obtain the understanding necessary to enable flexible
requirements in areas such as staffing, operator qualifications, and
HFE.
The NRC proposes Sec. 53.730(d) to require the submittal of both a
Functional Requirements Analysis and a Function Allocation. The
identification of design-specific safety functions and how they are
fulfilled serves as a primary means for achieving technology-inclusive
requirements within areas such as staffing, operator qualifications,
and HFE. The Functional Requirements Analysis and Function Allocation
processes (which are both HFE methods derived from systems engineering
principles), provide an effective means to identify both how safety
functions will be satisfied and how to characterize any associated
operator role in doing so. A Functional Requirements Analysis shows
what features, systems, and human actions are relied upon to
demonstrate safety (i.e., fulfill safety functions). A Function
Allocation then describes how safety functions are assigned to both
personnel and automatic systems. However, an important adaptation of
the Function Allocation for use under the proposed rule would be the
further need to not only describe allocations of safety functions to
human action and automation, but also to identify allocations made to
active safety features, passive safety features, or inherent safety
characteristics as well.
Operating experience provides an important source of information by
which to inform various aspects of facility design and operations.
Accordingly, the NRC proposes in Sec. 53.730(e) to adapt the
requirements of Sec. 50.34(f)(3)(i) for requiring an operating
experience program.
New technologies may involve concepts of operations that are more
conducive to customizable licensed operator staffing requirements than
the prescriptive requirements of Sec. 50.54(m). Analyses and
assessments that are based on HFE principles provide a performance-
based means of determining licensed operator and senior operator
staffing needed to support safe operations. In contrast, for those
facilities required to be staffed by GLROs, the NRC anticipates that
the operator staffing plans will reflect a simpler approach of showing
that a continuity of responsibility will be maintained for facility
operations throughout the operating phase, with at least one GLRO
providing continuous oversight and remaining immediately available when
any units are fueled. Additionally, a revised approach to the
traditional position of the shift technical advisor that focuses on the
availability of engineering expertise as a means of addressing
uncertainties and abnormal circumstances is more suitable within the
context of part 53 and is intended to be applicable to all facilities,
irrespective of other design and staffing considerations.
Consistent with this approach, the NRC proposes under Sec.
53.730(f) to require the submittal of a staffing plan that details
operations staffing, how engineering expertise will be provided, and
what staffing will be available to provide other needed support
functions. The NRC intends that this requirement would be associated
with staff guidance for reviewing operations staffing plans that is
planned to accompany part 53 and that, following NRC approval of the OL
or COL, the staffing plan would become a condition of the facility
license. The NRC intends that, at a minimum, the approved licensed
operator and senior operator (or, if applicable, GLRO) staffing,
positions, and personnel locations will be incorporated into
corresponding requirements within the facility TS and that a license
amendment would thus be required for any subsequent changes.
Operator training and qualification programs provide an essential
component of supporting human performance in implementing tasks with
safety implications. Such programs must include components that cover
the stages of initial training, examination, and continuing training.
Additionally, recognizing the potential for varying concepts of
operations to affect traditional, prescriptive approaches to operator
proficiency, the NRC proposes under part 53 to allow facilities to
develop operator proficiency programs based on facility-specific
considerations.
Therefore, the NRC proposes in Sec. 53.730(g)(1) to require
approval as part of its approval of the OL or COL, of the programs that
will be used for the initial training, initial examination,
requalification training and examination, and proficiency of both
licensed operators and senior operators. In a corresponding manner, the
NRC proposes in Sec. 53.730(g)(2) to require approval of the programs
that will be used for the GLRO equivalents of each of these programs
for facilities with such staffing. The NRC intends that examination
program requirements would be associated with staff guidance for the
review of tailored examination processes that are planned to accompany
part 53. Following the completion of an initial training program,
continuing training programs provide an important means of sustaining
the knowledge and abilities of individuals. The NRC is proposing to
adapt the requirements of Sec. 50.54(i-1) in Sec. 53.730(g)(3) to
require that operator continuing training programs be in effect to
support operator performance. Under part 53, the NRC proposes to
require these programs to be in effect concurrent with when the initial
operator examinations first commence, in effect putting the programs in
place only when they are needed. This represents a modification of the
comparable requirement of Sec. 50.54(i-1), which links the
commencement of these programs to a timeline driven by the licensing of
the facility.
The authorization to manipulate controls of the facility that
directly affect reactivity or power level is restricted to individuals
who are either licensed operators, licensed senior operators, or GLROs.
However, for practical purposes, situations in which
[[Page 86936]]
an individual is participating in an approved training program or
reestablishing proficiency may also call for them to operate the
controls of the facility under the cognizance of a licensed individual.
The NRC is proposing to adapt the requirements of Sec. 55.13 in Sec.
53.735 to accomplish this, with a notable difference being the
incorporation of GLROs.
Section 53.740 would provide requirements for OL and COL holders
under part 53. Portions of Sec. 53.740 would be adapted from the
conditions of Sec. 50.54. In general, the conditions for operations
staffing under part 53 would reflect considerations for potential
technological differences and varying concepts of operation that are
expected among part 53 facility licensees. Additionally, certain
requirements would be specific to the operating phase while others
would remain in effect following the permanent cessation of facility
operations during the decommissioning phase.
All commercial nuclear plants licensed under part 53 would require
some form of licensed operator staffing, whether it be by specifically
or generally licensed operators. Consistent with this, the NRC is
proposing under Sec. 53.740(a) to require facility licensees to
demonstrate compliance with the programmatic requirements for either
specifically licensed operators and senior operators or for GLROs, as
applicable to the facility.
The NRC recognizes that technology-inclusive facility staffing will
need to account for a potentially wide range of concepts of operations;
for this reason, flexible and performance-based approaches for
establishing required facility staffing are appropriate. However, once
the appropriate facility staffing has been determined and approved by
the NRC, such staffing must be maintained to ensure that the
appropriately qualified individuals will be available when needed to
support the safe operation of the facility. Therefore, the NRC is
proposing under Sec. 53.740(b) to require that the staffing described
within the approved facility staffing plan be maintained as a condition
of the facility license as opposed to prescriptive staffing
requirements like those of Sec. 50.54(k) and (m).
Because operation of facility controls directly affects reactivity
or power level, only those individuals who possess appropriate levels
of qualification and authorization are permitted to operate those
controls. The NRC is proposing to adapt the requirements of Sec.
50.54(i) in Sec. 53.740(c) to require that only specifically licensed
operators and senior operators or, alternatively, GLROs, may operate
facility controls, with allowance for specified exceptions for the
purposes of operator training or proficiency.
Senior operators, by virtue of their license level, are qualified
and authorized both to perform certain important responsibilities and
to direct the licensed activities of licensed operators. Therefore,
facilities that are required to be staffed by specifically licensed
operators must also include senior operators within their staffing. In
contrast, facilities staffed with GLROs only have a single license
level available and, therefore, there is no equivalent provision for
such facilities. The NRC is proposing to adapt the requirements of
Sec. 50.54(l) in Sec. 53.740(d) to require the licensing and
designation of senior operators at facilities staffed by specifically
licensed operators.
In contrast with control manipulations that directly affect reactor
power and reactivity (e.g., control rod movement, control drum
rotation, recirculation pump speed adjustment, reactor coolant system
boration or dilution, etc.) and are therefore restricted to performance
only by licensed operators, other types of plant operations that may
result in reactor power and reactivity changes via means that are
indirect in nature (e.g., electrical generation changes, turbine bypass
valve operation, steam usage by process heat applications, etc.) may be
implemented by non-licensed personnel. However, due to the potential
influence of such operations on reactor power and reactivity, the
continuous oversight of reactor parameters by a licensed operator is
necessary during these operations. The NRC is therefore proposing to
adapt the requirements of Sec. 50.54(j) in Sec. 53.740(e) to require
appropriate oversight of operations, other than those associated with
the controls themselves, that may affect reactivity or power level.
Load following where plant output automatically changes in response
to externally originated instructions or signals is not permitted under
the existing regulations of Sec. 50.54. However, new technological
considerations and concepts of operation may justify such an
operational approach under appropriate circumstances. The NRC
recognizes that, beyond electrical power generation, load following may
also affect other applications of plant output, such as hydrogen
production, desalination, or district heating. For load following to be
permissible, measures must be in place to provide assurance that plant
output considerations are not permitted to lead to challenges to safe
reactor operations. These measures may consist of automated control
systems, automatic protective features, or the continuous oversight and
immediate intervention capability of an appropriately qualified and
authorized individual. Section 53.740(f) would allow for load
following, provided that appropriate measures are in place. In
considering the acceptability of the measures associated with load
following, the NRC expects that any automatic protection relied upon
would be separate from that credited for reactor protection purposes
and would employ setpoints that are set so as to prevent actuation of
the reactor protection system while accomplishing its functions to the
extent practical.
Core alterations such as refueling are associated with specific
considerations that warrant limiting the oversight of such operations
to appropriately qualified and authorized individuals. Unlike other
types of fuel handling operations, core alterations occur within the
confines of a reactor vessel that is specifically designed to support
and sustain nuclear criticality, thereby justifying the imposition of
higher qualification levels within such contexts. The NRC is proposing
to adapt the requirements of Sec. 50.54(m)(2)(iv) in Sec. 53.740(g)
to require the supervision of core alterations by either a specifically
licensed senior operator, a specifically licensed senior operator whose
license is limited to fuel handling, or by a GLRO, as applicable to the
facility. Because certain commercial reactor designs may be capable of
refueling while at power and, in any event, overall facility oversight
would already be required by either a specifically licensed senior
operator or by a GLRO, the NRC proposes to omit this requirement as
redundant during periods where core alterations occur while the plant
is operating.
It is impossible to predict every possible scenario that a
commercial nuclear plant might potentially encounter. Therefore, it is
prudent to grant the authority for appropriately qualified individuals
to depart from facility license conditions when emergency circumstances
dictate that doing so is in the interest of public health and safety.
The NRC is proposing to adapt the requirements of Sec. 50.54(x) and
(y) in Sec. 53.740(h) to permit specific individuals to authorize
departures from facility license conditions or TSs when emergency
conditions warrant doing so for the protection of the public health and
safety. Recognizing that certain facilities licensed under part 53 may
be staffed by GLROs in lieu of specifically licensed senior operators,
the NRC proposes to extend this authority to
[[Page 86937]]
GLROs. While it is not anticipated that GLROs will have a role in the
fulfillment of safety functions at self-reliant-mitigation facilities
and, furthermore, that operators at such facilities would not be in a
position by which to significantly influence radiological safety
outcomes, the very nature of the Sec. 50.54(x) and (y) and the
proposed Sec. 53.740(h) provisions concern situations that are
unanticipated and, therefore, unforeseeable. Thus, it is appropriate to
grant GLROs a comparable authority to that of senior licensed operators
and certified fuel handlers as it relates to invoking this provision
under emergency conditions as a means of accounting for such
possibilities.
Due to the unique authorities and responsibilities of both
specifically and generally licensed reactor operators, it is essential
that any individual fulfilling such a role demonstrate compliance with
the regulatory requirements for operator licensing. Section 107 of the
Act authorizes the Commission to prescribe conditions for the licensing
of operators and to issue licenses consistent with those conditions.
The NRC is proposing to adapt the requirements of Sec. 55.3 in Sec.
53.745 to require that any person performing the function of an
operator, senior operator, or GLRO must be authorized by a license
issued by the Commission.
The NRC proposes to license individuals as operators under both
specific and general licensing frameworks. Specific licenses would be
for licensed operators (i.e., reactor operators) and senior operators
(i.e., senior reactor operators) and would be issued to a named person
upon approval by the Commission of an application for that named
person. In contrast, GLROs would perform duties under the provisions of
a general license that would be effective without the filing of an
application with the Commission or the issuance of licensing documents
to a particular person. The NRC proposes requirements for the use of a
specific licensing process for licensed operators and senior operators
under Sec. Sec. 53.760 through 53.795, with Sec. 53.760 addressing
applicability.
Medical fitness is an important component of the overall process of
specifically licensing operators because it provides assurance that
operators will be able to carry out important duties without being
precluded from doing so by health-related issues. Medical fitness also
provides assurance that such issues will not adversely affect the
performance of assigned job duties or cause operational errors that
endanger public health and safety. In addition to a requirement for
medical fitness, a medical examination by a physician to confirm
compliance with this requirement is necessary. The NRC is proposing to
adapt the requirements of Sec. Sec. 55.21, 55.23, and 55.27 under
Sec. 53.765 to require medical fitness, examinations by physicians,
and medical certification for specifically licensed operators and
senior operators. In recognition of the fact that GLROs are not
expected to have a role in the fulfillment of safety functions at the
facilities at which they are licensed, the NRC proposes to not extend a
comparable medical requirement to GLROs.
The NRC is also proposing to adapt the requirements of Sec. Sec.
55.25 and 50.74(c) in Sec. 53.770 to require that timely notifications
be made to the NRC if a specifically licensed operator or senior
operator develops a permanent physical or mental condition that
adversely affects the performance of assigned operator job duties or
could cause operational errors endangering public health and safety.
Notwithstanding this requirement related to permanent medical
conditions, the NRC continues to recognize that it is appropriate for
facility licenses to impose administrative restrictions and conditions
upon specifically licensed operators and senior operators in response
to temporary medical conditions.
The process of specifically licensing individuals as licensed
operators or senior operators requires the submittal of applications to
the NRC for review. These applications must detail certain elements
associated with licensing, including the demonstration of compliance
with examination, experience, and medical requirements. The NRC is
proposing to adapt the requirements of Sec. Sec. 55.31 through 55.35
in Sec. 53.775 to include requirements for the applications associated
with the specific licensing of licensed operators and senior operators
at commercial nuclear plants licensed under part 53. In contrast with
the part 55 requirements, the NRC proposes to provide additional
flexibility by locating certain details associated with the preparation
and submittal of these applications within guidance in lieu of
placement within this proposed rule itself.
The NRC proposes overall programmatic requirements for specifically
licensed operator and senior operator training, examination, and
proficiency in Sec. 53.780. In general, the proposed requirements are
adapted from those in part 55, with several additional flexibilities
being incorporated to better account for potential variations in
reactor technologies and concepts of operations. The requirements
proposed in Sec. 53.780 cover, in part, the initial training, initial
examination, requalification training, requalification examination, and
proficiency of specifically licensed operators and senior operators.
The initial training process provides individuals with the
knowledge and abilities needed to subsequently fulfill assigned duties
as licensed operators or senior operators in a safe and reliable
manner. The use of a systems approach to training (SAT) ensures that
the training program is based upon job requirements in a manner that
can be adapted to account for differences in plant technology, concepts
of operations, and operator roles in the fulfillment of design-specific
safety functions. The NRC is proposing under Sec. 53.780(a) to require
facility licensees to implement a SAT-based training program for the
initial training of licensed operator and senior operator applicants.
The program must be adequate to ensure that applicants will be capable
of performing the duties necessary both to protect public health and
safety and to maintain plant safety functions. The NRC further proposes
that such programs be subject to NRC approval and subsequent change
control processes of an appropriate nature.
Examinations provide a means of assessing that individuals have
achieved a degree of knowledge and ability that is sufficient to carry
out assigned duties as licensed operators or senior operators in a
manner that is safe and reliable. The NRC is proposing to adapt the
requirements of Sec. Sec. 55.40, 55.41, 55.43, and 55.45 in Sec.
53.780(b) to require that facilities establish and implement an initial
examination program. However, a key difference from the comparable
requirements of part 55 would be that facilities have the flexibility
to propose, subject to NRC approval, the examination methods and
criteria to be used in assessing satisfactory applicant performance.
Such examination programs (including those used within the scope of
requalification training) would need to provide for acceptable levels
of both test validity and test reliability in order to be considered
acceptable. The NRC intends that staff guidance would be available to
facilitate the review of licensing examination programs that are
proposed by facility licensees and that, following NRC approval,
initial examination programs would be subject to an appropriate change
control process. Furthermore, the NRC proposes that holders of licenses
to operate commercial nuclear
[[Page 86938]]
plants under part 53 be provided the alternative of administering their
own approved licensing examinations. The NRC would continue to exercise
appropriate oversight of the program, make operator licensing decisions
based upon the examination results, and reserve the right to administer
the examinations in lieu of permitting the facility to do so. However,
irrespective of the provided flexibilities in examination format and
structure, at a minimum, topics from the following general categories
of knowledge and abilities should be sampled in such examinations:
Reactor Theory, Thermodynamics, and Chemical Interactions
Plant Systems and Components
Reactivity Management and Manipulations
Radiation Control and Safety
Emergency, Abnormal, and Normal Operations
Administrative Requirements and Conditions of the Facility
License
Requalification training programs provide for the continuing
training and examination of specifically licensed operators and senior
operators to ensure that they maintain the knowledge and abilities
needed to support the safe and reliable performance of job duties
following the completion of an initial training and examination
program. The NRC is proposing to adapt the requirements of Sec. 55.59
in Sec. 53.780(c) to require that facilities implement both a SAT-
based requalification training program and a biennial requalification
examination program. However, a notable difference from the biennial
requalification examinations required under part 55 would be that
distinct annual operating test and biennial written examination
components would not be mandated, with the facility licensee instead
proposing the examination methods and criteria to be used in assessing
satisfactory performance. The NRC intends that guidance would be
available to facilitate the review of the requalification examination
programs that are proposed by facility licensees and that, following
NRC approval, requalification examination programs would be subject to
an appropriate change control process.
For examinations to provide for valid assessments of the knowledge
and abilities of individuals, the examinations must remain free from
compromises that could affect their underlying integrity. The NRC is
proposing to adapt the requirements of Sec. 55.49 in Sec. 53.780(d)
to require that examinations and related activities remain free from
any compromise that might affect the integrity of the examination
process.
Simulators provide a valuable means of training and evaluating
plant operators, and the NRC is specifically authorized under the
Nuclear Waste Policy Act of 1982, as amended (NWPA), section 306 (42
U.S.C. 10226) to establish regulations for the use of simulators within
such context. The NRC is proposing to adapt the requirements of Sec.
55.46 in Sec. 53.780(e) to address the use of simulation facilities
for training, examinations, and applicant experience requirements, as
well as to address the maintenance of simulator fidelity. However, the
proposed requirements of part 53 would not mandate that full scope,
plant-referenced simulators be used and would allow the use of
alternative simulation facilities consisting of, for example, partial
scope simulators or the plant itself, provided that all associated
requirements can be demonstrated to be met using alternative approaches
and methods. Additionally, in allowing for the possibility that an
applicant or licensee might demonstrate compliance with training,
examination, or experience requirements using the plant itself, the NRC
is not allowing the initiation of transients on the actual plant.
Consistent with this, aside from controlled reactivity manipulations
that are conducted for the purposes of demonstrating compliance with
experience requirements, actual plant components may not be operated
for these purposes. Rather, the NRC perspective is that the use of the
plant for training and examination purposes should be restricted to
techniques such as walkthroughs, job performance measures, simulated
tasks, use of augmented reality technology, and similar approaches that
provide training and examination value while avoiding the operation of
actual plant components.
There may be situations in which applicants for operator or senior
operator licenses have previous training and experience that justifies
waiving some, or all, of the initial examination requirements. The NRC
is proposing to adapt the requirements of Sec. 55.47 in Sec.
53.780(f) to allow for consideration of requests for waivers of
examinations requirements. In contrast with the part 55 requirements,
the NRC proposes to locate certain details associated with such waiver
requests within guidance documentation in lieu of placement within the
rule itself.
For licensed operators and senior operators to perform their
assigned duties safely and reliably, it is essential that they perform
those duties frequently enough so as to maintain a sufficient degree of
proficiency. The NRC is proposing to adapt the requirements of Sec.
55.53(e) and (f) in Sec. 53.780(g) to require that specifically
licensed operators and senior operators maintain proficiency and, if
proficiency is not maintained, regain proficiency prior to resuming
licensed duties. However, in recognition of the fact that varying
concepts of operations are possible for advanced reactor facilities,
the NRC is proposing, in contrast with the requirements of part 55, to
allow facility licensees to establish their own programs for operator
proficiency, subject to NRC approval.
As the holders of specific licenses, licensed operators and senior
operators must be subject to license conditions on an individual basis
to ensure that the basis upon which the licenses were issued remains
valid. The NRC is proposing to adapt the requirements of Sec. 55.53 in
Sec. 53.785 to require appropriate conditions of licenses for
specifically licensed operators and senior operators. However, in
contrast with the requirements of Sec. 55.53(e) and (f), the NRC is
proposing to allow certain aspects of operator proficiency to be
addressed by an NRC-approved facility proficiency program.
Licenses for specifically licensed operators and senior operators
are issued by the NRC and must remain subject to modification or
revocation. The NRC is proposing to adapt the requirements of
Sec. Sec. 55.51 and 55.61 in Sec. 53.790 to address the issuance,
modification, and revocation of licenses issued to specifically
licensed operators and senior operators.
The licenses issued to specifically licensed operators and senior
operators are valid for a period of six years, after which they expire,
unless otherwise renewed. The NRC is proposing to adapt the
requirements of Sec. Sec. 55.55 and 55.57 in Sec. 53.795 to address
the expiration and renewal of licenses issued to specifically licensed
operators and senior operators.
In developing this proposed rule, the NRC has discussed with
stakeholders the considerations that might justify the omission of the
specifically licensed operators and senior operators. However, even for
an inherently safe reactor with autonomous operation features, certain
important administrative functions (e.g., compliance with TS,
operability determinations, NRC notifications, emergency declarations,
risk assessment, maintenance oversight, and radiological release limit
compliance) would still need to be accomplished by
[[Page 86939]]
appropriately qualified and authorized individuals. Additionally, the
NRC recognized that manual manipulations of facility reactivity
controls must only be performed by individuals who have been
appropriately licensed by the Commission. The NRC therefore proposes
under Sec. 53.800 to establish a new class of facility (defined as a
self-reliant-mitigation facility), according to the criteria contained
in Sec. 53.800 for part 53. These facilities would employ GLROs rather
than specifically licensed operators and senior operators. The GLRO
regulations offer enhanced flexibilities and targeted relaxations in a
manner that is commensurate with the modified role of such operators to
ensure the safe operation of the associated facilities. In contrast,
those facilities not meeting the criteria of Sec. 53.800 would instead
be considered interaction-dependent-mitigation facilities and would
require staffing by specifically licensed operators and senior
operators. The terminology used to designate these facility types
reflects differences in how operators are anticipated to need to
interact with their plant systems in mitigating events and achieving
safe outcomes; such systems may either need operators to interact with
them in some manner (i.e., be interaction-dependent) or may instead be
able to rely fully upon their own capabilities independent of operator
interaction (i.e., be self-reliant).
Generally licensed reactor operators would differ from specifically
licensed operators because the latter would be directly and
independently evaluated by the NRC as part of their licensing process.
This direct and independent evaluation remains appropriate when
operators may reasonably be expected to exert a significant influence
on public health and safety outcomes. Therefore, a key determinant as
to whether generally licensed reactor operators can be utilized in
facility staffing is the assessment of the operator's role in
maintaining and fulfilling safety functions at the facility, such as
through the performance of credited actions for the mitigation of plant
events.
The criteria proposed in Sec. 53.800 would designate self-reliant-
mitigation facilities. These criteria are derived from the following
set of considerations:
no human action needed to satisfy radiological consequence
criteria;
no human action needed to address LBEs;
safety functions not allocated to human action;
reliance upon robust and highly reliable safety features;
and
adequate defense in depth achieved without reliance on
human action.
It should be noted that those facilities not meeting the criteria
proposed in Sec. 53.800 would instead be classified as interaction-
dependent-mitigation facilities and would require staffing by
specifically licensed operators and senior operators instead.
Generally licensed reactor operators would perform duties under the
provisions of a general license that would be effective without the
filing of an application with the Commission or the issuance of
licensing documents to a particular person. The NRC proposes
requirements for the general licensing process for GLROs under
Sec. Sec. 53.805 through 53.820. The requirements for GLROs would
parallel those for senior operators in regard to their comparable
administrative responsibilities. Nonetheless, the requirements for
GLROs would be relaxed and incorporate greater flexibilities compared
to the requirements for specifically licensed operators in a manner
that is consistent with the GLRO's role in safety at self-reliant-
mitigation facilities.
In order to use GLROs in lieu of specifically licensed operators
and senior operators, a OL/COL applicant would need to demonstrate that
its proposed facility is a self-reliant-mitigation facility, i.e., that
it will comply with the following requirements on an ongoing basis:
maintaining GLRO qualifications for the performance of important
functions and tasks; incorporating relevant programmatic controls into
TS; administering the related programs for training, examination, and
proficiency; and ensuring that the relevant provisions of parts 26 and
73 are met. Additionally, to provide for an accurate accounting of what
individuals are licensed under the general license, facility licensees
would be required to report the identities of all generally licensed
reactor operators to the NRC on an annual basis. Furthermore, a
facility licensee must ensure that the facility design and performance
continue to meet the technological criteria to be classified as a self-
reliant-mitigation facility (i.e., the criteria of Sec. 53.800) on a
continual basis during the operating phase, as the relaxations afforded
to such facilities in the areas of operator licensing, staffing, and
HFE would be predicated on this assumption. The NRC therefore proposes
under Sec. 53.805 to establish requirements for facility licensees
that address issues such as these. Finally, the failure of a self-
reliant-mitigation facility to subsequently meet the criteria of Sec.
53.800 after the issuance of an OL or COL would constitute a reportable
event (i.e., an unanalyzed condition that significantly degrades plant
safety) under the provisions of Sec. 53.1630.
The NRC proposes the general license for GLROs under Sec. 53.810.
GLROs would be licensed as a class of individuals under the provision
of Sec. 53.810(a) and would be subject to the conditions specified in
Sec. 53.810(b) through (g). Portions of these conditions are adapted
from Sec. 55.53 and from those conditions currently included in the
licenses issued to specifically licensed operators and senior
operators. The NRC would retain the ability to suspend or prohibit
individuals from operating under the general license should such action
be warranted.
The NRC proposes overall programmatic requirements for GLRO
training, examination, and proficiency under Sec. 53.815. In general,
these proposed requirements are adapted from those of part 55 and
parallel those also proposed for specifically licensed senior operators
in Sec. 53.780. These requirements include increased flexibilities and
several targeted relaxations that reflect the limited role of GLROs in
facility safety. The requirements proposed under Sec. 53.815 cover, in
part, the initial training, initial examination, continuing training,
requalification examination, and proficiency of GLROs. Section 53.805
would require the facility licensee to develop, implement, and maintain
these programs. Section 53.810, in turn, would prescribe that the
requirements of Sec. 53.805 would need to be met as a requirement of
the general license. The implication of this structure is that the
facility licensee would need to implement these programs for training,
examination, and proficiency, and GLROs would need to participate in
these programs to demonstrate compliance with the requirements of the
general license.
The initial training process provides GLROs with the knowledge and
abilities needed to fulfill assigned duties as GLROs. The use of a SAT
serves to ensure that the training program is based upon job
requirements in a manner that can be adapted to account for differences
in plant technology and concepts of operations. The NRC is proposing
under Sec. 53.815(b) to require facility licensees to implement a SAT-
based training program for the initial training of GLROs that is
adequate to ensure that they have the necessary knowledge, skills, and
abilities to perform their duties. The NRC further proposes that such
programs would be subject to NRC approval, oversight, and appropriate
change control processes. The training program must ensure that
[[Page 86940]]
GLROs maintain the necessary knowledge, skills, and abilities.
Examinations provide a means of assessing that individuals have
achieved a degree of knowledge and ability that will be sufficient to
enable them to carry out assigned duties as GLROs in a manner that is
both safe and reliable. The NRC proposes to adapt the requirements of
Sec. Sec. 55.40, 55.41, 55.43, and 55.45 in Sec. 53.815(b) to require
that facility licensees establish and implement an initial examination
program. A key difference from the comparable requirements of part 55
would be that facility licensees would be afforded the flexibility to
propose, subject to NRC approval, the examination methods and criteria
to be used in assessing satisfactory individual performance. Such
examination programs (including those used within the scope of
continuing training) would need to provide for acceptable levels of
both test validity and test reliability in order to be considered
acceptable. The NRC intends that staff guidance would be available to
facilitate the review of initial examination programs that are proposed
by facility licensees and that approved initial examination programs
would be subject to an appropriate change control process. In contrast
with both the requirements of part 55 and the proposed requirements of
Sec. 53.780, the NRC does not intend to administer or evaluate these
initial examinations. However, the examination processes themselves
will continue to be subject to ongoing NRC oversight. Irrespective of
the provided flexibilities in examination format and structure, topics
from the following general categories of knowledge and abilities should
be sampled in such examinations:
Reactor Theory, Thermodynamics, and Chemical Interactions
Plant Systems and Components
Reactivity Management and Manipulations
Radiation Control and Safety
Emergency, Abnormal, and Normal Operations
Administrative Requirements and Conditions of the Facility
License
Continuing training programs provide the ongoing training and
examination of GLROs to ensure that they maintain the knowledge and
abilities needed to support the safe and reliable performance of job
duties following the completion of an initial training and examination
program. The NRC is proposing to adapt the requirements of Sec. 55.59
in Sec. 53.815(b) to require that facility licensees implement both a
SAT-based continuing training program and a requalification examination
program. However, a notable difference from the examinations required
under part 55 would be that distinct annual operating test and biennial
written examination components would not be mandated. The facility
licensee would instead propose examination methods and criteria to be
used in assessing satisfactory performance. Furthermore, unlike the
comparable requirements of part 55 and those proposed for specifically
licensed operators and senior operators, a biennial periodicity for
requalification examinations would not be prescribed. However, adequate
justification for the proposed periodicity of requalification
examinations would be required. The NRC intends that staff guidance
would be available to facilitate the review of the requalification
examination programs that are proposed by facility licensees. Approved
requalification examination programs would be subject to an appropriate
change control process.
For examinations to provide for valid assessments of the knowledge
and abilities of individuals, the examinations must remain free from
compromises that could affect their underlying integrity. The NRC is
proposing to adapt the requirements of Sec. 55.49 in Sec. 53.815(d)
to require that examinations and related activities remain free from
any compromise that might affect the integrity of the examination
process.
Simulators provide a valuable means of training and evaluating
plant operators and the NRC is specifically authorized under the NWPA,
section 306 (42 U.S.C. 10226) to establish regulations for the use of
simulators within such context. The NRC is proposing to adapt the
requirements of Sec. 55.46 in Sec. 53.815(e) to address the use of
simulation facilities for training and examinations, and experience
requirements, as well as to address the maintenance of simulator
fidelity. The use of full scope, plant-referenced simulators would not
be mandated. The potential use of alternative simulation facilities
consisting of, for example, partial scope simulators or the plant
itself, would be allowed provided that all associated requirements
could be demonstrated to be met using alternative approaches and
methods. Additionally, in allowing for the possibility that an
applicant or licensee might demonstrate compliance with training and
examination requirements using the plant itself, the NRC is not
allowing the initiation of transients on the actual plant. Consistent
with this, aside from controlled reactivity manipulations that are
conducted for the purposes of demonstrating compliance with experience
requirements, actual plant components may not be operated for these
purposes. Rather, the use of the plant for training and examination
purposes should be restricted to techniques such as walkthroughs, job
performance measures, simulated tasks, use of augmented reality
technology, and similar approaches that provide training and
examination value while avoiding the operation of actual plant
components.
There may be situations in which GLROs have previous training and
experience that justifies waiving some, or all, of the initial
examination. Therefore, the NRC is proposing under Sec. 53.815(f) to
allow facility licensees to waive some, or all, portions of initial
examinations provided that such waivers are consistent with a program
that has been approved by the NRC.
For GLROs to safely and reliably perform their assigned duties, it
is essential that they perform those duties frequently enough so as to
maintain a sufficient degree of proficiency. However, the NRC
recognizes that facilities that utilize GLROs may have concepts of
operation that warrant unique proficiency considerations. Therefore,
the NRC is proposing in Sec. 53.815(g) to require that facility
licensees develop, implement, and maintain programs to maintain and
reestablish, if needed, the proficiency of GLROs. This could occur, for
example, if an individual's extended absence from watch standing has
rendered proficiency requirements unmet.
The general license should remain in effect for an individual only
while that individual remains employed in a position that may call for
the individual to manipulate the reactivity controls of the facility.
The NRC proposes under Sec. 53.820 to require that the general license
would cease to be applicable on an individual basis when an
individual's employment status becomes such that this is no longer the
case. However, the NRC recognizes that for some types of self-reliant-
mitigation facilities, very long periods may elapse between
circumstances that necessitate manual manipulation of reactivity
controls. Therefore, the general license remains in effect for an
individual as long as the individual's current position could
potentially require that individual to manipulate reactivity controls
at some point within the course of the individual's assigned job
duties.
The NWPA, section 306 (42 U.S.C. 10226) authorizes and directs the
NRC to, in part, issue regulations and guidance that address the
training and
[[Page 86941]]
qualifications of civilian nuclear power plant operators, supervisors,
technicians, and other appropriate operating personnel. The NRC
implements this in part 50 through the requirements of Sec. 50.120,
``Training and qualification of nuclear power plant personnel.'' The
NRC is proposing under Sec. 53.830 to adapt, with modifications, the
requirements of Sec. 50.120 for use in part 53 to provide more
flexible personnel training and qualification requirements than those
in Sec. 50.120 and better reflect diverse concepts of operations.
The NRC recognizes that the categories of nuclear power plant
personnel in Sec. 50.120 may not be needed for the diverse concepts of
operations, staffing models, and non-traditional personnel roles and
responsibilities anticipated under proposed part 53; conversely, and
for the same reasons, additional categories of plant personnel may need
to be covered by part 53. The NRC also recognizes that the timeframe
prescribed in Sec. 50.120 for the establishment of training programs
may not be aligned with the schedules associated with the startup of
certain types of commercial nuclear plant facilities. However, the NRC
also recognizes that the SAT-based training required under Sec. 50.120
remains an appropriate means by which training programs should continue
to be developed and implemented. Therefore, the approach taken by the
NRC in addressing the training of certain plant staff under the
proposed part 53 reflects greater flexibilities in personnel categories
and programmatic timeframes, while still retaining the requirement that
such training programs be based on SAT.
The NRC is proposing under Sec. 53.830 to require SAT-based
training programs with the timeframe for when such programs are
required being based upon when the associated personnel are needed to
support facility-specific needs. The training programs would cover the
training and qualification of plant personnel in the general categories
of supervisors, technicians, and other appropriate operating personnel.
The licensee would not be required to seek NRC approval of a training
program prior to usage. However, the licensee is required to
accommodate NRC inspection of the training program. The NRC intends to
develop guidance to facilitate the inspection of these training
programs but does not intend for such guidance to preclude the
potential for the training programs to be maintained by a separate,
NRC-approved accreditation process.
The proposed Sec. 53.845 would require programs to be developed,
implemented, and maintained to help ensure that design features and
human actions have the capabilities and reliabilities necessary to
demonstrate compliance with the safety criteria in subpart B throughout
the operating life of each commercial nuclear plant. The proposed
programmatic requirements in subpart F would also address areas such as
radiation protection needed to control routine effluents during normal
operations. The proposed Sec. Sec. 53.850 through 53.910 would require
programs to support specific activities needed to ensure the prevention
or mitigation of unplanned events or to support normal operations for
any reactor design. However, each holder of an OL or COL would be
required to assess whether additional programs are needed for the
specific reactor design and location of the commercial nuclear plant.
Licensees would be able to combine, separate, and otherwise organize
programs and related documents as appropriate for the technologies and
organizations associated with the commercial nuclear plant.
Proposed Sec. 53.850 would require a radiation protection program
associated with the requirements in subparts B and C for public doses
resulting from normal operations and the protection of plant workers.
The proposed requirements related to doses from normal operations,
including routine effluents, would be similar to those specified in
Sec. 50.36a, ``Technical specifications on effluents from nuclear
power reactors,'' and related requirements in standard TS for offsite
dose calculation manuals. While the proposed section would include
requirements that are technically and programmatically similar to part
50, proposed Sec. 53.850 would not include a requirement for effluent-
related TS as is required in Sec. 50.36a. A proposed requirement
similar to that found in the administrative controls section of TS for
operating reactors licensed under parts 50 and 52 would be included for
programmatic controls of solid wastes to complement the design
requirements in proposed Sec. 53.425.
Proposed Sec. 53.855 would require an emergency response plan that
demonstrates compliance with the requirements in appendix E to part 50
and Sec. 50.47(b) or Sec. 50.160. The regulations in Sec. 50.47
stating that the NRC will not issue certain licenses unless it finds
that there is reasonable assurance that adequate protective measures
can and will be taken to protect public health and safety in the event
of a radiological emergency apply equally to applications under part 53
complying with the applicable standards set forth in either Sec.
50.160 or the requirements in appendix E to part 50 and Sec. 50.47(b).
In its 2008 Advanced Reactor Policy Statement, the Commission
stated their expectation that ``the safety features of advanced reactor
designs will be complemented by the operational program for Emergency
Planning (EP). This EP operational program, in turn, must be
demonstrated by inspections, tests, analyses, and acceptance criteria
to ensure effective implementation of established measures.''
Consistent with this policy statement, emergency plans and emergency
planning zones are not safety features in the design. In SECY-97-020,
``Results of Evaluation of Emergency Planning for Evolutionary and
Advanced Reactors,'' dated January 27, 1997, the staff indicated that
the rationale upon which EP for current reactor designs is based, that
is, potential consequences from a spectrum of accidents, is appropriate
for use as the basis for EP for evolutionary and passive advanced LWR
designs and is consistent with the Commission's defense-in-depth safety
philosophy. Also, in its Safety Goals Policy Statement the Commission
stated that: ``A defense-in-depth approach has been mandated in order
to prevent accidents from happening and to mitigate their consequences.
Siting in less populated areas is emphasized. Furthermore, emergency
response capabilities are mandated to provide additional defense-in-
depth protection to the surrounding population.'' Consistent with this
policy statement, proposed Sec. 53.855 contributes an additional
independent layer of defense in depth for commercial nuclear plants.
Therefore, the emergency plans and emergency planning zones under
proposed Sec. 53.855 are not used to demonstrate compliance with
subpart B and subpart C of this part. Rather, compliance with the
requirements in proposed Sec. 53.855 would provide reasonable
assurance that adequate protective measures can and will be taken to
protect public health and safety in the event of a radiological
emergency.
Proposed Sec. 53.860 would identify the applicable regulations for
part 53 applicants related to the programs for physical security,
cybersecurity, FFD, AA, and information security. These programs are
discussed in more detail in section V, ``Changes to Other Parts of 10
CFR,'' of this document.
Proposed Sec. 53.860(a) would establish the physical protection
program and present a graded approach to physical protection
requirements. If a licensee can meet the proposed criterion in
[[Page 86942]]
Sec. 53.860(a)(2)(i), then the requirement to protect against the
design-basis threat (DBT) of radiological sabotage would not be
applicable. The criterion in Sec. 53.860(a)(2)(i) would require a
licensee to show that potential consequences resulting from a DBT
initiated event would result in offsite doses below the values in Sec.
53.210 even if licensee mitigation and recovery actions, including any
operator action, are unavailable or ineffective. Where the criterion is
met, the resulting physical protection requirements would be those for
protection of SNM and Category 1 and Category 2 radioactive material,
if applicable. This proposal would apply a new regulatory approach for
certain commercial nuclear plants in which the DBT of radiological
sabotage would not be applicable.
For those licensees able to meet the criterion in Sec.
53.860(a)(2), the NRC would not conduct Force-On-Force (FOF) exercise
inspections. Section 170D.a of the Act permits the Commission to
determine which licensed facilities are part of a class of licensed
facilities where NRC-conducted FOF exercises are appropriate to assess
the ability of a private security force of a licensed facility to
defend against any applicable DBT. For the class of licensees that meet
the criterion of Sec. 53.860(a)(2), it would not be appropriate to
conduct FOF exercises to evaluate performance at commercial nuclear
plants where the DBT of radiological sabotage is not applicable and the
facility poses a lower risk to public health and safety from potential
radiation exposure. These facilities would still have tailored security
requirements and oversight consistent with their relatively low risk.
For those licensees not able to meet the criterion in Sec.
53.860(a)(2), proposed Sec. 53.860(a) would permit the licensee to
choose one of two paths to provide physical protection: (1) the current
set of requirements in Sec. 73.55, which would include any changes
resulting from the ongoing proposed rulemaking on Alternative Physical
Security Requirements for Advanced Reactors \2\ that provides pre-
determined physical security alternatives; or (2) the performance-based
requirements in proposed Sec. 73.100. In either case, the licensee
would be subject to NRC-conducted FOF inspections.
---------------------------------------------------------------------------
\2\ SECY-22-0072, ``Proposed Rule: Alternative Physical Security
Requirements for Advanced Reactors,'' dated August 2, 2022.
---------------------------------------------------------------------------
Proposed Sec. 53.860(b) would require licensees to establish,
implement, and maintain an FFD program under part 26. Section 53.860(c)
would require licensees to establish, implement, and maintain an AA
program in accordance with either Sec. 73.56 or proposed Sec. 73.120,
as appropriate. Section 53.860(d) would require licensees to establish,
implement, and maintain a cybersecurity program in accordance with
either Sec. 73.54 or proposed Sec. 73.110. Section 53.860(e) would
require licensees to establish, implement, and maintain an information
protection system that complies with the requirements of Sec. Sec.
73.21, 73.22, and 73.23, as applicable.
Proposed Sec. 53.865 would establish requirements for quality
assurance and refer to appendix B of part 50 for the part 53
requirements for SR design features. Proposed requirements related to
evaluating and reporting changes to the quality assurance program would
be included in proposed subpart I and would be equivalent to those
found in Sec. 50.54.
The proposed Sec. 53.870 would require licensees to actively
assess possible degradation of SSCs from the effects of aging, fatigue,
and environmental conditions. The proposed inclusion of requirements
related to designing and monitoring for possible degradation mechanisms
reflects important lessons learned from the history of LWRs and the
likely introduction of new design features and materials in future
commercial nuclear plants. The allowable combinations of design
features, operating experience, testing, and monitoring during
operations would support performance-based approaches to the initial
licensing of new technologies. The proposed performance-based approach
to integrity assessment programs would also allow for the subsequent
consideration of operating experience and appropriate corrective
actions or allowable relaxations for ensuring that design features
comply with the proposed functional design criteria of Sec. Sec.
53.410 and 53.420. The proposed program would be based upon a
comprehensive and integrated evaluation of the aging and other
degradation mechanisms applicable to the design; identification of the
affected SSCs; the allowances provided in the design of the SSCs for
degradation; and schedules and procedures for determining if and at
what rate degradation is occurring, as well as its cause. Risk insights
could be used to prioritize the monitoring, evaluation, and management
of degradation based upon the importance of the SSC to safety and the
time frame for when the effects of degradation could be of concern.
Proposed Sec. 53.875 would establish requirements for a fire
protection program supporting operations similar to Sec. 50.48. The
proposed fire protection program during operations would work in
concert with specific fire protection requirements proposed in subpart
C for design and analyses and in proposed subpart E for construction
and manufacturing.
Proposed Sec. 53.880 would establish requirements for an inservice
inspection (ISI) and inservice testing (IST) program, which are
historically important activities conducted in accordance with ASME
codes and regulations in Sec. 50.55a. While the proposed part 53 would
not incorporate specific consensus codes and standards into the
regulations, Sec. 53.880 allows for the use of generally accepted
codes and standards. The proposed requirement for an ISI and IST
program would reinforce the need to develop monitoring programs to be
conducted during a plant's operations phase to complement the design
process and address inherent uncertainties. The NRC encourages the
continued use of consensus codes and standards supporting design,
testing, and inspections to support integrated and performance-based
approaches in demonstrating compliance with the proposed requirements
in part 53.
Proposed Sec. 53.910 would establish requirements for developing,
implementing, and maintaining procedures (e.g., operations and
emergency operating procedures) and guidelines (e.g., accident
management guidelines). The programmatic requirements for many of the
procedures listed in this proposed section would be similar to the
requirements found in the administrative controls section of TS for
plants licensed under parts 50 and 52. The proposed inclusion, where
appropriate, of accident management guidelines in these requirements is
intended to ensure that an integrated set of procedures and guidelines
would be established by licensees to ensure command and control across
the spectrum of possible event sequences. The proposed required
procedures would also include those needed to complement the design
requirements in proposed Sec. 53.440(m) related to criticality alarms
and the equivalent of the procedures required in Sec. 50.54(hh) to
address notifications of potential aircraft threats.
Subpart G--Decommissioning Requirements
The proposed subpart G would provide the regulatory requirements
for the decommissioning phase of the life cycle of a commercial nuclear
plant.
[[Page 86943]]
The requirements being proposed in subpart G for the decommissioning of
a commercial nuclear plant are adapted from the current regulations in
Sec. 50.75, ``Reporting and recordkeeping for decommissioning
planning,'' Sec. 50.82, ``Termination of license,'' and Sec. 50.83,
``Release of part of a power reactor facility or site for unrestricted
use.'' Although the requirements from those sections of part 50 have
been copied into proposed subpart G with relatively few changes, the
requirements are reorganized to fit within the part 53 structure. The
few changes made were primarily to make the proposed requirements more
technology inclusive by adding alternatives within sections, whereas
some requirements in part 50 were developed specifically for LWRs.
As an example, Sec. 50.75 provides minimum amounts of
decommissioning funds required to demonstrate reasonable assurance of
funds for decommissioning LWRs. Such generic amounts have not been
developed for all reactor technologies that may be licensed under part
53. Therefore, the Commission proposes in Sec. 53.1020, ``Cost
estimates for decommissioning,'' that site-specific cost estimates for
decommissioning must be developed considering costs in such areas as
engineering, labor, and waste disposal. The derivation of the generic
cost estimates for LWRs in Sec. 50.75 is provided in NUREG/CR-5884,
``Revised Analyses of Decommissioning for the Reference Pressurized
Water Reactor Power Station,'' and NUREG/CR-6187, ``Revised Analyses of
Decommissioning for the Reference Boiling Water Reactor Power
Station.'' Similar to part 50, a provision for an annual adjustment of
decommissioning cost estimates would be included in proposed Sec.
53.1030.
The NRC is currently pursuing another rulemaking, ``Regulatory
Improvements for Production and Utilization Facilities Transitioning to
Decommissioning,'' which was published as a proposed rule for public
comment on March 3, 2022 (87 FR 12254). As these rulemakings progress,
the NRC will consider revisions to part 53 to align the two rulemaking
efforts. For example, the proposed Sec. 53.1075 could be expanded to
include or reference requirements for decommissioning in areas such as
EP and security in addition to the proposed decommissioning fire
protection plans that would provide an equivalent to Sec. 50.48(f).
Subpart H--Licenses, Certifications, and Approvals
Proposed subpart H would provide requirements related to
applications under part 53 for NRC licenses, certifications, or
approvals for commercial nuclear plants.
Proposed subpart H would specify requirements applicable to all
part 53 applications as well as requirements specific to part 53
applications for LWAs, ESPs, standard design approvals, standard DCs,
MLs, CPs, OLs, and COLs. Proposed subpart H would be equivalent to and
include all existing licensing, certification, and approval processes
currently covered under parts 50 and 52, with the exception of the
process for early review of site suitability issues. Interactions with
external stakeholders during the development of the proposed rule did
not identify significant interest in or need for including the process
for early review of site suitability issues in part 53.
Much of the proposed subpart H regulatory text is identical to the
corresponding language in parts 50 and 52, with minor changes to
account for cross references in part 53, to make language technology
neutral, or to reflect the unique analytical approach in part 53. In
these instances, this preamble discussion will describe the language as
``equivalent'' to the existing corresponding requirement in part 50 or
part 52 and will describe any deviations, where applicable.
Because part 53 carries over the majority of the licensing options
from parts 50 and 52, there are several sections in proposed subpart H
that are similar to existing regulations in parts 50 and 52. Proposed
Sec. 53.1100 would address filing of applications for licenses,
certifications, or approvals under oath or affirmation and is
equivalent to Sec. 50.30. The proposed Sec. 53.1100 does not include
the current requirement in Sec. 50.30(a)(2) that the applicant
maintain the capability to generate additional copies, because it is
unnecessary in the age of electronic submissions. In addition, the
existing requirement on applications for OLs in Sec. 50.30(d) is
included in proposed Sec. 53.1124(g)(2), ``Relationship between
sections,'' covering OLs, rather than in proposed Sec. 53.1100.
Proposed Sec. 53.1101 would lay out activities requiring an NRC
license and is equivalent to Sec. 50.10(b). Proposed Sec. 53.1103
would address combining applications and is equivalent to Sec. Sec.
50.31, 50.52, and 52.8. Proposed Sec. 53.1103(b) would continue the
Commission's practice of combining multiple authorizations for a
facility under parts 30, 40, 50, 52, and 70 into one license based on
the Commission's authority under Section 161h. of the Act to combine
NRC licenses. Proposed Sec. 53.1106 would address elimination of
repetition and is equivalent to Sec. 50.32.
Proposed Sec. 53.1109 would provide general information
requirements for the content of applications submitted to the NRC under
part 53 and is equivalent to Sec. 50.33, with the exception of Sec.
50.33(f) on financial qualifications, which is covered in proposed
subpart J, and Sec. 50.33(h) on earliest and latest dates for
completion of construction, which is covered in Sec. 53.1306 of this
subpart. Each application would need to include information to address
the items in proposed Sec. 53.1109 as cited in the appropriate section
of this subpart for the application type.
One change from current requirements can be found in proposed Sec.
53.1109(i), which is not limited to electricity generation as it is
currently in part 50. Some prospective NRC applicants are considering
development of nuclear plants for other commercial ventures, such as
process heat generation or hydrogen production. In addition, Sec.
53.1109(j), which requires applications containing classified
information to separate that information from the unclassified
information in the application, refers to ``Restricted Data or
classified National Security Information'' instead of the term used in
the corresponding provision in Sec. 50.33(j), ``Restricted Data or
other defense information.'' This change was made to use the defined
term in part 95 rather than ``defense information'' as used in Sec.
50.33(j). The usage in Sec. 50.33(j) dates back to the Atomic Energy
Commission amendment of that section on January 19, 1956 (21 FR 355,
357) and was not changed with the issuance of part 95 (45 FR 14476;
March 5, 1980) after the establishment of the NRC and the 1975
reissuance of the former Atomic Energy Commission regulations. The
revised terminology also aligns with its usage in Sec. 53.1115.
Proposed Sec. 53.1112 would address environmental conditions and
is equivalent to Sec. 50.36b. Proposed Sec. 53.1115 would address
requirements for agreements limiting access to classified information
and is equivalent to Sec. 50.37. Proposed Sec. 53.1118 would address
ineligibility of certain applicants and is equivalent to Sec. 50.38.
Proposed Sec. 53.1120 would address exceptions and exemptions from
licensing requirements for Department of Defense and DOE facilities and
is equivalent to Sec. 50.11. Proposed Sec. 53.1121 would address
public inspection of applications and is equivalent to Sec. 50.39.
Proposed Sec. 53.1124 would address the relationship between the
various licenses, certifications, and approvals
[[Page 86944]]
provided in this subpart, and the requirements are equivalent to a
number of similar provisions in parts 50 and 52 including Sec. Sec.
50.10, 52.13, 52.43, 52.73, 52.133, and 52.153. New provisions are
provided in Sec. 53.1124(c) and (d), that would allow an application
for either a standard design approval or a standard DC under part 53 to
reference applicable licensing-basis information that supported
issuance of an OL or COL under part 53. These provisions are being
proposed to offer additional flexibility beyond what is currently
allowed under parts 50 or 52 for an applicant who may wish to license a
first-of-a-kind reactor for operation prior to seeking generic approval
or certification of the standard design.
Proposed Sec. 53.1124(e) would address the limitations that a
manufactured reactor may only be transported to a site with a COL and
is equivalent to Sec. 52.153. Proposed Sec. 53.1130 would address
LWAs and is equivalent to Sec. 50.10.
Proposed Sec. Sec. 53.1140 through 53.1188 would govern the
content of ESP applications. Proposed Sec. 53.1140 is equivalent to
Sec. 52.12. Proposed Sec. 53.1143 would address filing of
applications and is equivalent to Sec. 52.15. Proposed Sec. 53.1144
would address general information requirements for the content of
applications and is equivalent to Sec. 52.16.
Proposed Sec. 53.1146 would specify requirements for the technical
contents of applications and is equivalent to Sec. 52.17. Proposed
Sec. 53.1146(b)(2) provides applicants for ESPs a regulatory option to
propose major features of the emergency plans or complete and
integrated emergency plans in accordance with either the requirements
in Sec. 50.160 of this chapter, or the requirements in appendix E to
part 50 of this chapter and Sec. 50.47(b) of this chapter, as
applicable.
Proposed Sec. 53.1149 would address standards for review of ESP
applications and administrative review of applications, including
hearings, and is equivalent to Sec. Sec. 52.18 and 52.21. Proposed
Sec. 53.1155 would address referral to the ACRS and is equivalent to
Sec. 52.23. Proposed Sec. 53.1158 would address issuance of ESPs and
is equivalent to Sec. 52.24. Proposed Sec. 53.1161 would address the
extent of activities permitted and is equivalent to Sec. 52.25.
Proposed Sec. 53.1164 would address the duration of an ESP and is
equivalent to Sec. 52.26. Proposed Sec. 53.1167 would address
provisions for requesting a LWA after issuance of an ESP and is
equivalent to Sec. 52.27. Proposed Sec. 53.1170 would address
transfers of ESPs and is equivalent to Sec. 52.28. Proposed Sec.
53.1173 would address applications for ESP renewals and is equivalent
to Sec. 52.29. Proposed Sec. 53.1176 would address criteria for
renewal of an ESP and is equivalent to Sec. 52.31. Proposed Sec.
53.1179 would address the duration of an ESP renewal and is equivalent
to Sec. 52.33. Proposed Sec. 53.1182 would address the use of a site
for purposes other than those described in the permit and is equivalent
to Sec. 52.35. Proposed Sec. 53.1188 would address finality of ESP
determinations and is equivalent to Sec. 52.39.
Proposed Sec. Sec. 53.1200 through 53.1221 would govern the
contents of standard design approval applications. Proposed Sec.
53.1200 is equivalent to Sec. 52.131. Proposed Sec. 53.1203 would
address filing of applications and is equivalent to Sec. 52.135.
Proposed Sec. 53.1206 would address general information requirements
for the content of applications and is equivalent to Sec. 52.136.
Proposed Sec. 53.1209 would address requirements for the technical
content of applications and is largely equivalent to Sec. 52.137. In
proposed Sec. 53.1209(a), the NRC proposes text that expands the
discussion of ``major portion'' standard design approvals. Additional
discussion regarding standard design approvals for a major portion of a
standard design can be found in the NRC's ``A Regulatory Review Roadmap
for Non-Light Water Reactors,'' which considers the Nuclear Innovation
Alliance report ``Clarifying `Major Portions' of a Reactor Design in
Support of a Standard Design Approval.'' Proposed Sec. 53.1209(b)
outlines the required content of the Final Safety Analysis Report
(FSAR). Proposed requirements in Sec. 53.1209(b)(2) for portions of
the application addressing design information state that the
application must include design information equivalent to that required
for a standard DC. This reference to the pertinent DC requirements
(specifically, those in proposed Sec. 53.1239(a)(2) through (27)) is
an efficiency that would prevent the need to repeat many of the same
requirements for the content of a standard design approval application.
Proposed Sec. 53.1210 would address requirements for the content
of a standard design approval application other than the FSAR. Proposed
Sec. 53.1210(a) would require the inclusion of a description of
availability controls that are not included in the FSAR.
Proposed Sec. 53.1212 would address standards for review of
applications and is equivalent to Sec. 52.139. Proposed Sec. 53.1215
would address referral to the ACRS and is equivalent to Sec. 52.141.
Proposed Sec. 53.1218 would address staff approval of designs and
duration of design approvals and is equivalent to Sec. Sec. 52.143 and
52.147. Proposed Sec. 53.1221 would address finality of standard
design approvals and information requests and is equivalent to Sec.
52.145 with the exception that it extends such finality to a standard
approval referenced in a DC application. Standard design approvals
issued to date under part 52 have been issued during the NRC's review
of the standard DC application and have relied on the same application
content. However, a future scenario could arise where the DC
application is not submitted until after a design approval has been
granted. The NRC would apply the same finality provisions in this
situation as in the situation where a standard design approval is
referenced in a COL application.
There is no equivalent to proposed Sec. 53.1221(d) in part 52 for
standard design approvals. This provision would state that the
Commission will require, before granting a CP, COL, OL, or ML which
references a standard design approval, that engineering documents be
completed and available for audit. A similar provision is included in
part 52 in relation to a standard DC; and the NRC would require that
design and analysis information needed for the Commission to make its
safety determination be complete and available for any application the
NRC is reviewing. Making this explicit provides increased clarity to
future standard design approval applicants under part 53.
Proposed Sec. Sec. 53.1230 through 53.1263 would address standard
DCs. Proposed Sec. 53.1230 would address general provisions for
standard DCs and is equivalent to Sec. 52.41. Proposed Sec. 53.1233
would address filing of applications and is equivalent to Sec. 52.45.
Proposed Sec. 53.1236 would address general information requirements
for the content of applications and is equivalent to Sec. 52.46.
Proposed Sec. 53.1239 would address requirements for the technical
content of applications and is equivalent to Sec. 52.47(a). The
requirements in proposed Sec. 53.1239 have been modified from the
analogous requirements in Sec. 52.47(a) to align with the technical
requirements in proposed part 53.
Proposed Sec. 53.1241 would address requirements for the content
of a standard DC application other than the FSAR and is equivalent to
Sec. 52.47(b) and (d).
[[Page 86945]]
Proposed Sec. 53.1242 would address review of applications and is
equivalent to Sec. Sec. 52.48 and 52.51. Proposed Sec. 53.1242(c)
would include a provision that would allow a DC applicant to reference
applicable licensing-basis information for an OL or COL issued under
part 53. As explained previously, this provision is being proposed to
explicitly allow flexibility for an applicant who may wish to license a
first-of-a-kind reactor for operation prior to seeking certification of
the generic reactor design. For NRC findings on a reactor design in an
OL or COL proceeding, this proposal would provide finality in a
subsequent DC application that references information on the OL or COL
proceeding's docket. This finality accorded to the OL or COL findings
would bind the NRC staff and the ACRS but would not bind members of the
public or the Commission. (To the extent an Atomic Safety and Licensing
Board (ASLB) might have a role in a DC rulemaking, the OL or COL
findings would not bind the ASLB either.) Specifically, members of the
public would have the opportunity to comment on a proposed DC rule
under well-established NRC practice. The rationale for binding the NRC
staff and ACRS is similar to the rationale for a COL applicant
referencing a standard design approval under part 52.
Proposed Sec. 53.1245 would address referral to the ACRS and is
equivalent to Sec. 52.53. Proposed Sec. 53.1248 would address
issuance of standard DCs and is equivalent to Sec. 52.54. Proposed
Sec. 53.1251 would address duration of certifications and is
equivalent to Sec. 52.55(c). Proposed Sec. 53.1254 would address
application for renewal and is equivalent to Sec. 52.57. Proposed
Sec. 53.1257 would address criteria for renewal and is equivalent to
Sec. 52.59. Proposed Sec. 53.1260 would address duration of renewals
and is equivalent to Sec. 52.61. Proposed Sec. 53.1263 would address
finality of standard DCs and is equivalent to Sec. 52.63.
Proposed Sec. Sec. 53.1270 through 53.1291 would address MLs
covering manufacturing activities at one or more licensee facilities.
Proposed Sec. 53.1270 would address the scope of these sections and is
equivalent to Sec. 52.151.
Proposed Sec. 53.1273 would address filing of applications for an
ML and is equivalent to Sec. 52.155(a).
Proposed Sec. 53.1276 would address general information
requirements for the content of ML applications and is equivalent to
Sec. 52.156, with one exception. Proposed Sec. 53.1276 would require
each application for an ML to also include the information required by
Sec. 53.1109(e). This information includes the type of license applied
for, the use to which the facility will be put, the period of time for
which the license is sought, and a list of other licenses, except
operator's licenses, issued or applied for in connection with the
proposed facility to address the potential variations in how MLs might
be formulated under the proposed part 53.
Proposed Sec. 53.1279 would address requirements for the technical
content of applications for MLs to be included in the FSAR and is
equivalent to Sec. 52.157. In addition, the requirements in proposed
Sec. 53.1279(a) and (b) have been modified from the analogous
requirements in Sec. 52.157 to align with the technical requirements
in proposed part 53. Proposed Sec. 53.1279(a)(2) outlines the required
content of the application addressing design information and states
that the application must include design information equivalent to that
required for a standard DC. This reference to the pertinent DC
requirements is an efficiency that would prevent the need to repeat the
same requirements for the content of an ML application.
Proposed Sec. 53.1279(c) would provide application requirements
related to the deployment of the completed manufactured reactor.
Proposed Sec. 53.1279(c)(1) would require inclusion of information
related to the procedures governing the preparation of the manufactured
reactor for shipping to the site where it is to be operated, the
conduct of shipping, and the verification of the condition of the
shipped items upon receipt at the site. Proposed Sec. 53.1279(c)(2)
would require that the application include information on the
interaction of the design, manufacture, and installation of a
manufactured reactor within the applicant's organization and the manner
by which the applicant will ensure close integration between the
designer, contractors, and any licensee of a facility in which the
manufactured reactor is to be installed. Finally, proposed Sec.
53.1279(c)(3) would require that the application include a description
of the measures used for the control of interfaces between the holder
of the ML and the holder of the COL for the commercial nuclear plant at
which the manufactured reactor is to be installed. This information is
necessary for the NRC to determine whether the applicant would have
appropriate controls in place to ensure coordination between parties
involved in the design, manufacture, and eventual operation of any
reactor manufactured under an ML.
Proposed Sec. 53.1279(d) would include additional requirements for
application content for applicants seeking an ML for manufactured
reactors that will be fueled at the factory under a 10 CFR part 70
license, consistent with the requirements in Sec. 53.620(d). These
provisions would require the application to include information related
to loading fuel and the required independent physical mechanisms to
prevent criticality and to otherwise provide assurance that the fueled
manufactured reactor can be successfully transported, installed, and
operated at a site for which the Commission has issued a COL that
authorizes construction and operation of a commercial nuclear plant
using the manufactured reactor.
Proposed Sec. 53.1282 would provide requirements for other
application content for MLs and is equivalent to Sec. 52.158. Proposed
Sec. 53.1282(a)(1) would provide requirements to include in the ML
application the ITAAC within the scope of the ML that the COL holder
referencing the ML must satisfy. Proposed Sec. 53.1282(a)(2) would
require that the ITAAC from a referenced standard design apply to the
portions of the ML design within the scope of the referenced standard
design. Proposed Sec. 53.1282(a)(3) would state that the COL
application may include a notification that required referenced
standard DC ITAAC have been satisfied at the manufacturing facility.
Proposed Sec. 53.1282(b) would require an ML application to
include an environmental report and, consistent with existing
requirements, proposed Sec. 53.1282(b)(2) would note that if the ML
application references a standard DC, the environmental report need not
contain a discussion of severe accident mitigation design alternatives
for the manufactured reactor as used in a commercial nuclear plant.
Proposed Sec. 53.1285 would provide standards for review of
applications and administrative review of applications for MLs,
including hearings, and is equivalent to Sec. Sec. 52.159 and 52.163.
Proposed Sec. 53.1286 would address referral of applications to
the ACRS and is equivalent to Sec. 52.165. Proposed Sec. 53.1287
would address issuance of an ML and is equivalent to Sec. 52.167.
Proposed Sec. 53.1288 would address finality of MLs and is
equivalent to Sec. 52.171. Proposed Sec. 53.1291 would address the
duration of MLs and is equivalent to Sec. 52.173. Proposed Sec.
53.1293 would address the transfer of MLs and is equivalent to Sec.
52.175. Proposed Sec. 53.1295 would address the renewal of MLs and is
equivalent to Sec. Sec. 52.177, 52.179 and 52.181, with a minor
exception. Proposed Sec. 53.1295(a)(3) would state that an ML
[[Page 86946]]
for which a timely application for renewal has been filed remains in
effect until the Commission has made a final determination on the
renewal application, provided, however, that the holder of an ML may
not begin manufacture of a manufactured reactor less than six months
before the expiration of the license. The proposed 6-month time frame
for this provision is changed from the 3-year period in the equivalent
provision in part 52 because future reactor applicants may present
smaller, simpler designs, to include micro-reactor designs, in ML
applications than those that were envisioned when the existing
requirements were written. A 6-month time frame for this provision
would provide greater flexibility for ML holders related to
manufactured reactors being produced when the ML expires.
Proposed Sec. Sec. 53.1300 through 53.1348 would address licensing
requirements for CPs. Proposed Sec. 53.1300 would set out general
requirements for CPs and is equivalent to Sec. 50.23. Proposed Sec.
53.1306 would address the general information requirements for the
content of applications for CPs and is equivalent to Sec. 50.33(f) and
(h).
Proposed Sec. 53.1309 would address requirements for the technical
content of applications for CPs and includes the requirement to submit
a Preliminary Safety Analysis Report (PSAR) that describes the facility
and presents a preliminary safety analysis of the facility as a whole.
This is in contrast to an OL application which is required to include
an FSAR that describes the facility and presents a final safety
analysis of the facility as a whole. Proposed Sec. 53.1309 is
equivalent to Sec. 52.17(a)(1)(iv) through (a)(1)(x) and 52.17(b),
with two exceptions. First, proposed Sec. 53.1309 would replace the
analysis of the dose criteria required by Sec. 52.17(a)(1)(ix) with
analysis to demonstrate compliance with the safety criteria defined in
Sec. Sec. 53.210 and 53.220. Second, proposed Sec. 53.1309(a)(2)
would add a requirement for a CP application to include several
categories of detailed design information, although Sec.
53.1309(a)(2)(ii) would allow certain relaxations of this requirement
in view of aspects of a design that may not yet be fully developed.
Section 53.1309 would reference the requirements for the content of an
ESP application to address application requirements related to siting
and would reference the requirements for the content of a DC
application to address application requirements related to design of
the commercial nuclear plant. Proposed Sec. 53.1309(a)(2)(ii) would
address the treatment of preliminary design information and notes that
information provided in the application may include some aspects of the
design that are not fully developed. This provision would require that
the completed design, including any changes during construction, be
described in the FSAR in an application for an OL. This would include
the requirement for a description of the PRA required by Sec.
53.450(a) and its results. Probabilistic risk assessments developed for
commercial nuclear plants prior to construction would be based on the
design and other information available at the time of the CP
application. PRAs performed in early design stages or prior to
construction may be inherently less detailed and may include projected
information that will be subsequently verified or revised when the
plant is built. Proposed Sec. 53.1309(a)(4) would address preliminary
description of the plans for coping with emergencies.
Proposed Sec. 53.1312 would address other application content for
CPs. Proposed Sec. 53.1312(a)(1) is equivalent to Sec. 52.80(b) but
is adapted for a CP application. Proposed Sec. 53.1312(a)(2) is
equivalent to Sec. 52.80(c) but is adapted for a CP application.
Proposed Sec. 53.1312(b)(1) is equivalent to Sec. 52.79(b), (c), and
(d) but is adapted for a CP application. Section 53.1312(b)(2) is
equivalent to portions of Sec. Sec. 52.63(b)(1), 52.79(b)(1) through
(b)(3), (c), and (d)(1) and (d)(3), 52.80, and 52.93(b), but is adapted
for a CP application. Guidance for equivalent requirements in parts 50
and 52 is also addressed in RG 1.206, ``Applications for Nuclear Power
Plants,'' Revision 1, section C.1.7.
Proposed Sec. 53.1315 would address standards for review of
applications and administrative review of applications, including
hearings, and is equivalent to Sec. Sec. 52.81 and 52.85, but is
adapted for a CP application.
Proposed Sec. 53.1318 would address finality of NRC approvals,
licenses, and certifications referenced in a CP application and is
equivalent to Sec. 52.83(a) but is adapted for a CP application.
Proposed Sec. 53.1324 would address referral to the ACRS and is
equivalent to Sec. 50.58(a) and to Sec. 52.87 but is adapted for a CP
application.
Proposed Sec. 53.1327 would address authorization to conduct LWA
activities and is equivalent to Sec. 52.91 but is adapted for a CP
application. Proposed Sec. 53.1327(a) is equivalent to Sec. 52.91(a)
but is adapted for a CP application. Proposed Sec. 53.1327(b) is
equivalent to Sec. 52.91(b) but is adapted for a CP application.
Proposed Sec. 53.1330 would address exemptions, departures, and
variances for CP applicants.
Proposed Sec. 53.1333 would address issuance of CPs. Proposed
Sec. 53.1333(a) is equivalent to Sec. 50.35(a). Proposed Sec.
53.1333(b) is equivalent to Sec. 50.35(b) and to Sec. 52.97(c) but is
adapted for a CP application. Proposed Sec. 53.1336 would address the
effect of CPs and is equivalent to Sec. 50.35(b). Proposed Sec.
53.1342 would address the duration of CPs. Proposed Sec. 53.1342(a) is
equivalent to Sec. 50.55(a). Proposed Sec. 53.1342(b) is equivalent
to Sec. 50.55(b). Proposed Sec. 53.1345 would address the transfer,
assignment, and disposal of CPs and is equivalent to Sec. 50.80.
Proposed Sec. 53.1348 would address the termination of CPs and is
equivalent to Sec. Sec. 52.3(b)(8) and 52.110(a)(1) but is adapted for
a CP application.
Proposed Sec. Sec. 53.1360 through 53.1405 address requirements
for OLs.
Proposed Sec. 53.1366 would address requirements for the general
content of applications for OLs. It would refer to general content
requirements in proposed Sec. 53.1109 and would require supplemental
information. Proposed Sec. 53.1366(a) is equivalent to Sec. 50.33(f).
Proposed Sec. 53.1366(b) is equivalent to Sec. 50.33(k).
Proposed Sec. 53.1369 would provide requirements for the technical
content of applications for OLs to be included in the FSAR and is
equivalent to Sec. 50.34(b) but has been modified to align with the
technical requirements in part 53. It would require that the FSAR
include and, as needed, update information provided in the PSAR that
was submitted and reviewed to support the associated CP application.
Similar to the proposed requirements for the content of CP
applications, proposed Sec. 53.1369(a) would reference the
requirements for the content of an ESP application to address
application requirements related to the site. Section 53.1369(b) would
reference the requirements for the content of a DC application to
address some of the application requirements related to design of the
commercial nuclear plant.
Proposed Sec. 53.1369(c) is equivalent to Sec. 50.34(b)(7).
Proposed Sec. 53.1369(d) would require a description of the Integrity
Assessment Program that would be required by proposed Sec. 53.870.
Proposed Sec. 53.1369(e) is equivalent to Sec. 50.34(e). Proposed
Sec. 53.1369(g) would provide requirements for OL application content
to support proposed Sec. 53.730 related to the role of personnel in
the operation of the commercial nuclear plant and is adapted from
requirements in part 55 and Sec. 50.34(f). Likewise, proposed Sec.
53.1369(h) would provide
[[Page 86947]]
requirements for OL application content related to training programs to
support proposed Sec. Sec. 53.730(g) and 53.830 and includes
requirements equivalent to Sec. 50.34(b)(8), Sec. 52.79(a)(33), and
part 55. Proposed Sec. 53.1369(i) would provide requirements for OL
application content related to emergency plans to support proposed
Sec. 53.855 and is equivalent to Sec. 50.34(b)(6)(v).
Proposed Sec. 53.1369(j) would provide requirements for OL
application content related to the applicant's organizational structure
and is equivalent to Sec. 50.34(b)(6)(i). Proposed Sec. 53.1369(k)
would provide requirements for OL application content related to the
applicant's proposed maintenance program to support proposed Sec.
53.715 and is equivalent to Sec. 50.34(b)(6)(iv). Proposed Sec.
53.1369(l) would provide requirements for OL application content
related to the applicant's quality assurance program to support
proposed Sec. 53.865 and is equivalent to Sec. 50.34(b)(6)(ii).
Proposed Sec. 53.1369(m) would provide requirements for OL application
content related to the applicant's proposed radiation protection
program to support proposed Sec. 53.850 and is equivalent to Sec.
50.34(b)(3).
Proposed Sec. 53.1369(n) through (p) would provide requirements
for OL application content related to the applicant's proposed physical
security program to support proposed Sec. 53.860(a) and are equivalent
to Sec. 50.34(c) and (d). Proposed Sec. 53.1369(q) would provide
requirements for OL application content related to the applicant's
proposed cybersecurity plan to support proposed Sec. 53.860(d) and is
equivalent to Sec. Sec. 52.79(a)(36)(iv) and 73.54. Proposed Sec.
53.1369(r) would provide requirements for OL application content
related to the implementation of proposed security, safeguards, and
cybersecurity plans to support proposed Sec. 53.860 and is equivalent
to Sec. 52.79(a)(35)(ii) and 52.79(a)(36)(iv) and (v).
Proposed Sec. 53.1369(s) would provide requirements for OL
application content related to the applicant's proposed fire protection
program to support proposed Sec. 53.875 and is equivalent to Sec.
52.79(a)(40). Proposed Sec. 53.1369(t) would provide requirements for
OL application content related to the applicant's proposed ISI and IST
program to support proposed Sec. 53.880 and is equivalent to part of
Sec. 52.79(a)(11). Proposed Sec. 53.1369(w) would provide
requirements for OL application content related to the applicant's
general employee training program to support proposed Sec. 53.830 and
is equivalent to Sec. 52.79(a)(33). Proposed Sec. 53.1369(x) would
provide requirements for OL application content related to the
applicant's FFD program to support part 26 and is equivalent to Sec.
52.79(a)(44). Proposed Sec. 53.1369(y) would provide requirements for
OL applicant's programs to demonstrate that any safety questions
identified at the CP stage have been resolved and is equivalent to
Sec. 50.34(b)(5). Proposed Sec. 53.1369(z) would provide requirements
for OL applicants to describe how the performance of each safety design
feature has been demonstrated capable of fulfilling functional design
criteria considering interdependent effects through either analysis,
appropriate test programs, prototype testing, operating experience, or
a combination thereof to support proposed Sec. 53.440(a). It is
largely equivalent to Sec. Sec. 50.34(b)(5) and 50.43(e). Proposed
Sec. 53.1369(aa) would provide requirements for OL application content
related to the applicant's proposed TS to support proposed Sec.
53.710(a) and is equivalent to Sec. 50.34(b)(6)(vi).
Proposed Sec. 53.1372 would address requirements for the content
of OL applications other than the FSAR. Proposed Sec. 53.1372(a) would
require submission of an environmental report and is equivalent to
Sec. 50.30(f) and Sec. 51.53(b). Proposed Sec. 53.1372(b) does not
have a direct parallel in parts 50 and 52 and would require the
inclusion of a description of availability controls that are not
included in the FSAR to support proposed Sec. 53.710(b).
Proposed Sec. 53.1375 would address standards for review of OL
applications and the administrative review of applications, including
hearings, and is equivalent to Sec. Sec. 52.81 and 52.85, except that
the NRC has omitted 10 CFR part 54, ``Requirements for Renewal of
Operating Licenses for Nuclear Power Plants,'' from the list of
standards in the proposed Sec. 53.1375(a). Proposed part 53 does not
include detailed requirements related to renewal of licenses, although
a general provision and possible placeholder for future requirements
has been included as proposed Sec. 53.1595. The NRC will decide after
the part 53 final rule is published whether this future section will be
retained in part 53 to address license renewal or whether the agency
will take another approach to address license renewal for part 53
licensees, such as amending part 54 to address part 53 licensees.
Proposed Sec. 53.1381 would address referral to the ACRS and is
equivalent to Sec. Sec. 50.58 and 52.87. Proposed Sec. 53.1384 would
address exemptions, departures, and variances for OL applicants.
Section 53.1384(a) is equivalent to Sec. 52.93 but is adapted for OLs.
Proposed Sec. 53.1384(b) is equivalent to Sec. Sec. 52.39(d) (with
respect to ESPs) and 52.93 but is adapted for OLs.
Proposed Sec. 53.1387 would address issuance of OLs. The proposed
introductory paragraph is equivalent to Sec. 50.56. Proposed Sec.
53.1387(a)(1)(i) is equivalent to Sec. Sec. 50.50 and 50.57(a)(1).
Proposed Sec. 53.1387(a)(1)(ii) is equivalent to Sec. 50.50. Proposed
Sec. 53.1387(a)(1)(iii) is equivalent to Sec. 50.57(a)(2). Section
53.1387(a)(1)(iv) is equivalent to Sec. 50.57(a)(3). Proposed Sec.
53.1387(a)(1)(v) is equivalent to Sec. 50.57(a)(4). Proposed Sec.
53.1387(a)(1)(vi) is equivalent to Sec. 50.57(a)(6). Proposed Sec.
53.1387(a)(1)(vii) is equivalent to Sec. 50.57(a)(5). Proposed Sec.
53.1387(a)(1)(viii) is equivalent to Sec. 52.97(a)(1)(vi) but is
adapted for OLs. Proposed Sec. 53.1387(c) is equivalent to Sec.
50.57(b). Proposed Sec. 53.1387(d) is equivalent to Sec. Sec.
50.36(b) and 50.50.
Proposed Sec. 53.1390 would address backfitting of OLs and is
equivalent to Sec. 52.98(a) but adapted for an OL application.
Proposed Sec. 53.1396 would address duration of an OL and is
equivalent to Sec. 50.51(a) and Sec. 52.104. Proposed Sec. 53.1399
would address transfer, assignment, and other disposition of an OL and
is equivalent to Sec. 50.80. Proposed Sec. 53.1402 would address
applications for renewal of an OL and refers to proposed Sec. 53.1595.
Proposed Sec. 53.1405 would address continuation of an OL and is
equivalent to Sec. 52.109 but is adapted to address an OL.
Proposed Sec. Sec. 53.1410 through 53.1461 would address
requirements for COLs. Proposed Sec. 53.1410 is equivalent to Sec.
52.71. Proposed Sec. 53.1413 would address general information
requirements for the content of applications for COLs and is equivalent
to Sec. 52.77, which references Sec. 50.33. Most of the provisions
from Sec. 50.33 are restated in proposed Sec. 53.1109. Some
requirements in Sec. 50.33 related to financial qualifications and
construction timelines are addressed in other sections of part 53.
Proposed Sec. 53.1416 would address the technical content to be
included in an FSAR for an application for a COL and is equivalent to
Sec. 52.79 except as modified to reflect the technical requirements in
part 53 and with one addition. Proposed Sec. 53.1416 includes the
statement that the Commission will require, before issuance of a COL,
that engineering documents, such as analyses, drawings, procurement
specifications, or construction and installation specifications, be
completed and available for audit if the more
[[Page 86948]]
detailed information is necessary for the Commission to verify the
information in the application and make its safety determination. This
statement is equivalent to DC application requirements in Sec. 52.47
and is included in proposed Sec. 53.1416 for clarity.
Similar to the proposed requirements for the content of OL
applications, proposed Sec. 53.1416(a)(1) would reference the
requirements for the content of an ESP application to address
application requirements related to siting. Section 53.1416(a)(2) would
reference the requirements for the content of a DC application to
address some of the application requirements related to design of the
commercial nuclear plant. The remaining items under proposed Sec.
53.1416(a) are likewise similar to the required content for OL
applications under proposed Sec. 53.1369(a). Proposed Sec. 53.1416(b)
would require COL applicants to provide a report documenting the
resolution of any safety questions for SSCs for which research and
development was necessary to confirm the adequacy of their design and
is equivalent to Sec. 50.34(b)(5). Proposed Sec. 53.1416(c) would
provide requirements for COL applicants to describe how the performance
of each safety design feature has been demonstrated capable of
fulfilling functional design criteria considering interdependent
effects through either analysis, appropriate test programs, prototype
testing, operating experience, or a combination thereof to support
proposed Sec. 53.440(a). It is largely equivalent to Sec. Sec.
52.79(a)(24) and 50.43(e). Proposed Sec. 53.1416(d) would address the
content of COL applications referencing an ESP. Proposed Sec.
53.1416(e) would address the content of COL applications referencing a
standard design approval. Proposed Sec. 53.1416(f) would address the
content of COL applications referencing a standard DC. Proposed Sec.
53.1416(g) would address the content of COL applications referencing an
ML.
Proposed Sec. 53.1419 would address other application content for
COLs and is equivalent to Sec. 52.80. Proposed Sec. 53.1419(a)(2) is
new and would require the inclusion of a description of availability
controls that are not required to be included in the FSAR.
Proposed Sec. 53.1422 would address standards for review of
applications and the administrative review of applications, including
hearings, and is equivalent to Sec. Sec. 52.81 and 52.85. The NRC has
removed part 54 from the list of standards in proposed Sec.
53.1422(a). Proposed part 53 does not include requirements related to
renewal of licenses, in relation to proposed Sec. Sec. 53.1422 and
53.1595.
Proposed Sec. 53.1425 would address the finality of NRC approvals
referenced in a COL application and is equivalent to Sec. 52.83(a).
Proposed Sec. 53.1431 would address the referral of COL applications
to the ACRS for review and is equivalent to Sec. 52.87. Proposed Sec.
53.1434 would address the authorization to conduct LWA activities and
is equivalent to Sec. 52.91. Proposed Sec. 53.1437 would address
exemptions, departures, and variances and is equivalent to Sec. 52.93.
Proposed Sec. 53.1440 would address issuance of COLs and is equivalent
to Sec. 52.97. Proposed Sec. 53.1443 would address finality of COLs
and is equivalent to Sec. 52.98.
Proposed Sec. 53.1449 would address inspection during construction
and is equivalent to Sec. 52.99. Proposed Sec. 53.1452 would address
operation under a COL and is equivalent to Sec. 52.103. Paragraph (a)
of proposed Sec. 53.1452 would include footnotes to provide that, for
licensees installing fueled manufactured reactors under a COL, (1) the
COL holder would notify the NRC of its scheduled date for initiating
the physical removal of any one of the independent physical mechanisms
to prevent criticality required under Sec. 53.620(d)(1) rather than
its scheduled date for the initial loading of fuel, and (2) the NRC
would time its publication of the notice of intended operation based on
the COL holder's schedule for initiating the physical removal of any
one of the independent physical mechanisms to prevent criticality
required under Sec. 53.620(d)(1) rather than the COL holder's
scheduled date for the initial loading of fuel. These footnotes are
consistent with the provisions of proposed Sec. 53.620(d)(1)(iv),
which would state that, upon initiating the physical removal of any one
of the independent physical mechanisms to prevent criticality in the
manufactured reactor's place of operation, the fueled manufactured
reactor has commenced operation. For reactors without the independent
physical mechanisms to preclude criticality under proposed Sec.
53.620(d)(1), operation begins with initial fuel load. In both cases,
removal of the physical features to prevent criticality (for reactors
with such features) and initial fuel load (for reactors without such
features) put a fully constructed utilization facility in a position to
sustain a nuclear chain reaction, and in both cases, the utilization
facility cannot sustain a nuclear chain reaction (for lack of
sufficient reactivity) until the action takes place. Therefore, the NRC
proposes that initiating the physical removal of any one of the
independent physical mechanisms to prevent criticality is the best
analogue to initial loading of fuel for reactors without such features.
The proposed footnote in Sec. 53.1452(a) regarding timing of the
notice of intended operation for fueled manufactured reactors with
independent physical mechanisms to prevent criticality also addresses
the requirements of Section 189a.(1)(B)(i) of the Act. This section
requires, in part, that ``[n]ot less than 180 days before the date
scheduled for initial loading of fuel into a plant by a licensee that
has been issued a combined construction permit and operating license
under section 185b., the Commission shall publish in the Federal
Register notice of intended operation.'' That section further requires
that this notice provide a 60-day period in which to request a hearing
``on whether the facility as constructed complies, or on completion
will comply, with the acceptance criteria of the license.'' In the case
where a fueled manufactured reactor arrives at the site where it is to
be operated by a COL holder, the manufacturer would have loaded fuel at
the factory under its part 70 license. Therefore, at the site of
operation, there would not be ``initial loading of fuel into a plant by
a licensee that has been issued a combined construction permit and
operating license'' (emphasis added). Under a literal reading of the
entry condition in Act Section 189a.(1)(B)(i), this situation would not
trigger its requirements. However, the purpose of the provision is to
offer the hearing opportunity at least 180 days prior to when the fuel
is loaded and ready for use at its authorized location. It would be
contrary to that purpose if, in this situation, the Commission did not
publish the notice of intended operation and opportunity for the public
to request a hearing on conformance with the acceptance criteria in the
COL for the site of operation. To fulfill the underlying purpose of the
law, the NRC proposes to time the notice of intended operation based on
the COL holder's schedule for initiating the physical removal of any
one of the independent physical mechanisms to prevent criticality
required under Sec. 53.620(d)(1). This action by the COL holder would
be the best analogue to initial fuel load by the COL holder for the
reasons stated previously. This analogue is adopted in other sections
of the proposed part 53 and related sections in parts 50 and 73 that
use initial fuel loading to identify
[[Page 86949]]
a transition point for the applicability of regulatory requirements. To
address the possible loading of fuel into a manufactured reactor for
subsequent transport to and use at a commercial nuclear plant, multiple
sections that determine the applicability of regulations have been
drafted or revised to allow for either initial fuel load or initiating
the physical removal of any one of the independent physical mechanisms
to prevent criticality required under Sec. 53.620(d)(1) for a fueled
manufactured reactor to determine the applicability of the requirement,
as appropriate.
Proposed Sec. 53.1455 would address duration of COL and is
equivalent to Sec. 52.104. Proposed Sec. 53.1456 would address the
transfer of a COL and is equivalent to Sec. 52.105. Proposed Sec.
53.1458 would address application for renewal and is equivalent to
Sec. 52.107. Proposed Sec. 53.1461 would address continuation of COL
and is equivalent to Sec. 52.109.
Proposed Sec. 53.1470 would address standardization of commercial
nuclear plant designs and licenses to construct and operate commercial
power reactors of identical design at multiple sites and is equivalent
to appendix N of part 52. This section would set out the particular
requirements and provisions applicable to situations in which
applications for CPs and subsequent OLs, or COLs, under this part are
filed by one or more applicants for licenses to construct and operate
nuclear power reactors of identical design (``common design'') to be
located at multiple sites. Additional information related to this
proposed section is provided in the final rule to revise part 52 (72 FR
49352; August 28, 2007).
Subpart I--Maintaining and Revising Licensing-Basis Information
Part 53 would establish requirements for the maintenance of
licensing-basis information in subpart I.
Section 53.1500 would describe the purpose of the subpart in terms
of the definition of licensing-basis information in subpart A. Subpart
I would be closely tied to the requirements in subpart H, which would
provide the requirements for contents of applications for the various
types of licenses issued under part 53. Subpart I would generally be
organized into sections dealing with: (1) licensing-basis information
that licensees are not authorized to change without NRC approval (e.g.,
licenses, regulations); and (2) licensing-basis documents that
licensees may change provided specified criteria are satisfied (e.g.,
FSAR, program descriptions). The subpart would also capture certain
general conditions on licenses and changes to the licenses related to
the transfer and termination of licenses.
Section 53.1502 would define specific terms and conditions of
licenses. These terms and conditions would be equivalent to the
regulations in: (1) Sec. 50.54(h) stating that each license is subject
to the provisions of the Act and requirements issued by the Commission;
(2) Sec. 50.54(s) stating the actions the Commission would take if it
makes a finding that there is not reasonable assurance that adequate
protective measures can and will be taken in the event of a
radiological emergency; (3) Sec. 50.54(aa) stating that each license
is subject to the specified sections of the Federal Water Pollution
Control Act; and (4) Sec. 50.54(dd) stating that a holder of an OL or
COL may take reasonable actions that depart from the license in a
national security emergency.
Section 53.1505(a) would serve as an introduction to and overview
of the sections that follow on changes to licensing-basis information
requiring prior NRC approval, namely the elements of licensing-basis
information defined by licenses, orders, and regulations. The related
sections within these subparts would primarily deal with the process of
how a licensee requests and the NRC issues an amendment to a license or
issues an order that modifies a license. Another important element of
licensing-basis information that a part 53 licensee would not be able
to change or deviate from without NRC approval would be the NRC
regulations themselves. Section 53.1505(b) would refer to Sec. 53.080
in subpart A that would provide the criteria for a licensee or other
party to satisfy when requesting an exemption from NRC regulations.
Section 53.1510 would be equivalent to Sec. 50.90 and would
require that a licensee submit an application to request an amendment
to a license. The required assessments that would be included within an
application to amend a license under part 53 would need to address the
safety criteria and analysis requirements of subparts B and C. As with
parts 50 and 52, licensees would be required to include in their
applications to amend a license an analysis of whether the amendment
involves no significant hazards consideration using the standards in
Sec. 53.1520, which would be equivalent to the standards in Sec.
50.92. Although this rulemaking provided an opportunity to revise the
terminology related to no significant hazards consideration
determinations, which dates to the early 1960s when applications were
supported by final hazard summary reports, the NRC is proposing to
maintain the same terminology used in part 50 to minimize the need for
associated changes in other regulations, guidance, and public notices.
Section 53.1515 would establish requirements for public notices and
state consultations associated with the NRC's processing of a license
amendment request. This section would be equivalent to Sec. 50.91 for
the NRC's processes related to applications to amend an OL or COL.
Section 50.91(b) stipulates that the Commission will make available to
the licensee the name of the appropriate State official designated to
receive such amendments. While the Commission intends to continue
following this practice, the Commission has not included this
administrative matter in proposed part 53. Proposed Sec. 53.1515(b)(3)
contains some modifications compared to Sec. 50.91(b)(3) for clarity;
these revisions are not intended to revise the substance of the
provisions in part 53 compared to part 50.
Section 53.1520 would be based on Sec. 50.92. The section would
continue to use the criteria in Sec. 50.92 for determining that a
proposed amendment involves no significant hazards consideration.
Although more specific terms such as event sequence are used throughout
part 53, Sec. 53.1520 would use the term ``accident'' to maintain
consistency with the long history of making no significant hazards
consideration determinations under part 50.
Section 53.1525 would provide requirements for holders of an OL or
COL requesting to revise information from a DC rule that was referenced
in the initial license application and included in or incorporated by
reference into the facility FSAR. In keeping with the current
requirements in part 52, the portion of the part 53 facility licensing-
basis information obtained from the certified design would be divided
into two categories. The most significant design information and the
ITAAC would be certified by rule and designated as ``certification
information.'' The remaining information, which makes up the majority
of the design information approved as part of the DC, would not be
certified by rule and is not considered ``certification information.''
Part 52 refers to these categories of information as Tier 1 and Tier 2
information, respectively, and refers to a change made to that
information on a plant-specific basis as a departure. Under part 52, a
departure from Tier 1 information requires an exemption and,
[[Page 86950]]
for information incorporated into the license, a license amendment.
Part 53 would dispense with the Tier 1 and Tier 2 terminology.
Rather, Sec. 53.1525 would use the term ``certification information''
in place of Tier 1, and a plant-specific departure from the
certification information would require both a request for an exemption
from the associated DC rule and, for information such as ITAAC
incorporated into the license, a license amendment. However, as would
be provided in Sec. 53.1525(c), a plant-specific departure from the
information approved by the NRC as part of the DC rule but which is not
certification information (i.e., Tier 2 information under part 52)
would be assessed using the process and criteria defined in Sec.
53.1550 for changes to a FSAR. An applicant or licensee would need to
identify such a change as a departure from the referenced standard
design in the updated FSAR. The process for making a generic change to
a certified design would be described in the associated section in
subpart H.
Section 53.1530 would not allow the holder of an ML or the holder
of a COL using a manufactured reactor to make changes to the design of
the manufactured reactor without requesting a license amendment from
the NRC. This section would provide the equivalent requirements as
those in Sec. Sec. 52.98 and 52.171.
Section 53.1535 would establish requirements for license amendments
during construction. The section would provide the equivalent options
and requirements for the holders of a CP as those in Sec. 50.35(b).
The regulations would allow but do not require the holder of a CP or
LWA to request an amendment under Sec. 53.1510 if the licensee desires
to obtain NRC approval of a specific design feature or specification.
The requirements for obtaining an amendment to a COL to address changes
during construction would also be provided in Sec. 53.1535. The
proposed process would differ from the current requirements in part 52
by adopting a requirement that would explicitly support a change
process like that described in RG 1.237, ``Guidance for Changes During
Construction for New Nuclear Power Plants Being Constructed Under a
Combined License Referencing a Certified Design Under 10 CFR part 52.''
The proposed regulation would allow the holder of a COL to proceed
at its own risk in making a change during the construction process and
would require that licensee to submit a license amendment request no
later than 45 days from the date the licensee begins to implement the
change or departure requiring NRC approval.
Section 53.1540 would serve as an introduction to the sections that
follow on changes to licensing-basis information that are primarily
under the control of a licensee but for which evaluations are made to
determine if a submittal to the NRC requesting approval would be
required. The section would also include definitions that would be
applicable when using the processes in Sec. Sec. 53.1545 through
53.1565. The definitions would be largely equivalent to those in Sec.
50.59(a) but include some revision to reflect the structure and
terminology in other subparts in part 53. For example, the definition
of ``Change'' in Sec. 53.1540(b) would address a ``design feature or
related functional design criteria'' rather than a ``design function,''
because the former are defined terms in part 53. Similarly, in Sec.
53.1540(b), the phrase ``design basis'' from Sec. 50.59(a)(2) would be
replaced with functional design criteria for SR SSCs.
Section 53.1545 would provide the proposed requirements for
updating of FSARs. While the process-related requirements proposed
under Sec. 53.1545 would be largely the same as those in Sec. 50.71,
the specifics of information to be updated would differ due to the role
of PRA in satisfying the requirements in subparts B and C.
Additionally, the use of the risk-informed approach in subpart C would
result in some but not all PRA information being in the FSAR or another
licensing basis document and therefore a separate PRA update
requirement similar to Sec. 50.71(h) is not included in proposed
subpart I.
Proposed Sec. 53.1239(a)(18) in subpart H and the related
references to this proposed requirement for the holders of OLs and COLs
would require a description of the PRA required by Sec. 53.450(a) and
its results to be included in FSARs. However, guidance documents are
planned to clarify the division of PRA-related information that would
need to be in the FSAR, in other possible licensing basis documents,
and controlled as plant records subject to inspections and audits. At a
minimum, the information from the PRA that would be needed to show
compliance with subpart C would be included in the FSAR (e.g., PRA
summary and analytical results for LBEs). The submittal of voluminous
PRA information was initially required under part 52, but that proved
to be impractical and was revised in the 2007 revision of part 52.
Guidance is being developed to ensure sufficient information is
submitted to the NRC to support the licensing process and the NRC's
regulatory findings under part 53 or similar applications using the LMP
under parts 50 or 52.
The NRC has posed a question in section VI, ``Specific Requests for
Comments,'' of this document that asks about the appropriate level of
detail for PRA-related information in an FSAR and whether other
licensing basis documents might be more appropriate to both provide
information to the NRC and ensure the PRA is maintained and updated as
proposed in subpart C. The program document would provide more detail
than the summaries in the FSAR but still be a much-condensed source of
information in comparison to the documentation of the PRA.
Section 53.1545(a)(3) and (4) would be based on the inclusion of at
least a summary of PRA results and the related margins to safety
criteria in the FSAR and would require updates to that information. The
routine reporting of these margins would also inform application of the
criteria for allowing changes without an amendment in the following
section (Sec. 53.1550) in subpart I.
Section 53.1550 would establish requirements for evaluating changes
to a facility as described in its FSAR. This proposed section would
provide the equivalent of the requirements in Sec. 50.59 for
evaluating changes to an FSAR (as updated) and determining if a license
amendment is required to implement a change to a facility or
procedures. The evaluation criteria proposed in Sec. 53.1550 would
reflect the role of the PRA in the safety analyses under part 53 and
would include several measures related to the changes in plant risk
resulting from a change in the plant design or plant procedures.
Examples would include criteria that rely on the identification of
risk-significant event sequences in accordance with the analysis
requirements of Sec. 53.450; exceeding the LBE evaluation criteria as
defined in Sec. 53.450; the consideration of potential reductions in
margin between the estimated comprehensive risk metrics and associated
risk performance objectives in the safety criteria in Sec. 53.220;
changes to the safety classification of SSCs; and consideration of
reductions in defense in depth.
Section 53.1550 would include a criterion related to a departure
from a method of evaluation used in the safety analyses. The NRC has
not yet developed draft guidance for use in applying proposed Sec.
53.1550 but anticipates that the NRC and stakeholders will assess the
potential need for such guidance and that such guidance would, if
needed, be
[[Page 86951]]
developed as part of ongoing or future activities.
Section 53.1550 would include certain concepts taken from existing
guidance for Sec. 50.59 in the proposed criteria related to DBAs.
Specifically, criterion (iv) for changes made to a method of evaluation
of DBAs under Sec. 53.450(f) would be equivalent to a change in a
method of evaluation under Sec. 50.59, and criterion (viii) on
assessing if a change creates a possibility for an accident of a
different type than previously analyzed in the FSAR would be similar to
the Sec. 50.59 criterion (v). Guidance documents will be prepared to
address the content of applications for PRA-related information under
proposed part 53, and this guidance will also influence how potential
changes in the evaluation of LBEs other than DBAs analyzed under Sec.
53.450(e) are evaluated and reported under the proposed criterion (iv).
Section 53.1550(a)(2)(x) would require evaluating plant changes to
ensure they would not prevent satisfying the design requirements in
Sec. 53.440(j) related to the impact of a large commercial aircraft.
The inclusion of a proposed requirement under Sec. 53.1550 related to
design features for protecting against aircraft impact would reflect
the proposed design requirement in subpart C and related proposed
requirements in subpart H to address the proposed design requirement in
FSARs.
Sections 53.1560 through 53.1565 in subpart I would define the
processes for a licensee to evaluate changes to the program documents
included in the licensing-basis information submitted to the NRC and to
modify such programs without NRC prior approval.
Section 53.1560 would include the proposed requirements for
updating program documents included in licensing-basis information and
would provide the equivalent of FSAR updates for key program documents.
The proposed requirements in these sections would provide a uniform
approach for updating program documents, which correspond to the
programs required under subpart F.
The proposed Sec. 53.1565 would provide a process for licensees to
make changes to program documents included in licensing-basis
information without obtaining prior NRC approval. The proposed
requirements would include several generic criteria that, if not
satisfied, would prompt the need for NRC approval of a change to a
program document. These generic criteria would include whether a change
would comply with TS and NRC regulations. Another proposed criterion
for evaluating changes to program documents would be conforming with
program-specific requirements, including NRC-approved program documents
with more specific criteria for a particular program, regulations,
administrative controls sections of TS, and NRC-approved program
documents.
Proposed Sec. 53.1565(d) would include specific criteria for
evaluating changes to several program documents that have well
established change processes and guidance for licensees under parts 50
and 52. The program documents specifically addressed in the proposed
section would include quality assurance programs that would be
equivalent to Sec. 50.54(a), an emergency preparedness program that
would be equivalent to Sec. 50.54(q), and the security program that
would be equivalent to Sec. 50.54(p).
The proposed Sec. 53.1570 would establish requirements for the
transfer of commercial nuclear plant licenses by providing the
equivalent requirements of Sec. 50.80 for the possible transfer of an
ESP, CP, OL, or COL. Likewise, the proposed Sec. 53.1575 would
establish requirements for the termination of an OL or COL by providing
the equivalent requirements of Sec. 50.82. Other proposed requirements
related to decommissioning and license termination would be included in
subpart G.
Section 53.1580 would establish requirements for information
requests the NRC could send to the various types of licensees and would
provide requirements that would be equivalent to requirements in Sec.
50.54(f). The proposed Sec. 53.1585 would provide the requirements
that would be equivalent to requirements in Sec. 50.100 to address
revocation, suspension, modification of licenses, and approvals for
cause. Section 53.1590 would propose to address backfitting
requirements by providing requirements that would be equivalent to
those in Sec. 50.109.
Proposed Sec. 53.1595 would address license renewals under part 53
with simple statements that licenses may be renewed. This section would
be expanded through future rulemakings to more fully describe or
reference the processes related to requesting and processing
applications to renew ESPs, OLs, and COLs issued under part 53 (if
finalized).
Subpart J--Reporting and Other Administrative Requirements
Part 53 would address various reporting and administrative
requirements in subpart J.
Section 53.1600 would explain the organization of the various
sections within the subpart related to providing unfettered access to
NRC inspectors; maintaining certain records and reporting specified
events or conditions; demonstrating compliance with financial
qualification requirements and providing specified financial reports;
and maintaining financial protections to address potential accidents.
Section 53.1610 would establish requirements for the provision of
facilities and unfettered access for inspections. These requirements
would be equivalent to Sec. 50.70 with only minor changes proposed to
provide additional flexibilities and address possible differences
related to reactors licensed under part 53 and the possibility that
some commercial nuclear plants may not be assigned resident inspectors.
Section 53.1620 would provide for maintenance of records and the
making of various reports to the NRC. These requirements would be
largely equivalent to Sec. 50.71. This section is not intended to
reflect all provisions in Sec. 50.71; several important requirements
in Sec. 50.71 would be captured in other sections of part 53. For
example, Sec. 53.1545 within subpart I would provide requirements that
would be equivalent to Sec. 50.71(e), updating FSARs, and Sec.
53.1680, ``Annual financial reports,'' would provide the equivalent of
Sec. 50.71(b), which covers financial reports. A reporting requirement
related to completion of power ascension testing would be added to
Sec. 53.1620 to support the assessment of annual fees under 10 CFR
part 171, ``Annual Fees for Reactor Licenses and Materials Licenses,
Including Holders of Certificates of Compliance, Registrations, and
Quality Assurance Program Approvals and Government Agencies Licensed by
the NRC,'' which normally commence upon completion of those testing
activities.
Section 53.1630 would establish requirements for immediate
notification requirements for operating commercial nuclear plants.
These requirements would be equivalent to Sec. 50.72 with minor
changes proposed to make the reporting criteria technology inclusive.
In addition, a new version of NRC Form 361 (NRC Form 361S) would be
created for use by part 53 licensees, but without LWR-specific
terminology to ensure technology inclusiveness. A separate rulemaking
activity, ``Reporting Requirements for Nonemergency Events at Nuclear
Power Plants,'' has been initiated to consider possible changes to the
requirements in Sec. 50.72. At a future date, the NRC may consider
reconciling future changes to Sec. 50.72 with the requirements
proposed in part 53, which have been taken or derived from the current
reporting requirements.
[[Page 86952]]
Section 53.1640 would address the licensee event report system.
These requirements would be equivalent to Sec. 50.73 with minor
changes proposed to make the requirements inclusive of various reactor
technologies and to reflect appropriate internal references to other
sections in part 53. In addition, NRC Forms 366, 366A, and 366B would
be revised to include corresponding check boxes for part 53 licensees.
Section 53.1645 would require periodic reporting of the quantity of
radionuclides released to unrestricted areas in liquid and gaseous
effluents, doses to members of the public, and the results of
environmental monitoring. These reporting requirements in the proposed
part 53 would be largely equivalent to those in the TSs required by
Sec. 50.36a, ``Technical specifications on effluents from nuclear
power reactors.'' The only difference would be that a Sec. 50.36a
requirement to specifically address conditions where the dose to the
maximally exposed individual could be significantly above design
objectives would refer to a design objective of 10 mrem/year total
effective dose equivalent, instead of referring to the design
objectives in appendix I to part 50. The proposed section would also
include an equivalent to the reporting requirement in section IV of
appendix I to part 50 if the radiation exposure to a member of the
public in any calendar quarter exceeds one-half of the annual ALARA
design objective.
Section 53.1650 would include a reporting requirement to support
safeguards agreements between the United States and the International
Atomic Energy Agency (IAEA) and would be equivalent to Sec. 50.78.
Section 53.1660 through 53.1700 would address financial
requirements and would be largely similar to existing regulations in
parts 50 and 52. Section 53.1670 would be entitled ``Financial
qualifications'' and would require applicants other than electric
utilities to possess or have reasonable assurance of obtaining funds
for the activities for which the license is being sought. The NRC is
seeking feedback on these sections and their ramifications for merchant
plants \3\ in section VI, ``Specific Requests for Comments,'' of this
document. The remaining financial reports in part 53 would be
equivalent to Sec. 50.71(b) for annual financial reports, Sec. 50.76
for a change of status, Sec. 50.54(cc) for the filing of a petition
for bankruptcy, and Sec. 50.81 for creditor regulations.
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\3\ A ``merchant plant'' is a plant licensed to a non-rate-
regulated entity (e.g., a nonutility) that engages in the business
of production, manufacturing, generating, buying, aggregating,
marketing, or brokering electricity for sale at wholesale or for
retail sale to the public.
---------------------------------------------------------------------------
Sections 53.1710 through 53.1730 would address financial protection
requirements. Section 53.1720 would require insurance to stabilize and
decontaminate a plant following an accident. These requirements would
be taken from Sec. 50.54(w) with the only notable change being the
addition of a provision allowing plant-specific estimates of costs to
stabilize and decontaminate a plant as an alternative to the $1.06
billion minimum coverage in Sec. 50.54(w). Section 53.1730 is
equivalent to Sec. 50.57(a)(5) and would refer to the requirements in
10 CFR part 140, ``Financial Protection Requirements and Indemnity,''
related to financial protection requirements and indemnity agreements,
including the financial protection requirements of the Price-Anderson
Act.
Subpart M--Enforcement
Subpart M would contain two provisions, Sec. 53.9000 and Sec.
53.9010, which are analogous to provisions contained in other parts of
10 CFR Chapter I imposing requirements on regulated entities. Section
53.9000 would provide notice of the Commission's authority under the
Act to obtain injunctions or other court orders for the enumerated
violations. Paragraph (a) of Sec. 53.9010 would provide notice to all
persons and entities subject to part 53 that they are subject to
criminal sanctions for willful violations, attempted violations, or
conspiracy to violate certain regulations under part 53. Criminal
sanctions would not apply to the regulations listed in paragraph (b).
The regulations for which criminal penalties would apply are limited to
those that establish either a regulatory obligation or prohibition.
V. Changes to Other Parts of 10 CFR Chapter I
10 CFR Part 26
A. Introduction
The NRC is proposing a technology-inclusive, risk-informed, and
performance-based approach for the application of drug and alcohol
testing and fatigue management requirements for facilities licensed
under part 53. The proposed requirements applicable to these
applicants, licensees, and other entities would be commensurate with
the radiological consequences presented by the applicants' facilities
and the operation of these facilities.\4\ The proposed FFD framework
would consist of a two-tiered graded approach similar to that currently
in part 26 and an optional third tier for part 53 commercial nuclear
plants that perform an analysis that demonstrates the facility and its
operation would satisfy the criterion in proposed Sec. 26.603(c),
which refers to Sec. 53.860(a). This proposed FFD framework would be
established in subpart M, ``Fitness for Duty Programs for Facilities
Licensed Under Part 53,'' of part 26.
---------------------------------------------------------------------------
\4\ The NRC uses the term ``operation'' in its part 26
discussion to focus on human performance, namely the necessity of
individuals to operate, maintain, surveil, and protect the facility
and respond to operational transients and unlikely event sequences.
---------------------------------------------------------------------------
The NRC used operating experience to provide regulatory flexibility
in the proposed subpart M of part 26 framework to help support a
licensee's or other entity's response to changes in societal drug use,
drug testing technologies and processes, and FFD program performance.
The flexibility would also help in FFD program implementation because
of the wide variety of staff sizes anticipated at commercial nuclear
plants licensed under part 53 and the geographically remote locations
in which commercial nuclear plants may be sited.
The proposed first-tier FFD program requirements would apply to
part 53 licensees and other entities of commercial nuclear plants under
construction who satisfy the criterion in Sec. 26.603(c) but elect not
to implement proposed Sec. 26.604, ``FFD program requirements for
facilities that satisfy the Sec. 26.603(c) criterion,'' or who do not
satisfy the criterion in Sec. 26.603(c), and to holders of MLs who are
assembling or testing manufactured reactors. These requirements would
be provided in proposed Sec. 26.605(a) and would be essentially
equivalent to those requirements in subpart K, ``FFD Program for
Construction,'' of part 26 as supplemented by select requirements from
subparts E, ``Collecting Specimens for Testing,'' and I, ``Managing
Fatigue,'' of part 26, and the requirements in subparts A,
``Administrative Provisions,'' and O, ``Inspection, Violations, and
Penalties,'' of part 26. The first-tier requirements would involve
policies, procedures, behavioral observation, fatigue management, drug
and alcohol testing, determinations of fitness, appeals, training,
sanctions, auditing, change control, performance monitoring,
recordkeeping, and reporting. These requirements would help deter
individuals subject to this section from illicit drug and/or alcohol
use and from being impaired from any cause including fatigue. These
proposed requirements would also help licensees
[[Page 86953]]
and other entities identify individuals as users of impairing
substances and demonstrate compliance with Sec. 26.23, ``Performance
objectives.''
The proposed second tier would include all the proposed first-tier
requirements, plus the more comprehensive set of FFD program
requirements in current subparts C, ``Granting and Maintaining
Authorization,'' D, ``Management Actions and Sanctions to be Imposed,''
H, ``Determining Fitness-for-Duty Policy Violations and Determining
Fitness,'' and N, ``Recordkeeping and Reporting Requirements,'' of part
26. These requirements would be provided in proposed Sec. 26.605(b)
and would be applicable to licensees and other entities satisfying the
Sec. 26.603(c) criterion, at their discretion. These requirements
would also apply to licensees or other entities not satisfying the
Sec. 26.603(c) criterion that implement an FFD program under subpart M
of part 26, before the loading of fuel onsite into a reactor vessel;
before receiving a manufactured reactor; or before operating, testing,
performing maintenance of, or directing the maintenance or surveillance
of security-related equipment or equipment that a risk-informed
evaluation process has shown to be significant to public health and
safety.
The second-tier requirements are based on the additional risk
presented by nuclear reactor assembly, testing, fueling, and operation
and the necessity for human actions in certain event sequences. The
inclusion of the current part 26 requirements would align proposed part
53 FFD and AA program requirements with the current FFD and AA programs
required for facilities licensed under parts 50 and 52. This approach
would ensure effective and consistent AA and FFD program implementation
across the commercial nuclear power industry, thereby ensuring uniform
requirements for individuals who may perform roles and responsibilities
for multiple facilities regardless of facility licensure.
Proposed Sec. 26.604 would offer an alternate option for an
applicant implementing an FFD program under subpart M of part 26. If
the applicant demonstrates that the criterion in proposed Sec.
26.603(c) is met, then the applicant (and the subsequent licensee or
other entity) must still implement an FFD program described in subpart
M of part 26; however, drug and alcohol testing would not be required
unless FFD performance declines or the applicant, licensee, or other
entity elects to implement drug and alcohol testing. The proposed Sec.
26.604 requirements are equivalent to those proposed in Sec. 26.605(a)
except for required drug and alcohol testing. This proposed framework
would focus on the human performance of individuals while they are
performing those duties and responsibilities that make them subject to
the FFD program. This performance would be verified through behavioral
observation, evaluation of any FFD concerns, performance monitoring,
fatigue management, and determinations of fitness. Applicants that do
not satisfy the criterion in proposed Sec. 26.603(c), or elect not to
perform the analysis required to demonstrate that the criterion in
Sec. 26.603(c) is met, would be subject to an FFD program described in
Sec. 26.605, ``FFD program requirements for facilities that do not
implement Sec. 26.604,'' or an FFD program that implements all part 26
requirements, except for those requirements in subparts K and M of part
26.
In establishing the minimum FFD program requirements in Sec.
26.604, the NRC reviewed current advanced reactor designs against that
of a non-power production or utilization facility (NPUF) that is not
required to implement an FFD program for those individuals who have
unescorted access to the controlled access area (and vital area for
some facilities), including NRC-licensed operators.\5\ This review was
performed because commercial nuclear plants licensed under part 53
could be designed with similar power levels and radiological
consequences as the currently licensed NPUFs. From this review, three
principal considerations supported the minimum set of requirements for
the Sec. 26.604 FFD program.
---------------------------------------------------------------------------
\5\ Controlled access area and vital area are defined in Sec.
73.2, ``Definitions.''
---------------------------------------------------------------------------
First, the radiological consequences presented by a part 53
licensed facility and its operation that satisfy the criterion in Sec.
26.603(c) may present a greater potential radiological consequence to
workers and the public in the vicinity of the facility than does an
NPUF. Second, the operating characteristics of a part 53 licensed
facility are unlike that of an NPUF because there may be a higher
reliance on individuals at the part 53 site to safely and competently
operate, maintain, surveil, and secure SSCs that may not be required at
an NPUF, such as systems that provide secondary heat transfer, reactor
coolant flow, pressure control, and at-power core refueling.
Differences in operating characteristics could include, for example:
long-term, full power operation with automated reactivity control
systems for load-following; active and passive safety and security
systems; innovative non-light-water heat transfer systems; and energy
storage and hazardous chemical systems. The individuals at part 53
facilities may also be required to communicate to individuals both
onsite and offsite, such as electrical load dispatchers, any conditions
adverse to safety, security, or quality. Third, part 53 licensed
facilities may be sited in geographically remote locations that may not
have a physically available administrative or corporate support team to
provide face-to-face oversight, engineering expertise, and maintenance
support like that at NPUFs. This places a higher reliance on those
individuals required at a part 53 facility being fit for duty and
trustworthy and reliable because a replacement individual may not be
readily available.
The NRC proposes to exclude drug and alcohol testing from the
proposed Sec. 26.604 framework for five reasons: (1) the Sec. 26.23
performance objectives can be met through effective implementation of
the defense-in-depth regulatory framework established by behavioral
observation, reporting of legal actions, the proposed performance
monitoring and review program (PMRP), FFD training, and requirements
from the physical protection, AA, cyber protection, and licensed
operator programs; (2) the PMRP would require the licensee or other
entity to monitor its FFD program performance (both qualitatively and
quantitatively) against its historical site performance, fleet-level
performance, if applicable, and industry performance. The licensee or
other entity would be required to implement corrective actions if site
FFD performance meets a licensee- or other entity-established threshold
or to resolve a finding resulting from a qualitative review or audit in
a manner that restores performance and corrects root causes,
contributing causes, or both; (3) the requirements in proposed Sec.
26.609, ``Behavioral observation,'' are more robust than those in Sec.
26.407, ``Behavioral observation,'' of subpart K of part 26 and are
proposed to synchronize with and reinforce the AA behavioral
observation requirements in Sec. 73.56, ``Personnel access
authorization requirements for nuclear power plants,'' or the proposed
requirements under Sec. 73.120, ``Access authorization program for
commercial nuclear plants''; (4) a part 53 commercial nuclear plant
that satisfies the Sec. 26.603(c) criterion will be designed,
operated, and secured with a radiological risk profile that is lower
than that described in Sec. 53.860(a)(2) and
[[Page 86954]]
perhaps will approach the radiological risk profile of an NPUF (which
does not implement an FFD program); and (5) the NRC is aware that a
part 53 commercial nuclear plant could be designed and constructed in
such a manner to reduce reliance on an onsite security force to protect
SSCs, NRC-licensed materials, and sensitive information, with enhanced
capabilities for the detection, assessment, and delay of a DBT
adversary.
Regarding fatigue management requirements, work hour controls would
be required for personnel at utilization and manufacturing facilities
in accordance with the existing scoping criteria in Sec. 26.4, ``FFD
program applicability to categories of individuals,'' as revised in
this proposed rule. The amended Sec. 26.4 also would be used to
determine whether an individual would be subject to drug and alcohol
testing. The applicability of these scoping criteria for certain
individuals (such as operators and maintenance personnel) would be
determined by the licensee or other entity through its risk-informed
evaluation process performed to assess the risk significance of the SSC
upon which work is being performed or directed by the individual. These
requirements also would be scaled based on the potential radiological
consequences presented by the facility. However, fatigue management
would be applied to all individuals subject to the FFD program, similar
to FFD program implementation by the current fleet of commercial
nuclear plants because fatigue management is a proactive requirement
designed to help prevent on-shift impairment through work hour
scheduling and time off. The behavioral observation program (BOP) would
be the principal requirement to provide reasonable assurance that
individuals on shift are not mentally or physically impaired due to
fatigue, which in any way could adversely affect their ability to
safely and competently perform their duties.
The NRC is proposing subpart M of part 26 for facilities licensed
under part 53, in lieu of subjecting all part 53 licensees to the same
part 26 requirements that apply to facilities licensed under part 50 or
52, for four principal reasons. First, subpart M of part 26 would apply
FFD requirements in a risk-informed manner commensurate with the
radiological consequences presented by facilities licensed under part
53. This regulatory strategy is consistent with the current part 26,
which provides a comprehensive set of deterministic requirements for
licensees and other entities at facilities that are operating. This
approach is also consistent with the current subpart K of part 26,
which provides a more flexible framework for nuclear power reactors
under construction, where the probabilities of serious radiological
accidents are lower and consequences from such accidents are less
severe than at operating plants.
Second, subpart M of part 26 would enable a part 53 licensee or
other entity to implement innovative drug testing technologies and
behavior observation techniques while continuing to demonstrate
compliance with the part 26 performance objective in Sec. 26.23(b) of
providing reasonable assurance that individuals are not under the
influence of any substance or mentally or physically impaired from any
cause, which in any way adversely affects their ability to safely and
competently perform assigned duties. These technologies include drug
and alcohol testing using oral fluid, urine, and hair specimens;
screening using point of collection testing and assessment (POCTA)
devices; and monitoring using passive drug and alcohol detection
instrumentation. Part of the basis to enable the use of innovative drug
and alcohol testing technologies is to maintain FFD program
effectiveness should the staff size at a part 53 commercial nuclear
plant be small and challenge the effective implementation of the
behavioral observation and drug and alcohol testing programs. Also, a
commercial nuclear plant that is sited at a geographically remote
location may present additional challenges to behavioral observation
and drug and alcohol testing that are not presented by traditional LWR
facilities licensed under part 50 or 52, such as: efficiency of postal
services for shipping and controlling biological specimens; proximity
to drug and alcohol collection facilities that are reasonably
equivalent to that described in subpart E of part 26; availability of
internet and cellular services to enable same-time discussions among
the Medical Review Officer (MRO), donor, and laboratory; accessibility
to substance abuse treatment services described in subpart H of part
26; and proximity to an MRO (or management and clinical staff) to
evaluate potential impairment caused by fatigue and/or substance use or
abuse, for-cause and post-event occurrences, and the individual's
potential to return to duty.
A part 53 commercial nuclear plant that is sited in a
geographically remote location and has a small staff size may present
implementation challenges and the potential for small group dynamics to
impact FFD program effectiveness. Particularly in isolated
environments, psychological phenomena known as ``groupthink'' may take
effect and could impact the effectiveness of BOPs and the ability to
effectively manage safety culture. For example, in circumstances where
small staffs are drawn from the same small town and thereby have a
potentially narrow experience base, it could be challenging to maintain
a safety conscious work environment in which personnel feel free to
raise safety concerns without fear of retaliation, intimidation,
harassment, or discrimination, and organizations may resultingly
experience groupthink-like effects. Groupthink is particularly
prevalent among cohesive and insulated groups that experience high
levels of decisional stress.\6\ Small staffs at part 53 commercial
nuclear plants may therefore be more susceptible to groupthink if they
are working in an isolated environment where decision-making pressures
may be high.
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\6\ See e.g., Irene W[aelig]r[oslash], Ragnar Rosness, and Stine
Skaufel Kilska, ``Human performance and safety in Arctic
environments,'' SINTEF (2018).
---------------------------------------------------------------------------
Groupthink could have adverse effects on workplace safety culture,
as studies show that individuals will be more hesitant to speak out
against practices they deem unsafe for fear of deviating from group
norms.\7\ Individuals may also be unaware of systematic biases in the
group decision-making process and may then be less likely to scrutinize
the potential risks of the group's decision or sufficiently contemplate
alternative paths of action.\8\ Furthermore, the literature indicates
that groups make riskier decisions than individuals acting alone due to
the diffusion of responsibility among group members.\9\
[[Page 86955]]
This phenomenon, known as ``the risky shift,'' also runs counter to a
safety culture. Accordingly, ``groupthink'' and ``the risky shift'' may
lead to group behaviors that render behavioral observation less
effective. As such, alternative approaches to behavior observation
programs, such as the utilization of video-based surveillance by
individuals separate from the onsite work unit, could serve to mitigate
potential issues associated with groupthink. The incorporation of
remote observation, performed by individuals physically separate from
the site, could help to bring in independent and objective perspectives
and help to break patterns of thought and communication that may result
in groupthink.
---------------------------------------------------------------------------
\7\ See e.g., Russell Mannion and Carl Thompson, ``Systematic
biases in group decision-making: implications for patient safety,''
International Journal for Quality I Health Care, Vol. 26, No. 6
(2014): 606-612 (arguing that small group dynamics in healthcare
teams produce systematic biases in group decision-making because
healthcare professionals may be reticent to vocalize concerns they
have about quality of care).
\8\ See e.g., W[aelig]r[oslash], Rosness, and Kilska (arguing
that groupthink leads teams to ``develop shared rationalizations
that bolster a proposed choice, rather than examining alternative
options and identifying the risks associated with the proposed
choice''). See also David Hofmann and Adam Stetzer, ``A Cross-Level
Investigation of Factors Influencing Unsafe Behaviors and
Accidents,'' Personnel Psychology, Vol. 49 (1996) (finding that in a
study of fatal accidents involving offshore oil rigs, in the absence
of standard operating procedures, workers ``equated normal work
methods (i.e. what everyone else does) with safe and/or ideal work
methods,'' revealing that the groupthink phenomena will further
cement modes of work that do not reflect safety protocols in small
groups that lack strong norms around workplace safety and tacitly
reward short-cuts that prioritize efficiency over safety).
\9\ Mannion and Thompson, ``Systematic biases in group decision-
making: implications for patient safety,'' International Journal for
Quality I Health Care, Vol. 26, No. 6 (2014): 606-612.
---------------------------------------------------------------------------
Even without the influence of small group dynamics, there are other
practical constraints to implementing FFD requirements, such as random
drug and alcohol testing, among small staffs. Random testing is less
effective when applied to small staff sizes because it may be easier
for staff to communicate and predict when individuals will be subject
to drug and alcohol testing. Furthermore, if a facility is sited in a
remote location, program implementation could be challenged by the
following factors: limited mail services to laboratories certified by
the U.S. Department of Health and Human Services (HHS), availability of
local clinical or medical options for treatment and determinations of
fitness by an MRO or Substance Abuse Expert, and use of offsite drug
and alcohol collection facilities.
The increased potential for small staff sizes to impact FFD policy
compliance warrants an approach to FFD that emphasizes performance over
prescriptive requirements that may be ineffective or infeasible at
these facilities. Therefore, the NRC proposes the subpart M of part 26
framework to provide a performance-based approach to FFD. For example,
proposed Sec. 26.603(d) would use existing part 26 auditing
requirements and the reporting requirement in Sec. 26.717, ``Fitness-
for-duty program performance data,'' and clarify how FFD performance
data would be used to maintain or improve, if necessary, FFD program
effectiveness. Specifically, Sec. 26.603(d) would require each
licensee and other entity that elects to implement subpart M of part 26
to monitor and assess their site-specific performance against the
preceding year's site performance, the licensee's most recent fleet-
level performance, and the most recent industry performance. Licensees
and other entities would use these datapoints to develop performance
measures, which would be qualitative descriptions of the specific FFD
program elements, and threshold values for each performance measure
that, if exceeded, would indicate a performance deficiency. Each
licensee and other entity would compare its site's current performance
data against the performance measures and, if a threshold is exceeded,
the licensee or other entity would be required to take corrective
actions to restore performance. Also, the NRC proposes a change control
requirement to allow a licensee or other entity to change its subpart M
of part 26 FFD program while ensuring that FFD program effectiveness is
maintained.
Lastly, subpart M of part 26 would consolidate the applicable FFD
requirements by placing in one subpart all proposed part 26
requirements (either new requirements or cross-references to existing
part 26 requirements) for part 53 licensees and other entities. This
should help licensees and other entities implement the requirements
because it would enable easy cross-reference to similar requirements in
other subparts that are being implemented by non-part 53 licensees and
entities subject to part 26. Understanding how other licensees or other
entities implement similar FFD requirements may facilitate the sharing
of operating experience in program implementation.
The use of innovative technologies and a risk-informed performance-
based framework parallels the considerations presented in the Advanced
Reactor Policy Statement. As stated in the policy statement,
``[S]implified systems should facilitate operator comprehension,
reliable system function, and more straightforward engineering
analysis.'' Furthermore, these same attributes may reduce potential
radiation exposures, help prevent the theft of nuclear materials, and
use technology and design innovations. Should these components and
systems be designed, implemented, and maintained to minimize reliance
on human actions and leverage technology and innovation, then the
robust and prescriptive FFD requirements in, for example, subparts B,
``Program Elements,'' and E of part 26 could be scaled to the part 53-
licensed facility and its operation. This strategy would be implemented
in the subpart M of part 26 framework.
Even though current subpart K of part 26, provides for FFD
requirements commensurate with the radiological consequences presented
by a nuclear power plant construction site, proposed subpart M of part
26 would not allow part 53 licensees and other entities to implement
the requirements in subpart K. The principal reasons are that (without
significant changes to subpart K that would be outside the scope of
this rulemaking): (1) subpart K does not apply to holders of MLs who
assemble or test a reactor; (2) subpart K only applies during
construction, whereas subpart M would apply during construction,
operation, and decommissioning through implementation of the insider
mitigation program (IMP) required by Sec. 73.55 or proposed Sec.
73.100; (3) subpart K does not address training, authorization as
defined in Sec. 26.5, and MRO performance; (4) subpart K does not
expressly authorize the use of innovative drug and alcohol testing
technologies; (5) subpart K does not describe the use of time-dependent
alcohol limits or special analysis testing of dilute urine specimens;
and (6) subpart K has less rigor in the protection of worker rights and
sensitive information than that proposed in subpart M.
Despite the differences between subparts K and M of part 26, the
requirements in subpart M would be essentially equivalent to many in
subpart K that were implemented by the licensees of Vogtle Nuclear
Station and V.C. Summer Nuclear Station when they were constructing
four commercial nuclear power reactors and NRC inspection and operating
experience evaluation determined that the use of subpart K contributed
to adequately protecting the public health and safety and the common
defense and security. Further, given the risk profile posed by
facilities licensed under part 53 and the proposed additional
requirements in subpart M of part 26 that were developed from operating
experience and other part 26 subparts (but are not included in subpart
K of part 26), the NRC concludes that if licensees and other entities
effectively implement the proposed requirements in subpart M of part
26, then individuals subject to the rule should be fit for duty and
trustworthy and reliable.
B. Proposed Changes to Part 26, Subparts A Through E and I
Section 26.3(d) is the applicability paragraph for contractor/
vendors (C/Vs) who implement FFD programs or program elements, to the
extent that the licensees and other entities specified in Sec. 26.3(a)
through (c) rely on those C/V FFD programs or program elements to meet
the requirements of part 26. Section 26.3(d) would be amended to
[[Page 86956]]
address part 53 licensees and other entities in proposed Sec. 26.3(f).
Proposed Sec. 26.3(f) would place part 53 licensees or other
entities within the scope of part 26. For licensees and other entities
of a part 53 commercial nuclear plant, except a holder of an ML, the
FFD program would be required to be implemented no later than the start
of construction activities. The holder of an ML would need to implement
its FFD program before commencing activities that assemble a reactor.
Current Sec. 26.4 describes FFD program applicability to
categories of individuals. These categories are based on the duties,
responsibilities, and the types of access an individual may possess.
The NRC proposes to amend Sec. 26.4 to include licensees and other
entities described in Sec. 26.3(f). The NRC expects that not all
categories of individuals described in current Sec. 26.4 would be
applicable to all part 53 facilities. The NRC is proposing regulatory
guidance in DG-5073, ``Fitness-of-Duty Programs for Commercial Nuclear
Plants and Manufacturing Facilities Licensed Under 10 CFR part 53,''
and DG-5078, ``Fatigue Management for Nuclear Power Plant Personnel at
Commercial Nuclear Plants Licensed Under 10 CFR part 53,'' to help
address program applicability to certain individuals.
Section 26.4(a)(1) and (a)(4) would be amended to account for the
possibility that certain individuals may perform or direct the
performance of operational and maintenance activities from a remote
facility (for example, a remote-control station) for licensees or other
entities licensed under part 53.
The framework of the current part 26 does not account for
individuals who perform operating and maintenance duties at remote
facilities. Although current Sec. 26.4(a)(1) does not limit the
operating of applicable SSCs to onsite operating, Sec. 26.5 limits the
definition of ``Maintenance,'' for the purposes of Sec. 26.4(a)(4), to
include only ``onsite maintenance activities.'' In the 2008 part 26
final rule preamble, the NRC explained that the work hour requirements
apply to those individuals who perform maintenance activities within
the licensee's owner-controlled area. Furthermore, regarding the
direction of applicable operations and maintenance activities, current
Sec. 26.4(a)(1) and (4) address only individuals who perform ``onsite
direction.''
Under the proposed amendments to part 26, the limitation of
``onsite'' activities to those performed within the owner-controlled
area would still apply to facilities licensed under part 50 or 52.
However, for licensees and other entities described in Sec. 26.3(f),
the NRC would remove the ``onsite'' limitation to include activities
performed both within the owner-controlled area as well as operations
and maintenance duties performed at remote facilities where safety-
significant systems and components are expected to be operated within
the design basis of the commercial nuclear plant.
In the 2008 part 26 final rule, the purpose of limiting
``directing'' activities to those ``directing'' activities that are
conducted onsite was to avoid requiring work hour controls for
individuals performing incidental duties, consistent with Sec.
26.205(b)(5), from an offsite location in instances where those duties
might be considered to be ``directive'' in nature. Under the proposed
amendments to part 26, the exclusion of incidental duties while
calculating work hours would still be applicable for licensees and
other entities licensed under part 53. However, for these licensees and
other entities, beyond instances of incidental duties, the direction of
operations and maintenance activities associated with safety-
significant SSCs, when performed at remote facilities, would be
considered in an equivalent fashion as direction performed at non-
remote facilities, for the purposes of administering work hour
controls.
Proposed Sec. 26.4(b) would include in an FFD program individuals
who are granted unescorted access to the protected area of a facility
licensed under part 53 and do not perform or direct the performance of
the duties described in Sec. 26.4(a). This requirement would
contribute to the defense-in-depth regulatory framework that helps
provide that individuals who have unescorted access are fit for duty,
trustworthy, and reliable. For example, the NRC is proposing amendments
to part 73 to require a part 53 licensee to subject individuals to a
series of reviews to help determine whether those individuals are
trustworthy and reliable before granting them unescorted access to the
facility's protected area.
The NRC would amend Sec. 26.4(c) to include in an FFD program
individuals who are assigned to physically report to the part 53
licensee's emergency response facility (or facilities) or participate
remotely in emergency response activities, and individuals without
unescorted access to the part 53 facility who, remotely or otherwise,
make decisions and/or direct actions regarding plant safety or
security. Part 53 commercial nuclear plants may be licensed for and
rely upon offsite facilities to fulfill the role of a Technical Support
Center or Emergency Operations Facility. Therefore, the proposed rule
would account for such offsite facilities or remotely performed
activities. Further, the use of personnel to operate systems and
components, maintain and surveil SSCs, and respond to plant conditions
and security events may be different than those included in the
Technical Support Center or Emergency Operations Facility team for
power reactors currently licensed under part 50 or part 52.
For the individuals whose duties for the licensees and other
entities in Sec. 26.3(c) require the individuals to have the types of
access or perform the activities listed in Sec. 26.4(e)(1) through (6)
at the location where the commercial nuclear plant will be constructed
and operated, current Sec. 26.4(e) requires them to be subject to an
FFD program that satisfies all the requirements of part 26 except
subparts I and K. The NRC would amend Sec. 26.4(e) to except subpart M
as well as subparts I and K. The NRC would also amend Sec. 26.4(e) to
include in an FFD program the individuals whose duties for the
licensees and other entities in Sec. 26.3(f) require the individuals
to have the types of access or perform the activities listed in Sec.
26.4(e)(1) through (6) or perform construction activities as defined in
Sec. 26.5.
Section 26.4(e)(4) would be revised to include in an FFD program
individuals who witness or determine inspections, tests, and analyses
certifications required under part 53 because current Sec. 26.4(e)(4)
includes the individuals who perform the same duties under part 52.
The proposed rule would amend Sec. 26.4(f) to require individuals
who construct or direct the construction of safety- or security-related
SSCs at facilities licensed under part 53 to be subject to an FFD
program under subpart M of part 26 or an FFD program that demonstrates
compliance with all of the requirements of part 26 except for subparts
I, K, and M of part 26.
Section 26.4(g) is the applicability paragraph for FFD program
personnel (e.g., the FFD manager, MRO, and technicians) and persons who
perform AA determinations (e.g., the licensee- or other entity-
designated Reviewing Official). This section would be amended to
address part 53 licensed facilities. Specifically, a part 53 licensee
or other entity would use FFD program personnel to implement its FFD
program as well as other assigned individuals who are not involved in
the day-to-day operations of the program to implement specific elements
of its FFD program, such as the collection of a
[[Page 86957]]
specimen for drug or alcohol testing. These individuals would be held
accountable for program implementation, including consistent
implementation of protections afforded to all individuals subject to
the FFD program.
Section 26.4(h) would be amended to include subpart M of part 26.
The NRC proposes to include several new definitions in Sec. 26.5,
``Definitions,'' and amend some existing definitions. The NRC is
proposing to add a definition for ``Biological marker.'' The proposed
definition would be consistent with ``Biomarker'' defined by the HHS in
its Mandatory Guidelines for Federal Workplace Drug Testing (HHS
Guidelines) using oral fluid as the biological specimen to be tested
(84 FR 57554; October 25, 2019). However, the proposed definition for
Sec. 26.5 would add that the endogenous substance used to validate
that the biological specimen ``was produced by the donor'' because
subpart M of part 26 proposes to have the MRO evaluate any discrepant
biological marker identified in a biological specimen collected from a
donor.
The NRC is proposing a definition for the word ``Change'' as used
in the proposed Sec. 26.603(e), ``FFD program change control,''
process. The proposed definition would be consistent with the
definition of ``Change'' for a part 50 or 52 licensee's emergency plans
in Sec. 50.54(q)(1)(i).
The NRC proposes to revise the definition of ``Constructing or
construction activities'' to clarify that for licensees or other
entities in Sec. 26.3(f), the definition of ``Construction'' would be
that as proposed in Sec. 53.020.
The definitions of ``Contractor/vendor'' (C/V) and ``Other entity''
would be revised to make them applicable to part 53 licensees. A holder
of an ML under part 53 could be a C/V under the proposed C/V
definition.
The NRC is proposing a definition for ``Illicit substance'' because
this phrase is used in subpart M of part 26 and would address
substances that cause impairment and possible addiction but are not an
``illegal drug'' as defined in Sec. 26.5. This proposal is based on
operating experience where individuals have admitted to using common
household, non-drug substances to achieve a high or satisfy an
addiction. These common household items include, but are not limited to
nitrous oxide, butane, propane, glue, paint vapors, lighter fluid, nail
polish remover, degreasers, permanent markers, and methyl alcohol
(which is found in hand sanitizer and mouthwash).
The definition of ``Questionable validity'' would be revised to
make it applicable to an FFD program implemented under subpart M of
part 26, which would include all biological specimens.
The NRC is proposing a definition for ``Reduction in FFD program
effectiveness'' because this phrase, similar to the proposed definition
for ``Change,'' is used in proposed Sec. 26.603(e). The proposed
definition is generally consistent with the definition of ``Reduction
in effectiveness'' provided for emergency plans in Sec.
50.54(q)(1)(iv).
The proposed rule would make the current definition of ``Reviewing
official'' applicable to those licenses and other entities in Sec.
26.3(f).
The current part 26 definition of ``Safety-related structures,
systems, and components'' would be amended to use the NRC's proposed
definition in Sec. 53.020 for the part 53 licensees and other entities
described in Sec. 26.3(d) and (f).
The NRC would amend the definition of ``Security-related SSCs'' in
Sec. 26.5 to make it applicable to a licensee or other entity
described in Sec. 26.3(d) and (f).
The NRC proposes a definition for ``Special Nuclear Material'' that
would refer to the definition in Sec. 70.4, ``Definitions,'' of part
70 to ensure consistency.
The NRC is proposing a revision of the definition of ``Unit
outage'' to account for the potential use of commercial nuclear plants
for purposes other than electricity generation.
Section 26.21, an applicability statement for part 26 FFD programs,
would be amended to include licensees and other entities described in
Sec. 26.3(f) that choose to implement an FFD program that implements
all part 26 requirements, except those in subparts K and M of part 26.
Section 26.51, ``Applicability,'' would be amended to apply to
licensees and other entities described Sec. 26.3(f) that elect not to
implement the requirements in subpart M of part 26 for the categories
of individuals in Sec. 26.4 and those licensees and other entities
that elect to implement the requirements in Sec. 26.605.
Section 26.53(e), (e)(1) and (3), and (g) through (i), which are
general provisions for granting and maintaining authorization, would be
amended to apply to licensees and other entities described Sec.
26.3(f).
Section 26.63(d), a suitable inquiry requirement, would be amended
to apply to licensees and other entities described Sec. 26.3(f).
Section 26.73, the applicability statement for subpart D of part
26, would be amended to apply to licensees and other entities described
Sec. 26.3(f) that elect not to implement the requirements in subpart M
of part 26 for the categories of individuals in Sec. 26.4 and those
licensees and other entities that elect to implement the requirements
in Sec. 26.605(b).
Section 26.81, the purpose and applicability statement for subpart
E of part 26, would be amended to apply to licensees and other entities
described in Sec. 26.3(f) that elect not to implement the requirements
in subpart M of part 26 for the categories of individuals in Sec. 26.4
and those licensees and other entities that implement proposed Sec.
26.605(a) or (b). The subpart E requirements to be implemented are
listed in proposed Sec. 26.607(c)(2)(i) and (c)(2)(ii) and (c)(3).
Section 26.201, the applicability statement for subpart I of part
26 would be amended to apply to licensees and other entities described
in Sec. 26.3(f). Also, the applicability statement would be divided
into two paragraphs for clarity.
The NRC proposes to add Sec. 26.202, ``General provisions for
facilities licensed under part 53,'' for licensees or other entities
described in proposed Sec. 26.3(f) that elect to implement the
requirements in subpart I of part 26 in accordance with Sec. 26.604
and Sec. 26.605. Section 26.202 would establish requirements
equivalent to those in current Sec. 26.203, ``General provisions,''
which is applicable to part 50 and 52 licensees. The NRC would add the
separate Sec. 26.202 because Sec. 26.203 refers to various
requirements under subpart B of part 26, which would not be applicable
to facilities licensed under part 53 that implement subpart M of part
26.
Additionally, Sec. 26.202(c), ``Training and assessments,'' unlike
Sec. 26.203(c), ``Training and examinations,'' would not include a
comprehensive examination requirement because trainee assessment is
conducted as part of a SAT that would be required as proposed under the
FFD program training requirements in Sec. 26.608.
Proposed changes in Sec. Sec. 26.205, 26.207, and 26.211 would add
references to new requirements in subparts I and M of part 26 that
would be applicable specifically to licensees and other entities in
Sec. 26.3(f). The NRC would not change the specific provisions for
work hour requirements in current Sec. 26.205(d). However, as
addressed in the discussion of proposed changes to Sec. 26.4(a),
whether a licensee or other entity under part 26 would need to
implement work hour controls
[[Page 86958]]
for certain individuals or groups would be dependent, in part, on
determinations reached by that licensee's risk-informed evaluation
process.
Proposed changes to Sec. Sec. 26.207(a)(1)(ii) and 26.211(b) would
allow licensees and other entities in Sec. 26.3(f) to perform face-to-
face assessments to support the approval of work hour control waivers
and the conduct of fatigue assessments, respectively, using electronic
communications. These proposals would allow supervisors to conduct such
assessments from a remote location under appropriate circumstances.
Such remotely conducted assessments would need to be supported by
someone who is present in-person with the individual being assessed and
who is trained in accordance with the requirements of either Sec.
26.29 and Sec. 26.203(c) or Sec. 26.608 and Sec. 26.202(c). The
reasoning for these proposals and the associated need for in-person
support to augment electronic communications is addressed further in
the preamble discussion of proposed Sec. 26.619.
C. Proposed Requirements for Part 26, Subpart M
The proposed rule would add a new subpart M to part 26 that would
provide alternative FFD requirements for part 53 licensees and other
entities.
Proposed Sec. 26.601 would make subpart M of part 26 applicable to
part 53 licensees and other entities, at their discretion. If a
licensee or other entity in Sec. 26.3(f) does not elect to implement
an FFD program that demonstrates compliance with the requirements of
subpart M, then the individuals specified in Sec. 26.4 would be
subject to an FFD program that demonstrates compliance with all part 26
requirements, except for those requirements in subparts K and M.
Proposed Sec. 26.603(a) would require an applicant to provide a
description of its FFD program and its implementation within its
application for a license. This requirement is equivalent to the
existing requirements in Sec. Sec. 26.401(b) and 52.79(a)(44). The
entities that would be required to submit these FFD program
descriptions are certain applicants that would comply with the part 53
application requirements in subpart H. In subpart H, Sec.
53.1309(a)(6) would require an applicant for a CP to provide a
description of its FFD program in its PSAR. Under Sec. Sec.
53.1279(b)(4), 53.1369(x), and 53.1416(a)(24), an applicant for an ML,
OL, and COL, respectively, would be required to provide a description
of its FFD program in its FSAR.
Unlike an application for a license, a description of an FFD
program does not receive NRC review for possible approval. The
applicant provides the NRC with information about the applicant's
proposed FFD program to inform the NRC's inspection program and to
demonstrate that the FFD program will be effectively implemented before
a licensee or other entity commences any activity making individuals at
the NRC-licensed facility subject to the FFD program.
Proposed Sec. 26.603(a)(1) would require a summary description of
the analysis described in Sec. 26.603(c), if performed. The analysis
should describe the operation of the facility. This would include
informing the Commission of: (1) the principal individuals assigned by
job title (work category) and a summary description of the human
actions (e.g., monitoring, operating, responding, surveillance,
oversight, etc.) that they perform to maintain the facility in a safe
operating or shutdown condition; (2) the principal individuals by job
title and a summarized description of the human actions to secure and
protect the facility (without providing sensitive information); (3) the
estimated total population of individuals subject to the FFD program
and per shift by job description; and (4) references to supporting
documentation. The purpose of these descriptions is to enable an NRC
assessment of the licensee's or other entity's analysis and the
required human actions to operate, monitor, surveil, maintain, and
secure the facility within its design and licensing basis so that if an
operational or security-related event were to occur, the facility would
respond as designed and licensed and the calculated radiological dose
consequences would not exceed the consequences described in Sec.
53.860(a)(2). This is important because facilities that implement Sec.
26.604 are expected to have very small staff sizes and may be sited in
geographically remote locations, both of which could challenge
effective implementation of the FFD program.
Proposed Sec. 26.603(a)(2) would require the applicant to state
what FFD program it plans to implement.
Proposed Sec. 26.603(a)(3) would require a discussion that informs
the NRC of the applicability of the applicant's FFD program to
individuals who perform safety- or security-significant activities.
This description should summarize any key differences between the staff
at the site and any remote facility and the categories of individuals
in Sec. 26.4. The principal purpose of providing this description
would be to inform the NRC of any substantial differences in the
applicability of the FFD program to the categories of individuals in
Sec. 26.4.
Proposed Sec. 26.603(a)(4) would require a description of the drug
and alcohol testing and fitness determination process to be implemented
through the licensee's or other entity's procedures, including the
collection and testing facilities to be used, biological specimens to
be collected, and sanctions to be imposed upon a confirmed FFD policy
violation. This process includes how individuals who test positive for
a drug or alcohol will be evaluated before being afforded unescorted
access to the protected area to perform or direct those duties or
responsibilities making them subject to the FFD program. The principal
purpose of describing this return-to-duty process is to inform the NRC
of the behavioral observation strategy (for those facilities that
implement Sec. 26.604) and/or drug screening and testing strategy.
Proposed Sec. 26.603(a)(5) would require a summary description of
the applicant's planned PMRP. This description must provide the
performance measures and thresholds that the applicant intends to use.
Proposed Sec. 26.603(b) would establish when the FFD program must
be implemented and the longevity of the FFD program. This proposal is
equivalent to the current Sec. 26.3, which states, in part, when
licensees and other entities must begin implementing their FFD
programs. Unlike the current part 26 regulations, proposed Sec.
26.603(b) would expressly state that an FFD program would not be
applicable during decommissioning of a part 53 facility for licensees
and other entities specified in Sec. 26.3(f). However, licensees of
facilities licensed to operate a reactor should be aware that the
physical protection program under Sec. 73.55, ``Requirements for
physical protection of licensed activities in nuclear power reactors
against radiological sabotage,'' and under proposed Sec. 73.100
include a requirement for the implementation of an IMP, even during
decommissioning.
Proposed Sec. 26.603(b) would also require the holder of an ML to
implement its FFD program no later than the start of activities that
assemble a reactor. The holder of the ML should establish in its
procedures when reactor assembly commences and what constitutes
assembly. For example, the FFD program would not need to be implemented
for the receipt, storage, inspection, and staging of components and
systems used to assemble (i.e., build or fabricate) the reactor because
this is not a current requirement for LWR facilities licensed under
part 50 or 52. Furthermore, the NRC currently does
[[Page 86959]]
not require that an FFD program be applied to the assembly or
manufacturing of components (or basic components as defined in Sec.
21.3), or systems that were fabricated or assembled outside the
footprint of a commercial power reactor, and this regulatory position
would also apply to a manufacturing facility.
Proposed Sec. 26.603(c) would require the applicant, licensee, or
other entity seeking to implement an FFD program under Sec. 26.604 to
perform a site-specific analysis to determine whether the facility and
its operation satisfy the criterion in Sec. 53.860(a)(2). If the
analysis is performed and demonstrates that the radiological
consequences presented by the facility and its operation satisfy the
criterion, then the licensee or other entity could implement the FFD
program detailed in Sec. 26.604. If the analysis does not demonstrate
that the facility and its operation satisfy the criterion, then the
licensee or other entity must implement the FFD program described in
either Sec. 26.605 or subparts A through I, N, and O of part 26.
Proposed Sec. 26.603(c) would also require licensees and other
entities that implement proposed Sec. 26.604 to update the technical
analysis used to justify compliance with the criterion in Sec.
53.860(a)(2). This analysis would be updated to reflect changes made to
the staffing, FFD programs, or offsite support resources described in
the analysis to show that the facility and its operation continue to
satisfy the criterion. This is important because facility, operation,
or staffing changes outside FFD program implementation (e.g., changes
in the facility safety analysis, physical protection strategies, or the
security plan, implementing procedures, or contingency response
strategies) could adversely impact the licensee's or other entity's
documented analysis demonstrating that the facility and its operation
satisfy the criterion if event sequences require human action.
Proposed Sec. 26.603(d) would require the establishment of a PMRP.
The concept of a PMRP is not new. This requirement would consolidate
for part 53 the requirements in current Sec. Sec. 26.41, ``Audits and
corrective actions''; 26.415, ``Audits''; 26.717, ``Fitness-for-duty
program performance data''; and 26.183(c), which describes MRO
responsibilities. The proposal would state that the licensee or other
entity must monitor the effectiveness of its FFD program by comparing
performance data against performance measures and thresholds. The
development of quantitative thresholds would be new, but this is born
from licensees and other entities with facilities licensed under parts
50 or 52 already collecting, reviewing, and reporting FFD performance
data. Additionally, the benefit of quantitatively measuring FFD program
performance against established thresholds benefits a licensee's and
other entity's determination of whether they are maintaining FFD
program performance in a manner that demonstrates compliance with the
performance objectives in Sec. 26.23.
The NRC is proposing the PMRP because the subpart M of part 26
requirements would enable a high degree of flexibility in FFD program
implementation (e.g., drug testing). A licensee or other entity would
not only have options in the type of FFD program they may implement
under part 26, but they would have options in the types of biological
specimens they may test for drugs, where to collect the biological
specimens (e.g., at the NRC-licensed facility or offsite at a local
hospital or clinic), and the use of collection and assessment devices
to screen individuals for drugs and alcohol. These FFD program
flexibilities could cause FFD programs under subpart M of part 26 to
become very site-specific, necessitating performance measures to enable
the licensee or other entity to maintain the effectiveness of its FFD
program.
Fitness-for-duty program effectiveness would be determined by
comparing actual performance against the performance measures and
thresholds. The result of that comparison would inform licensee or
other entity decisions whether to change FFD program elements to
address a performance deficiency. Also, the thresholds would have
sufficient margin, based on operating experience, before conditions
adverse to safety and security may occur should an individual be
identified as impaired or not trustworthy and reliable. The potential
of a human-related failure causing a condition adverse to safety and
security is dependent on the duties and responsibilities of the
individual and the defense-in-depth designed to prevent or mitigate an
adverse consequence. The PMRP would account for this by requiring the
review of FFD performance data, in part, by work category, C/V, and
individuals employed by the licensee who are not a C/V as defined in
Sec. 26.5 (i.e., a licensee employee).
Proposed Sec. 26.603(d)(1) would require the licensee or other
entity to document and maintain its PMRP. Proposed Sec.
26.603(d)(1)(i) would require that the performance measures be
identified and designed to monitor FFD program performance. Proposed
Sec. 26.603(d)(1)(i)(A) would require the FFD program of a licensee or
other entity subject to the requirements of Sec. 26.604 to include
monitoring of the BOP. The purpose of this monitoring is to help ensure
that individuals subject to the FFD program are observing the behaviors
of others, are being observed themselves, and are reporting FFD
concerns to licensee- or other entity-designated individuals. The other
performance measures would include occurrence of FFD policy violations
evaluated by licensee employee, C/V, and labor category, and occurrence
of individuals with potentially disqualifying information or who
possessed an FFD prohibited item.
Proposed Sec. 26.603(d)(1)(i)(B) would require the FFD program of
a licensee or other entity that is either subject to the requirements
of Sec. 26.604 and has implemented a drug testing program at its
discretion, or is subject to the requirements of Sec. 26.605, to
include the performance measures identified in Sec. 26.603(d)(1)(i)(A)
and those necessary to monitor the effectiveness of the drug and
alcohol testing program. The drug and alcohol measures would include
the monitoring of FFD performance data for pre-access and random
testing and subversion attempts by the categories of licensee employee,
C/V, and labor category.
Proposed Sec. 26.603(d)(1)(ii) would require the licensee or other
entity to establish thresholds for each performance measure. Initial
thresholds must be based on FFD performance data from comparable
facilities subject to part 26, the licensee's or other entity's fleet-
level program performance if applicable, and industry FFD performance
data. This provision introduces the requirement to ``maintain FFD
program effectiveness.'' This terminology describes a performance-based
regulatory strategy in which the licensee or other entity must
initially establish a level of performance that is representative of
other facilities in the licensee's fleet of facilities subject to part
26, if applicable, and the FFD performance of comparable facilities
subject to part 26.
Proposed Sec. 26.603(d)(1)(iii) would require that the licensee or
other entity evaluate FFD data as it is received to determine whether a
threshold has been exceeded. Historical FFD performance data for the
current LWR fleet indicates that, for particular work categories and
employment types, few FFD policy violations occur per year. Therefore,
for work categories that may be significant to worker safety (e.g.,
radiation protection technicians), physical
[[Page 86960]]
protection (i.e., security personnel), or safety (i.e., NRC-licensed
operators and individuals who perform or direct the performance of
activities that a risk-informed evaluation process has shown to be
significant to public health and safety), a single FFD policy violation
could be a significant occurrence and warrant corrective actions. Based
on licensee-submitted FFD-related reports under Sec. Sec. 26.417,
26.419, 26.717, and 26.719, licensees and other entities with
facilities licensed under parts 50 or 52 implement some form of
corrective action that is typically scaled to the significance of the
violation. These corrective actions have included counseling, follow-up
drug and/or alcohol testing, remedial training, generic announcements
to the workforce, and reviews of recently performed or directed work by
the individual suspected of being impaired. Proposed Sec.
26.603(d)(1)(iii) would require that the PMRP include a year-to-year
comparison of FFD performance data to help provide assurance that an
adverse trend in FFD program performance would be identified if
occurring. This proposed requirement was developed from the annual FFD
performance data reporting requirements in Sec. Sec. 26.417(b)(2) and
26.717. In particular, the proposed year-to-year comparison of FFD
performance data is equivalent to Sec. 26.717(c), which requires, in
part, licensees and other entities to analyze their performance data at
least annually and take appropriate actions to correct any identified
program weaknesses.
Proposed Sec. 26.603(d)(1)(iv) would require the licensee or other
entity to perform and document quantitative and qualitative reviews.
These reviews would be performed in three program areas: protections
afforded to individuals subject to the FFD program, laboratory test
results and MRO performance, and change control. The purpose of these
reviews would be to specifically target performance within the three
program areas to assess whether the outcomes resulting from the
implementation of procedure requirements are contributing to FFD
program effectiveness. The proposed reviews would not require the
establishment of measures and thresholds because the reviews are
expected to result in qualitative findings regarding program
effectiveness. Qualitative findings and observations could still result
in the consideration of corrective actions in the targeted program
areas.
Proposed Sec. 26.603(d)(1)(iv)(A) would require the licensee or
other entity to monitor whether its FFD program is affording
appropriate protections to individuals subject to the FFD program. The
review of these protections would include, in part, assessing the
licensee's or other entity's protection of the following: privacy
during the specimen collection process; specimen integrity, custody,
and control; information gathered from FFD program implementation; and
due process during appeals of FFD policy violations.
Proposed Sec. 26.603(d)(1)(iv)(B) would require, in part, a review
of laboratory test results and MRO performance. Effective performance
by the laboratory (e.g., obtaining and communicating accurate test
results) and MRO (e.g., correct evaluation of the laboratory test
results based on Sec. 26.185 or HHS Guidelines) would result in three
significant outcomes: (1) protection of the donor from an inaccurate
FFD policy violation determination; (2) protection of the donor, other
individuals, and the facility from potential harm should the donor be
impaired or not trustworthy and reliable; and (3) a performance-based
assessment of both the laboratory and MRO. This last outcome could
facilitate actions to improve laboratory performance, MRO training
under Sec. 26.607(m), or both. Proposed Sec. 26.603(d)(1)(iv)(B)
would also require a comparative analysis between the POCTA screening
result(s) and the corresponding specimen test results obtained from the
HHS-certified laboratory if the POCTA indicated a positive,
adulterated, substituted, or invalid screening result or discrepant
biological marker, to assess the effectiveness of the POCTA and to
inform MRO decisions under Sec. 26.185 or Sec. 26.607(m)(6). The
results of this biennial review could also inform the conduct of
laboratory audits.
Proposed Sec. 26.603(d)(1)(iv)(C) would require that the change
control requirement in proposed Sec. 26.603(e) be included in the
biennial program review to help ensure that changes implemented over
the life of the facility do not result in a reduction in program
effectiveness even if a mitigating action was implemented for the
specific change. This requirement was developed from Sec. Sec.
26.137(f) and 26.713(d). This part of the review would require an
assessment of all changes since the last review and their potential
aggregated impact on FFD program effectiveness. For example, if last
year the licensee elected to contract with a different MRO and this
year the licensee implemented a new type of POCTA device, each of those
program changes probably would not have resulted in a recognizable
reduction in FFD program effectiveness. But, if the drug testing
positivity rate (or FFD policy violations) for C/Vs decreased markedly
during a future maintenance outage that required many C/Vs, then the
reduction could indicate, for example, that the POCTA device was not as
effective as determined by a forensic toxicologist review under
Sec. Sec. 26.603(e) and 26.607(h) or that the new MRO was improperly
crediting prescription medication for laboratory-confirmed positive
test results.
Proposed Sec. 26.603(d)(2) would state when the licensee or other
entity must implement corrective actions. This requirement would be
equivalent to the requirement in current Sec. 26.415(b) and was
developed from requirements contained in Sec. Sec. 26.41(a) and (f),
26.127(e), 26.129(b)(1)(i), 26.137(f)(3) through (5), 26.155(a)(6),
26.157(e), 26.159(b)(1)(i), and 26.203(e)(2). Corrective actions must
be implemented to correct root causes, contributing causes, or both.
There is margin built into the FFD performance thresholds and
qualitative factors (e.g., to account for potential changes in drug and
alcohol testing performance data when there is a large influx of C/Vs
to perform maintenance) that may influence a licensee or other entity's
causal determination for an occurrence. Thus, generalized or
qualitative corrective actions may be implemented like informing
management and placing a sufficiently descriptive summary of the
occurrence in a corrective action program for future monitoring to
assess recurrence.
However, should the occurrence challenge safety or security or
significantly exceed a performance threshold even when considering
qualitative factors and margin, the licensee or other entity should
implement more robust corrective actions to resolve the cause. An
example of a challenge to safety or security would be the situation
when an NRC-licensed operator or maintenance professional had operated,
surveilled, or maintained safety-significant SSCs and was determined to
have been impaired by behavioral observation or potentially under the
influence of a narcotic as determined by an alcohol or drug test or
screening result. Immediate corrective actions could include, but would
not be limited to, a licensee or other entity assessment of the duties
and responsibilities recently performed by the individual. Operating
experience within the LWR operating reactor community demonstrates few
FFD policy violations per year per site have been caused by individuals
who perform or direct the performance of
[[Page 86961]]
safety or security-significant activities. Therefore, any such
violations of the FFD policy in a particular work category in one year
could be a significant performance deficiency. These violations could
be even more significant at part 53 facilities that have a very small
workforce subject to part 26.
Proposed Sec. 26.603(d)(3) would require the licensee or other
entity to biennially assess and document its FFD performance monitoring
program; this requirement was developed from Sec. 26.41(b). This
documented review would demonstrate that the performance measures and
thresholds are appropriate based on site- and licensee's fleet-level
program performance, if applicable, and industry performance and
adjusted to maintain FFD program effectiveness. Also, as a result of
this effort, the licensee or other entity would be in possession of
lessons learned from fleet-level performance, if applicable, and
industry performance that could contribute to their own performance
assessment to maintain program effectiveness.
Under proposed Sec. 26.603(d)(3)(i), the identified program
weaknesses and corrective actions resulting from the biennial review
would be required to be summarized in the licensee's or other entity's
annual report to the NRC in compliance with either Sec. 26.417(b)(2)
or Sec. 26.717, as applicable. This information would inform the NRC
of FFD program weaknesses to facilitate regulatory oversight and enable
the NRC to aggregate industry data for use in a licensee or other
entity PMRP.
Proposed Sec. 26.603(d)(3)(ii) would establish when the biennial
PMRP review must be completed and when corrective actions from the
review must be implemented. The NRC selected the May 15th date of odd-
numbered years to help ensure that all FFD programs will maintain their
previously determined performance measures and thresholds or reset them
based on FFD program performance early in the year in which the
biennial review was conducted. This would assist in obtaining quality
FFD performance data over two annual reporting cycles and evaluating
whether previous corrective actions were effective.
In proposed Sec. 26.603(e), the NRC proposes a change control
requirement for subpart M of part 26 FFD programs. Requiring licensees
and other entities to demonstrate compliance with certain requirements
before implementing changes to their FFD programs would be necessary
for two primary reasons. First, proposed changes to a licensee's or
other entity's FFD program could affect the analysis performed by the
licensee or other entity under proposed Sec. 26.603(c), which helps
determine the FFD program requirements that must be implemented. If
this analysis changes, then the licensee's or other entity's FFD
program requirements might change. Second, the requirements in subpart
M of part 26 are performance based. Therefore, FFD program
implementation may change periodically in response to societal changes
in substance abuse or from PMRP implementation. Change control
therefore relies on the licensee or other entity maintaining its
procedures in a manner that details how its FFD program is to be
implemented while incorporating changes, with documentation that
justifies the changes to support the PMRP, audits, and NRC inspection.
Proposed Sec. 26.603(e)(1) would permit the licensee or other
entity to implement changes to its FFD program if it performs and
retains an analysis demonstrating that the change does not reduce the
effectiveness of the FFD program or the change was necessitated or
justified by a change to part 26, laboratory processes, or guidance
issued by the HHS or NRC. The proposed change control requirement would
enable flexibility in program implementation should the NRC or HHS
change its drug testing procedures (as implemented by the licensee or
other entity through its procedures) in response to changes in societal
substance abuse or drug testing technologies.
The proposed change control requirement was developed from the
change control requirements in Sec. 50.54(p) and (q)--the change
control requirements for security and emergency plans, respectively.
However, unlike these two requirements, the NRC does not review and
approve a licensee's or other entity's FFD program or its implementing
procedures, and the FFD program is not licensing-basis information as
described in Sec. 53.1300.
Proposed Sec. 26.603(e)(2) would require that if a change reduces
FFD program effectiveness, then the licensee must implement a
mitigating strategy so the FFD program, as revised, will continue to
demonstrate compliance with the performance objectives in Sec. 26.23
and not result in a reduction in program effectiveness.
Proposed Sec. 26.603(e)(3) would prohibit, with one exception, the
use of the change control process to reduce the minimum panel of drugs
to be tested and would reference the drugs listed in proposed Sec.
26.607(c)(1). Proposed Sec. 26.607(c)(1) would reference current Sec.
26.31(d)(1), which states that, at a minimum, licensees and other
entities shall test for marijuana metabolite, cocaine metabolite,
opioids (codeine, morphine, 6-acetylmorphine, hydrocodone,
hydromorphone, oxycodone, and oxymorphone), amphetamines (amphetamine,
methamphetamine, methylenedioxymethamphetamine, and
methylenedioxyamphetamine), phencyclidine, and alcohol. The testing of
these drugs and drug metabolites, except phencyclidine, and alcohol is
necessary for the FFD program to remain effective. Also, there is no
proposed subpart M of part 26 requirement stating that this panel of
drugs and drug metabolites needs to consist of only scheduled
drugs.\10\ This flexibility would account for the situation where an
impairing substance becomes prevalent in society and a licensee or
other entity elects to add the substance to their panel of substances
to be tested prior to it being scheduled by the Drug Enforcement
Administration.
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\10\ The Drug Enforcement Administration classifies drugs,
substances, and certain chemicals used to make drugs into five (5)
distinct categories, depending upon the drug's acceptable medical
use and the drug's abuse or dependency potential. These categories
appear as Schedules I through V of section 202 of the Controlled
Substances Act (21 U.S.C. 812). Schedule I drugs have a high
potential for abuse, have no currently accepted medical uses in
treatment in the United States, and lack accepted safety for use
under medical supervision. At the other end of the classification
scheme, Schedule V drugs have the least potential for abuse among
the five categories of drugs, have a currently accepted medical use
in treatment in the United States, and abuse of the drug may lead to
limited physical dependence or psychological dependence. For more
information, see https://www.dea.gov/drug-information/drug-scheduling.
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The exception in proposed Sec. 26.603(e)(3) would be that, should
HHS elect to remove phencyclidine from the panel of drugs and drug
metabolites to be tested, a licensee or other entity could make this
change in its FFD program without resulting in a reduction in FFD
program effectiveness. This outcome would be justified based on the
very infrequent occurrence rate of FFD policy violations due to
phencyclidine use since 2010. However, if HHS proposes to remove a
class of drugs from the panel of drugs to be tested that is listed in
Sec. 26.31(d)(1), except for phencyclidine, then a licensee or other
entity may not make a similar change to its panel of drugs to be
tested, because this change would be a reduction in FFD program
effectiveness even with a mitigative strategy implemented.
Changes in the HHS panel of drugs and drug metabolites to be tested
may also shift from one metabolite to a
[[Page 86962]]
different metabolite for the same drug class (e.g., amphetamines,
opioids) to be tested. Should HHS issue such a change to its panel,
this would not be expected to result in a reduction in FFD program
effectiveness because HHS would be targeting a more prevalent or
effective metabolite in its drug testing program. This situation could
occur as HHS gathers more operating experience from Federal Government
implementation of its HHS Guidelines, or data generated by drug testing
laboratories and federally mandated drug testing programs required by
Federal agencies such as the NRC and U.S. Departments of
Transportation, Energy, and Defense.
Proposed Sec. 26.603(e)(4) would require that change control
records be maintained for a 5-year record retention period based on the
current NRC practice to conduct triennial inspections of licensees' and
other entities' FFD programs. This would afford the NRC an opportunity
to review the licensee's or other entity's determination that FFD
program changes have not reduced the effectiveness of their FFD
program. Licensees and other entities would also be required to
summarize each change made under proposed Sec. 26.603(e) in their
annual FFD performance reports required by Sec. 26.617(b)(2) or Sec.
26.717, as applicable.
Proposed Sec. 26.604 would establish the minimum set of FFD
program requirements for licensees and other entities who have a
documented analysis that demonstrates that the facility and its
operation satisfy the criterion in Sec. 53.860(a)(2). For these
licensees, compliance with the performance objectives in Sec. 26.23
would be ensured through the BOP; defense-in-depth measures proposed in
subpart M of part 26 like the PMRP, change control, and audits; and
other requirements, such as those for AA, physical protection, and
licensed operators. The adequacy of these measures in satisfying the
performance objectives is supported by operating experience, which
demonstrates margin between an FFD-related occurrence and a condition
adverse to safety or security, as illustrated by for-cause, post-event,
and random testing data. A facility that satisfies the criterion in
proposed Sec. 53.860(a)(2) would present a smaller potential
radiological consequence than a facility that does not satisfy the
criterion, so the requirements in proposed Sec. 26.604 are scaled to
the lower risk presented consistent with the Commission's Advanced
Reactor Policy Statement.
The disadvantages of implementing the FFD program described in
proposed Sec. 26.604 would be few. Since drug and alcohol testing
would not be required, behavioral observation would be the keystone
requirement in this performance-based framework to provide that
individuals are fit for duty, trustworthy, and reliable, and can safely
and competently perform the duties and responsibilities making them
subject to the FFD program. If not, the individuals would be assessed
in accordance with the licensee's or other entity's procedures similar
in manner to that required by subpart K of part 26, and the proposed
PMRP would require corrective actions should a threshold be exceeded.
If a licensee or other entity elects not to perform the analysis in
proposed Sec. 26.603(c) to determine whether it satisfies the
criterion in proposed Sec. 53.860(a)(2); performs the analysis and
finds that the facility and its operation does not satisfy the
criterion in proposed Sec. 26.603(c); or is a holder of an ML, the
licensee or other entity could not implement the FFD program described
in Sec. 26.604. Instead, the licensee or other entity would implement
either the program described in proposed Sec. 26.605 or an FFD program
that demonstrates compliance with all the requirements in current
subparts A through I, N, and O of part 26.
Proposed Sec. 26.605 would establish requirements in a graded
manner similar to the regulatory framework established by the
requirements in subparts A through I, N, O, and K of part 26. This
existing graded approach consists of an FFD program for construction of
a commercial nuclear plant and a more robust program that must be
implemented before reactor operation. The former is the FFD program in
proposed Sec. 26.605(a), and the latter is proposed Sec. 26.605(b).
Like that for an FFD program under Sec. 26.604, the FFD program under
Sec. 26.605 would include FFD program elements similar to those in
subpart B of part 26, but the proposed requirements are less
prescriptive, enabling more flexibility in program implementation like
that offered in subpart K of part 26. For example, the requirements in
subpart B of part 26 are explicit requirements for, in part, the
collection and analysis of urine specimens. Subpart B of part 26 does
not enable the use of oral fluid for drug testing or screening, except
under very limited situations as described in subpart E of part 26, or
the use of hair specimens, unlike proposed Sec. 26.605. Proposed Sec.
26.605 would require drug and alcohol testing based on either the
requirements in part 26 or the HHS Guidelines. The principal benefit of
the proposed Sec. 26.605 FFD program is that it would provide a
regulatory framework that is consistent with the radiological
consequences for a facility that does not satisfy the criterion in
proposed Sec. 53.860(a)(2) while affording flexibilities in the
conduct of drug and alcohol testing.
Proposed Sec. 26.605(a) would apply to licensees and other
entities who perform the Sec. 26.603(c) analysis and satisfy the
criterion in Sec. 53.860(a)(2) but decide not to implement the FFD
program described in proposed Sec. 26.604, licensees and other
entities who do not perform the Sec. 26.603(c) analysis, and licensees
and other entities who perform the analysis but their analysis does not
demonstrate that their facility and its operation satisfy the criterion
in Sec. 53.860(a)(2). These entities must establish, implement, and
maintain an FFD program under Sec. 26.605(a) either during
construction activities as defined in Sec. 26.5, or during activities
performed under an ML that allows the assembly, testing, or both, of a
manufactured reactor. This FFD program implements all the FFD program
requirements in Sec. 26.604 plus drug and alcohol testing.
The timing element of the proposed applicability statement of Sec.
26.605(a) is equivalent to that for an LWR licensee or other entity who
is performing those same activities at a facility licensed under part
50 or 52 and helps provide assurance that those individuals who
assemble, test, or perform construction activities as defined in Sec.
26.5 or direct these activities are fit for duty and trustworthy and
reliable. This is important because assembly and testing a manufactured
reactor and the construction and testing of SSCs required for facility
operation require, in part, adherence to procedures, possible
implementation of unique and precise assembly techniques, and quality
assurance and controls. Additionally, SSCs within a manufactured
reactor may not be accessible, testable, or available for quality
assurance and verification after the reactor is assembled. This
requirement is also proposed to address solo-assembly activities that
may cause latent failures and passive SSCs located internal to a
reactor (for example, a fusible link designed to melt at a particular
temperature to trigger an actuation mechanism) that are relied upon for
safe operation but cannot be inspected or tested for proper
installation, configuration, or operation after installation. A Sec.
26.605(a) FFD program for these types of activities is equivalent to
the FFD program applicable to the assembly of the reactor vessel
internals and testing of the SSCs internal to the reactor at an LWR
licensed under part 50 or 52.
[[Page 86963]]
Proposed Sec. 26.605(b) would apply to the same licensees and
other entities as in proposed Sec. 26.605(a) but before the loading of
fuel onsite into a reactor vessel; before receiving a manufactured
reactor; or before individuals subject to part 26 operate, test,
perform maintenance of, or direct the maintenance or surveillance of
security-related equipment or equipment that a risk-informed evaluation
process has shown to be significant to public health and safety. These
entities must establish, implement, and maintain an FFD program that
implements all the requirements in Sec. 26.605(a), except proposed
Sec. Sec. 26.610, ``Sanctions''; 26.617, ``Recordkeeping and
reporting''; and 26.619, ``Suitability and fitness determinations'';
plus additional requirements due to the increased radiological
consequences presented by a part 53 commercial nuclear plant as the
licensee readies it for operation. These additional requirements
include those in subparts C, D, H, and N of part 26, some of which
would replace Sec. Sec. 26.610, 26.617, and 26.619.
Proposed Sec. 26.605(b) would also enable the licensee or other
entity to better integrate its facility into the LWR fleet and Category
I fuel cycle facilities because subparts C, D, and H of part 26 would
be required. These subparts would be required, in part, because it is
expected that: (1) individuals will be able to work at any part 50, 52,
or 53 commercial nuclear plant and will possess a nuclear safety
culture and desirable qualifications, skills, expertise, or services;
and (2) licensees and other entities of facilities licensed under parts
50, 52, and 70 may venture to construct or operate a facility licensed
under part 53. Therefore, the implementation of these subparts would
help ensure that all individuals subject to part 26, except those
individuals subject to an FFD program under Sec. 26.604, Sec.
26.605(a), or subpart K of part 26, would be subject to FFD programs
that provide reasonable assurance that the individuals are fit for
duty, trustworthy, and reliable.
Proposed Sec. 26.606, ``Written policy and procedures,'' would
require licensees and other entities to implement and maintain an FFD
policy and procedures for their FFD programs. This section would
establish requirements equivalent to those in current Sec. 26.403,
``Written policy and procedures,'' of subpart K. However, a principal
difference is that proposed Sec. 26.606 is written to enable the use
of urine, oral fluid, and hair for drug testing and screening.
Proposed Sec. 26.606(a)(1) would require each licensee and other
entity to provide a written FFD policy statement to individuals subject
to the FFD program before the individuals are subjected to behavioral
observation and any FFD program drug and alcohol test. This would be a
protection measure afforded to individuals subject to the FFD program
to help ensure that they know what is expected of them before being
subject to the FFD program and potential consequences should they
violate the FFD policy or procedures. This requirement would also
contribute to safety and security because understanding FFD program
responsibilities may enhance an individual's safety culture or the
individual may self-select out of the licensee's or other entity's
hiring process.
Proposed Sec. 26.606(a)(2) would require that the FFD policy
statement describe the performance objectives in Sec. 26.23, which are
the same FFD program performance objectives required for facilities
licensed under parts 50, 52, or 70. Having a standard performance
outcome based on a licensee or other entity satisfying the Sec. 26.23
performance objectives would enhance consistency in FFD program
implementation across all entities subject to part 26. It would also
generate confidence that individuals subject to part 26 will safely and
competently perform their duties and responsibilities and use NRC-
licensed materials in a manner that will protect the public health and
safety and common defense and security.
Proposed Sec. 26.606(a)(3) would require that the FFD policy
statement describe the minimum days off requirements in Sec.
26.205(d)(3) or maximum average work hours requirements in Sec.
26.205(d)(7).
Proposed Sec. 26.606(a)(4) would require the FFD policy statement
be written in sufficient detail to provide affected individuals with
information on what is expected of them and what consequences may
result from a lack of adherence to the policy, including those elements
described in Sec. 26.603(b), part 26-required sanctions, and required
medical/clinical treatment and follow-up testing for FFD policy
violations. This requirement is equivalent to Sec. 26.403(a) of
subpart K but includes an additional description of what the policy
statement must include. For example, the policy would describe the NRC-
required sanctions to help deter substance abuse and required medical/
clinical treatment and follow-up testing for FFD policy violations.
This provision would provide a protection measure by helping the
individual get the assistance they need and help ensure that the
individual refrains from substance abuse.
Proposed Sec. 26.606(a)(5) would require that the FFD policy
statement describes the individual's responsibilities to report for
work in a physiological and psychological condition that enables the
safe and competent performance of assigned duties and responsibilities
and inform a licensee- or other entity-designated representative when
the individual determines that this cannot be accomplished.
Proposed Sec. 26.606(b) would require licensees and other entities
implementing a FFD program in accordance with subpart M of part 26 to
establish, implement, and maintain written procedures for their FFD
programs. This requirement would be equivalent to that in Sec.
26.403(b) of subpart K.
Proposed Sec. 26.606(b)(1) would establish requirements for a
subpart M of part 26 FFD program in which the licensee or other entity
implements a drug and alcohol testing program. This provision would be
equivalent to the requirements in current Sec. 26.403(b)(1) of subpart
K, but Sec. 26.606(b)(1)(i) through (iv) proposes additional clarity
and specificity that licensees and other entities must detail in their
procedures to address new testing methods in subpart M of part 26 that
are not permitted under the current part 26 framework. Clarity and
specificity in procedural instructions would support consistent program
implementation, which protects all individuals subject to the program.
Proposed Sec. 26.606(b)(1)(iv) would require that if the licensee
or other entity elects to use the HHS Guidelines for the conduct of
drug testing, the FFD program procedures must include the name of the
specific HHS Guideline and revision being implemented by the licensee
or other entity and a description of the specific sections in the
guideline that are being implemented, including specimen collections,
drug testing, laboratory procedures, and evaluation of test results.
This requirement would help ensure the following: the validity and
accuracy of drug testing because the specimens would be subject to
laboratory testing that has been certified by the HHS; protection of
worker rights equivalent to the privacy, information, and due process
protections afforded to Federal workers under the HHS Guidelines
because the HHS Guidelines are used in the Federally mandated drug
testing programs; consistency in program implementation because all
individuals subject to the FFD program would be subject to the same
collection,
[[Page 86964]]
testing, and evaluation processes; and FFD program effectiveness
because the effectiveness of the HHS Guidelines have been verified by
HHS's National Laboratory Certification Program (NLCP). Detailed
procedures would enhance MRO and FFD program personnel reviews of
individual test results because instructions would be provided for, in
part, the evaluation of specific test results (e.g., positive,
negative, biological markers), the conduct of additional testing for
invalid or dilute specimens, and the assessment of subversion attempts
(e.g., adulterated or substituted). This would benefit FFD program
effectiveness and help prevent misunderstanding of program requirements
and processes.
Proposed Sec. 26.606(b)(2) would require licensees and other
entities to include in their written procedures the immediate and
follow-up actions that would be taken, and the procedures that would be
used, in certain situations specified in proposed Sec. 26.606(b)(2)(i)
through (vi). Proposed Sec. 26.606(b)(2) would be equivalent to the
requirements in current Sec. 26.403(b)(2), which provides the same
requirement under an FFD program for construction for part 50 or 52
licensees and other entities. This would help ensure the effectiveness
of the FFD program and its consistent implementation, because part 53
licensed facilities would be implementing procedures to address the
same requirements and with individuals who would understand what is
expected of them no matter what part 53 facility they were assigned.
The situation specified in proposed Sec. 26.606(b)(2)(i) would
arise when individuals subject to the FFD program have been involved in
the use, sale, or possession of illegal substances, illegal drugs, or
illicit substances. This provision would be equivalent to current Sec.
26.403(b)(2)(i), except that the phrase ``illegal drugs'' would be
replaced with ``illegal substances, illegal drugs, or illicit
substances.'' Illegal substances would include legal substances used in
a manner inconsistent with Federal or State law.
The situation specified in proposed Sec. 26.606(b)(2)(ii) would
arise when individuals who are subject to the FFD program are impaired
by any substance or the consumption of alcohol as determined by
behavioral observation or a test that measures blood alcohol
concentration, as defined in Sec. 26.5. Except for a few differences,
this provision would be equivalent to current Sec. 26.403(b)(2)(ii) of
subpart K. The NRC would not include the phrases ``to excess'' and
``accurately'' in proposed Sec. 26.606(b)(2)(ii). Subpart M of part 26
is a performance-based framework that focuses on impaired human
performance, and for alcohol, impairment is determined by behavioral
observation or by blood alcohol concentrations exceeding the limits in
Sec. 26.103, ``Determining a confirmed positive test result for
alcohol,'' using an evidentiary breath testing (EBT) device for alcohol
(not whether an individual drank ``to excess''). If impairment is
determined by an individual's behavior, it must be based on
physiological indications of alcohol impairment. These indications are
well established in medical, clinical, and law enforcement
organizations, and could be used by the licensee or other entity
through its procedures and training.\11\
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\11\ By ``well established'' the NRC means that there are
Federal, State, and non-governmental organizations with reputable
and scientifically based resources available for a licensee or other
entity to use in its procedures or training to inform individuals of
the physiological indications of alcohol impairment or intoxication.
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The NRC would include the phrase ``illegal substances, illegal
drugs, and illicit substances'' in proposed Sec. 26.606(b)(2)(ii)
based on operating experience and the terminology in current Sec.
26.23(b). There are far more substances that may cause impairment than
just drugs, drug metabolites, and alcohol. The phrase ``before or while
constructing or directing construction of safety- or security-related
SSCs'' in current Sec. 26.403(b)(2)(ii) would not be included in
proposed Sec. 26.606(b)(2)(ii) because proposed Sec. 26.606 would
apply during construction, operation, and decommissioning, if
applicable. The NRC would include the term ``behavioral observation''
in proposed Sec. 26.606(b)(2)(ii) because impairment can be visibly or
audibly observed in an individual, and individuals subject to subpart M
of part 26 would be trained in behavioral observation under proposed
Sec. 26.608.
The situation specified in proposed Sec. 26.606(b)(2)(iii) would
arise when individuals who are subject to an FFD program that includes
drug and alcohol testing attempt to subvert the testing process by
adulterating or diluting specimens (in vivo or in vitro), substituting
specimens, or by any other means. Except for one difference, this
provision would be equivalent to current Sec. 26.403(b)(2)(iii). The
NRC would include the phrase ``if drug and alcohol testing is
conducted'' to address the licensee or other entity who implements
Sec. 26.604, which does not require drug and alcohol testing. The
purpose underlying this requirement has increased in significance since
issuance of the 2008 part 26 final rule because subversion attempts
have accounted for about one-third of all FFD policy violations every
year since 2016.
The situation specified in proposed Sec. 26.606(b)(2)(iv) would
arise when individuals, who are subject to an FFD program that includes
drug and alcohol testing, refuse to provide a specimen for analysis or
refuse to follow instructions provided by FFD program personnel. Except
for two differences, this provision would be equivalent to current
Sec. 26.403(b)(2)(iv). As with proposed Sec. 26.606(b)(2)(iii), the
NRC would include the phrase, ``if drug or alcohol testing is
conducted,'' to account for an FFD program implemented under Sec.
26.604. The NRC would include the phrase ``or follow the instructions
provided by FFD program personnel'' based on an existing requirement in
Sec. 26.89(c) that the collector must inform the donor that if the
donor refuses to cooperate in the specimen collection process, then
such refusal will be considered a refusal to test and sanctions for
subverting the testing process will be imposed.
The situation specified in proposed Sec. 26.606(b)(2)(v) would
arise when individuals who are subject to an FFD program had legal
action taken relating to drug or alcohol use. This requirement would be
equivalent to current Sec. 26.403(b)(2)(v).
The situation specified in proposed Sec. 26.606(b)(2)(vi) would be
when individuals subject to an FFD program demonstrated character or
actions indicating that the individual cannot be trusted or relied upon
to perform those duties and responsibilities or maintain access to NRC-
licensed facilities, SNM, or sensitive information. This includes
character traits beyond those attributed to drug or alcohol use. This
proposal would help ensure that the licensee or other entity will
implement an FFD program designed to demonstrate compliance with the
Sec. 26.23(c) performance objective that FFD programs must provide
``reasonable measures for the early detection of individuals who are
not fit to perform the duties that require them to be subject to the
FFD program.'' An individual who is not trustworthy and reliable is not
fit to perform or direct the performance of those duties and
responsibilities or be afforded those types of access that make the
individual subject to an FFD program.
This proposed requirement also would help to align the subpart M of
part 26 BOP with the BOP implemented under Sec. 73.56(f) and proposed
Sec. 73.120 and the purpose of the IMP as described in Sec.
73.55(b)(9) and proposed
[[Page 86965]]
Sec. 73.100(b)(9).\12\ The demonstrated character and actions of an
individual can indicate whether the individual can be trusted and
relied upon to safely and competently perform assigned duties and
responsibilities or be afforded those types of access making the
individual subject to the FFD program. This holds true for any
demonstrated adverse character indication or action on- or offsite.
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\12\ The IMP must monitor the initial and continuing
trustworthiness and reliability of individuals granted or retaining
unescorted AA to a protected or vital area and implement defense-in-
depth methodologies to minimize the potential for an insider to
adversely affect, either directly or indirectly, the licensee's
capability to protect against radiological sabotage.
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The phrase ``character or actions'' would be used in proposed Sec.
26.606(b)(2)(vi) to focus on observed examples that indicate an
individual subject to subpart M of part 26 may not be fit for duty or
trustworthy and reliable. Character traits include but are not limited
to personality, temperament, honesty, carelessness, apathy, psychosis,
and commitment to safety culture. Assessment of an individual's
character should consider the potential for changes in these traits
when compared to a previous baseline. Actions would include a physical
or verbal demonstration of a character trait that could call into
question an individual's fitness, trustworthiness, or reliability. For
example, the individual does something physically, verbally, or in
writing (e.g., falsifying records, driving while impaired, or harming
or threatening to harm oneself, others, or property) that compels
another individual to conclude that the observed individual cannot be
trusted or relied upon. Unlike the background investigation and reviews
of ``character and reputation'' in Sec. 73.56(d)(6) and (k)(1)(v) and
proposed Sec. 73.120, which are principally retrospective reviews of
an individual and may be based on third-party information (i.e.,
information from individuals not subject to NRC requirements), the
``character or action'' focus of proposed Sec. 26.606(b)(2)(vi) would
be a present observation of an individual subject to the FFD program
and performed by an individual who is also subject to the FFD program.
Whether the information would be received from an individual subject to
the FFD program or someone who is not subject to the FFD program, the
licensee or other entity would need to review this information (i.e.,
determine if the information and its source are credible) to determine
whether the individual should maintain authorization.
Proposed Sec. 26.606(b)(3) would require licensees and other
entities to address in their procedures the process, including the
duties and responsibilities of FFD program personnel, to be followed if
an individual's behavior or condition raises an FFD concern. This
provision would also require a process to be conducted when credible
information is received by the licensee or other entity that the
individual is not fit for duty, trustworthy, and reliable.
With a few exceptions, proposed Sec. 26.606(b)(3) would be
equivalent to current Sec. 26.403(b)(3). Instead of the phrase ``while
constructing or directing the construction of safety- or security-
related SSCs'' in current Sec. 26.403(b)(3), the NRC would use ``on
the NRC-licensed facility'' in proposed Sec. 26.606(b)(3) because this
provision would apply during commercial nuclear plant construction,
operation, and decommissioning, if applicable, in addition to holders
of an ML as described in Sec. 26.3(f). The requirement that the roles
and responsibilities of FFD program personnel be described was
developed from current Sec. Sec. 26.4(g) and 26.31(b) and operating
experience, which has demonstrated that clear job descriptions help
ensure that individuals know who is designated by the licensee or other
entity to make decisions regarding FFD program implementation and who
can be approached when physiological or psychological help is needed.
This is principally a protection consideration afforded to individuals
subject to the FFD program.
The proposed requirement would also include two conditions not
found in current Sec. 26.403(b) that would clarify the initiation of
the fitness determination process should an individual's behavior or
condition raise an FFD concern. The phrase, ``impairment from any cause
that in any way could adversely affect the individual's ability to
safely and competently perform the individual's duties,'' would reflect
the Sec. 26.23(b) performance objective. The condition, ``the receipt
of credible information indicating that the individual cannot be
trusted or relied on to perform those duties and responsibilities
making the individual subject to this part,'' would reflect the Sec.
26.23(a) performance objective. In either case, as required by Sec.
26.23(c), the FFD program must provide reasonable measures for the
early detection of individuals who are not fit to perform the duties
that require them to be subject to the FFD program.
Proposed Sec. 26.606(b)(4) would require licensees and other
entities to have written procedures that address the operation and
oversight of an onsite or offsite collection facility. This requirement
would be equivalent to current Sec. Sec. 26.403(b) and 26.405(e) and
is developed from Sec. 26.41(b), which states that each licensee and
other entity who is subject to subpart B of part 26, shall ensure that
the entire FFD program is audited, which is part of a licensee's or
other entity's oversight of the facility, and Sec. 26.87(a), which
states that each FFD program must have one or more designated
collection sites that have all necessary personnel, materials,
equipment, facilities, and supervision to collect specimens for drug
testing and to perform alcohol testing. Having procedures for the
operation and oversight of the onsite or offsite collection facility
would enhance consistency in program implementation, protect
individuals subject to testing, and account for the flexibilities
afforded in the types of biological specimens than may be collected
under an FFD program subject to subpart M of part 26. Section
26.606(b)(4), when used with the PMRP described in Sec. 26.603(d) and
the proposed audit requirement in Sec. 26.605(a), would help maintain
FFD program effectiveness and prevent subversion attempts at facilities
that may not be under the direct day-to-day oversight of FFD program
personnel.
Proposed Sec. 26.606(b)(5) would require licensees and other
entities to have written procedures that address the fatigue management
requirements in Sec. 26.202(b), ``Procedures,'' and either Sec.
26.205(d)(3) or (d)(7).
Proposed Sec. 26.606(b)(6) would require licensees and other
entities to have written procedures that provide measures to prevent
subversion of drug and alcohol tests conducted onsite and offsite. This
proposal was developed from Sec. 26.27(c)(1).
Proposed Sec. 26.607, ``Drug and alcohol testing,'' would
establish drug and alcohol testing requirements for licensees and other
entities implementing proposed Sec. 26.604, at their discretion, and
licensees and other entities implementing proposed Sec. 26.605. Except
for a few differences, proposed Sec. 26.607 would be equivalent to
current Sec. 26.405, which requires licensees and other entities
implementing an FFD program under subpart K of part 26 to have a drug
and alcohol testing program that demonstrates compliance with the
requirements in Sec. 26.405(b) through (g). The differences are
commensurate with the risk consequences presented by a part 53-licensed
facility as compared to a part 50 or 52 nuclear power plant. These
proposed requirements would improve flexibility in the conduct of
[[Page 86966]]
drug and alcohol testing while maintaining protections afforded to
individuals subject to the FFD program.
Proposed Sec. 26.607(a) would require licensees and other entities
to obtain a split specimen for all drug tests using oral fluid or urine
for all test conditions in Sec. 26.607(b), (h) and (j). Neither
current subpart K nor current subparts B or E of part 26 require a
split specimen. However, the majority of the LWR fleet uses split
specimens for drug testing and commercially available drug screening
products use a split specimen technique. Since publication of the 2008
part 26 final rule, the HHS has issued guidelines for urine and oral
fluid that require split specimens, and the draft proposed HHS
Guidelines for hair requires split specimens, as well.
The required use of a split specimen process would protect the
individual because, upon a donor-alleged discrepant or questionable
test result, the donor may provide permission to test the split
specimen (specimen B) in an effort to refute the laboratory test
results for specimen A. The requirement also would enable the MRO to
direct laboratory testing of specimen B if specimen A were invalid;
though the NRC expects specimens becoming invalid at the laboratory to
be a rare occurrence as testing would be conducted in HHS-certified
laboratories with trained collectors. In the event that a specimen is
determined to be invalid, then the occurrence would likely warrant
further investigation by the MRO and laboratory to identify the cause.
This protocol would be equivalent to the special analysis testing in
current Sec. 26.163(a)(2) for dilute specimens in that additional
laboratory analysis is performed because of a questionable test result.
If a split specimen is tested by an HHS-certified laboratory, then
the test result from specimen B must be used as part of the
determination for an FFD policy violation as required by Sec.
26.185(n), ``Evaluating results from a second laboratory.'' However,
this is not to say that the test results from specimen A should be
discarded. Since the HHS-certified laboratory should report all test
results from all specimens tested to the MRO, like the information
described in Sec. 26.169, ``Reporting results,'' test result
differences between specimens A and B can be used to inform the MRO as
to what should be reported to the licensee or other entity to either
facilitate medical or clinical assistance for the individual, inform an
FFD-policy violation determination, or both.
The proposed Sec. 26.607(a) requirement would also state that if
the licensee or other entity elects to use a POCTA device for screening
during random testing or portal area monitoring (e.g., pre-access
screening), a split specimen would not need to be taken. The reason for
this exception would be that the requirements in Sec. 26.607(h)(4)
establish the process to be implemented when a screening test indicates
a presumptive positive, adulterant, or a discrepant biological marker,
if applicable. This process includes collecting and testing a specimen
for analysis at an HHS-certified laboratory.
Proposed Sec. 26.607(b) would require the licensee or other entity
to subject individuals identified in Sec. 26.202 to drug and alcohol
testing under the five conditions listed in Sec. 26.607(b)(1) through
(5). Proposed Sec. 26.607(b) would be equivalent to current Sec.
26.405(c).
Proposed Sec. 26.607(b)(1) would require pre-access testing
similar to current Sec. 26.405(c)(1), which requires testing before
assignment to construct or direct the construction of safety- or
security-related SSCs. Unlike current Sec. 26.405(c)(1), the proposed
requirement would not include the phrase, ``construct or direct the
construction of safety- or security-related SSCs,'' because, for
licensees or other entities under part 53, the pre-access test
condition applies to construction, operation, and decommissioning, if
applicable, to help inform a licensee's or other entity's authorization
determination. The proposal also would use ``pre-access'' instead of
``pre-assignment,'' which is used in current Sec. 26.405(c)(1).
A pre-access test would require the collection of an oral fluid or
a urine specimen no more than 14 days before the individual is granted
unescorted access. Although this change has roots in the 2008 part 26
final rule, which reduced the period within which pre-access testing
must be performed from 60 days to 30 days or less, the 14-day proposal
is based on three lessons learned from operating experience.
First, the 14-day period would be a large enough window of time to
collect the specimen and evaluate test results because licensees or
other entities typically receive laboratory test results within 5
business days of laboratory receipt of the biological specimen. At the
same time, the 14-day period would be small enough to help ensure that
the test results are representative of the individual's forensic
toxicology before being granted authorization.
Second, the 14-day window would enable the licensee or other entity
to conduct an unannounced pre-access drug and alcohol screening using a
hair specimen or a POCTA. This would help prevent an individual from
attempting to subvert the drug and alcohol test by temporarily
abstaining from drug or alcohol abuse or adulterating or substituting
their specimen to obtain a non-positive test result.
Third, the NRC does not expect licensees and other entities
licensed under part 53 to have the large and periodic influxes of
individuals (either licensee employees or C/Vs) that LWRs have to
support facility operation, maintenance, engineering design changes, or
nuclear refueling. Therefore, these licensees or other entities would
not be periodically challenged to in-take a large workforce within the
proposed 14-day pre-access testing window.
Proposed Sec. 26.607(b)(2) would require the licensee or other
entity to conduct random drug and alcohol testing of all individuals
subject to the FFD program. With one exception, this proposed
requirement would be equivalent to current Sec. 26.405(b). Section
26.405(b) gives licensees and other entities that implement an FFD
program subject to subpart K of part 26 the option to impose random
drug and alcohol testing. Proposed Sec. 26.607(b)(2) would not offer
that option because subpart M of part 26, unlike subpart K, would not
allow a licensee or other entity to implement a fitness monitoring
program under current Sec. 26.406 instead of a random testing program.
The principal reasons for not allowing this flexibility would be that
no licensee or other entity has ever implemented a fitness monitoring
program (i.e., there is no operating or regulatory experience on which
to judge the effectiveness of a fitness monitoring program) and the
proposed subpart M framework already uses behavioral observation to
help ensure FFD program effectiveness. Supplementing the proposed Sec.
26.609 BOP with an additional observation technique (i.e., the fitness
monitoring program) would not result in a level of deterrence or
detection equivalent to that which would be obtained through behavioral
observation and random drug and alcohol testing.
Proposed Sec. 26.607(b)(2)(i) through (v) would provide specific
requirements for the conduct of a random testing program. These
paragraphs would be equivalent to Sec. 26.405(b)(1) through (4),
although with a few differences. The similar provisions would be
proposed in Sec. 26.607(b)(2)(i), (b)(2)(iii), and (b)(2)(iv).
The differing provisions would include proposed Sec.
26.607(b)(2)(ii), which would refer to an ``FFD program procedure''
instead of the reference to an ``FFD program policy'' in Sec.
26.405(b)(2) because procedures
[[Page 86967]]
contain the instructions that implement FFD program requirements, but
the FFD policy need not contain specific instructions. Section
26.607(b)(2)(ii) would also require individuals who are selected for
random testing to report to the onsite collection site, as opposed to
the collection site in Sec. 26.405(b)(2) because alcohol metabolism
necessitates a relatively timely alcohol test. This change is also
proposed because the NRC expects that part 53 licensees and other
entities may use a combination of onsite (for random, for-cause, and
post-event testing) and offsite (for pre-access, post-event, and
follow-up testing) collection facilities for drug and alcohol testing
and may have to afford reasonable accommodation to certain individuals,
which would add complexity in the licensee's or other entity's
procedurally determined time period in which an individual must report
to the collection facility.
Another difference from Sec. 26.405(b) would be proposed Sec.
26.607(b)(2)(v), which would establish the random testing rate for the
population of individuals subject to testing. Subpart K of part 26 does
not establish a random testing rate. The proposed requirement would be
equivalent to current Sec. 26.31(d)(2)(vii), which requires that the
sampling process used to select individuals for random testing provides
that the number of random tests performed annually is equal to at least
50 percent of the population that is subject to the FFD program. The
NRC would revise that slightly for proposed Sec. 26.607(b)(2)(v) to
require a 50 percent random testing rate for the licensee employee
population and a 50 percent random testing rate for the C/V population.
The NRC proposes this change for two reasons.
First, although operating experience has demonstrated that Sec.
26.31(d)(2)(vii) helps provide reasonable assurance that individuals
are fit for duty and trustworthy and reliable through the detection and
deterrence of substance abuse, this same operating experience
demonstrates that, on many occasions, the C/V population has been
tested at a rate lower than 50 percent, even though this population
results in the majority of all FFD policy violations. This bias occurs
because C/Vs are available for testing only during short periods of
time or periodically throughout the year, whereas licensee employees
are essentially always available for a test.
A second reason why the NRC is proposing a different 50 percent
random testing protocol than in the current part 26 requirements is
that the flexibilities afforded to part 53 licensees or other entities
in subpart M of part 26 are not afforded to licensees or other entities
that must implement an FFD program under subparts A through I, N, and O
of part 26. These flexibilities include enabling the use of a POCTA
device to screen individuals during the random testing process and the
use of offsite collection facilities for pre-access testing. The
potential reduction in FFD program effectiveness caused by licensee or
other entity implementation of these options would be offset by subpart
M requirements that mitigate possible challenges to the FFD program,
such as the 50 percent random testing rate for the licensee employee
population and 50 percent random testing rate for the C/V population.
Proposed Sec. 26.607(b)(3) would require for-cause testing
equivalent to that used in current FFD programs implementing Sec.
26.405(c)(2). The NRC would require for-cause testing, like random
testing, to be conducted onsite to ensure that the test is conducted as
soon as reasonably practicable. This is an important consideration when
for-cause testing for alcohol or using oral fluid for drug screening or
testing because human metabolism continually lowers the concentrations
of the drugs, drug metabolites, and alcohol perhaps to concentrations
lower than the initial or confirmatory testing cutoffs. Additionally,
for facilities that are sited in geographically remote locations, an
offsite collection facility might be too far away or not readily
accessible.
Proposed Sec. 26.607(b)(4) would require post-event testing in a
manner equivalent to current Sec. 26.405(c)(3) with a few adjustments.
For part 53 licensees or other entities, the NRC proposes post-event
testing under two conditions: events involving human errors that may
have caused or contributed to the events (proposed Sec.
26.607(b)(4)(i)), and events not involving human error that result in
adverse health consequences or damage to any safety- or security-
related SSC (proposed Sec. 26.607(b)(4)(ii)). The word ``significant''
would not be used in Sec. 26.607(b)(4)(ii)(A) to describe the
``illness or personal injury'' as used in Sec. 26.405(c)(3)(i) because
Sec. 26.607(b)(4)(ii)(A) would describe which illnesses or injuries
are covered. Proposed Sec. 26.607(b)(4)(ii)(B), unlike Sec.
26.405(c)(3)(ii), would not use the word ``significant'' to describe
the damage to safety- or security-related SSCs because any damage to
safety- or security-related SSCs would require testing within four
hours of the event unless immediate medical intervention precludes the
conduct of the test on the individual(s) who caused or contributed to
the event. Proposed Sec. 26.607(b)(4)(ii)(B) also would not use the
word ``construction'' as in Sec. 26.405(c)(3)(ii) because Sec.
26.607(b)(4) would apply to construction, operation, and
decommissioning, if applicable.
Proposed Sec. 26.607(b)(4)(i) would require the licensee or other
entity to define in its procedures the terms ``human error'' and
``event.'' These terms may take on various meanings and they are not
defined in the current or proposed rule, so the licensee or other
entity would be required to describe or define these terms to help
ensure consistent implementation of subpart M of part 26 and that the
post-event test condition would be consistently applied to all
individuals subject to the FFD program. The Sec. 26.405(c)(3)(i)
requirement that ``the event is recordable under the Department of
Labor standards contained in 29 CFR 1904.7, and subsequent amendments
thereto,'' would not be carried over to proposed Sec. 26.607(b)(4).
Instead, the NRC proposes to prescribe the post-event test conditions
in Sec. 26.607(b)(4), in part so they would not change unless the NRC
amends the requirement.
Proposed Sec. 26.607(b)(5) would require follow-up testing. This
requirement would be equivalent to current Sec. 26.405(c)(4), although
the proposed Sec. 26.607(b)(5) would further describe follow-up
testing. The NRC proposes to describe follow-up testing as part of a
series of tests for drugs, alcohol, or both, which are performed after
an individual subject to part 26 has violated the FFD policy on
substance use or abuse, or the sale, use, or possession of illegal
drugs. Follow-up testing would be used to verify an individual's
continued abstinence from substance abuse. The NRC would not include a
reference to a follow-up plan as in Sec. 26.405(c)(4) because the
intent of a follow-up plan is to conduct a series of drug tests,
alcohol tests, or both, to verify continuing abstinence from substance
abuse. Nevertheless, individuals who violate an FFD policy on substance
use or abuse, or the sale, use, or possession of illegal drugs, should
have a follow-up plan that includes a definition of ``abstinence'' from
the medical professional prescribing the plan.
Proposed Sec. 26.607(c) would provide additional testing
requirements. This proposed requirement would be equivalent to Sec.
26.405(d) and would require implementation of select requirements from
current subpart E of part 26. The proposed requirements would govern
directly observed collections, shy bladder situations, special analysis
testing, and alcohol testing. These requirements would be necessary to
maintain FFD program
[[Page 86968]]
effectiveness equivalent to that currently implemented by the LWR
fleet.
Proposed Sec. 26.607(c)(1) would require validity testing and
establish the minimum panel of drugs and drug metabolites to be tested.
This panel would be the same as those in Sec. Sec. 26.31(d)(1) and
26.405(d) because, based on operating experience from LWR FFD program
implementation, this panel has been determined to contribute to a
licensee or other entity satisfying the FFD performance objectives in
Sec. 26.23(a) through (d).
Proposed Sec. 26.607(c)(1) would differ from Sec. 26.405(d)
because it would require testing of oral fluid and urine specimens for
validity, including at least one biological marker (developed from an
HHS Guidelines provision) and one adulterant (equivalent to current
validity testing for urine specimens in part 26). Section 26.405(d)
requires that urine specimens collected for drug testing be subject to
validity testing. The addition of oral fluid validity testing is
important because, just as there are publicly available kits to subvert
a urine drug test, kits that may be used to subvert a drug test that
uses oral fluid as a biological specimen are also readily available.
Proposed Sec. 26.607(c)(2) would include requirements that already
exist in the part 26 framework that provide protections for individuals
subject to the FFD program and contribute to testing effectiveness when
collecting and assessing a urine specimen. Specifically, current Sec.
26.115, ``Collecting a urine specimen under direct observation,''
describes the exclusive grounds for performing a directly observed
collection and the process to be followed to protect the privacy of the
individual. Section 26.119, ``Determining `shy' bladder,'' establishes
the process to be followed when a donor is not able to produce a
sufficient amount of urine for testing, and Sec. 26.163(a)(2) requires
special analysis testing when a specimen is dilute to help prevent a
subversion attempt.
Proposed Sec. 26.607(c)(3) would require implementation of all the
current alcohol testing requirements in Sec. 26.91, ``Acceptable
devices for conducting initial and confirmatory tests for alcohol and
methods of use,'' through Sec. 26.103, ``Determining a confirmed
positive test result for alcohol.'' Using the same alcohol testing
framework for parts 50, 52, 70, and 53 licensees and other entities
would provide for regulatory consistency, protections for individuals
subject to the FFD program (e.g., the quality controls and verification
applied to the EBT device), and FFD program effectiveness (e.g.,
accuracy of test results). For alcohol testing, unlike drug testing,
there is a preponderance of evidence that correlates blood alcohol
concentrations to impairment and intoxication. Furthermore, FFD
performance data has demonstrated that the time-dependent alcohol
cutoffs in Sec. 26.103 have increased the detection of individuals who
are under the influence of alcohol. For these reasons, the current
alcohol requirements in part 26 are proposed for FFD programs under
subpart M.
Proposed Sec. 26.607(c)(4) would establish additional testing
requirements. This proposal would be equivalent to current Sec.
26.405(f) for facilities licensed under part 53 for the conduct of drug
testing. Unlike Sec. 26.405(f), proposed Sec. 26.607(c)(4) would not
reference validity screening and initial drug and validity tests at
licensee testing facilities as this would be required in proposed Sec.
26.607(c)(1). Another minor difference between Sec. 26.405(f) and
proposed Sec. 26.607(c)(4) would reflect the requirement in subpart M
of part 26 to use an HHS-certified laboratory for all biological
specimens collected and not just for urine specimens.
Consistent with Sec. 26.405(f), proposed Sec. 26.607(c)(4) would
require the use of an HHS-certified laboratory for all test conditions
listed in Sec. 26.607(b), MRO-directed tests, and the testing of a
split specimen. Further, HHS-certified laboratory test results using
urine or oral fluid would be required for the issuance of an FFD policy
violation and part 26-required sanction.
All drug testing would need to be performed at an HHS-certified
laboratory to help ensure FFD program effectiveness and to protect the
donor from a false positive test result and an unwarranted FFD policy
violation. The donor would be protected because laboratory procedures
for specimen accessioning, testing, custody and control, and evaluation
of test results and the training and qualification of laboratory
personnel are evaluated by HHS as part of the NLCP. This provides
assurance that the drug testing results are accurate and attributed to
the donor. Urine, oral fluid, and hair specimens may also be screened
and tested for drugs and alcohol as described in Sec. 26.607. Drug and
alcohol screening results obtained from urine and oral fluid specimens
collected and analyzed using a POCTA device and screening results
obtained from a hair specimen or a portal monitor may only be used as
potentially disqualifying information for a licensee's or other
entity's authorization determination (i.e., used to assess the fitness,
trustworthiness, and reliability of the individual). These screening
results may not be used for the administration of an FFD policy
violation and sanction, except as proposed Sec. Sec. 26.607(i)(3) and
26.610 for subversions, as defined in Sec. 26.5, of the drug and
alcohol screening process.
There are three phrases or requirements in Sec. 26.405(f) that the
NRC does not propose to use in Sec. 26.607(c)(4). The first is the
phrase, ``consistent with its standards and procedures for
certification,'' regarding the operation of an HHS-certified
laboratory, because the laboratory would not be HHS-certified if it
were not following ``its standards and procedures for certification.''
The second is the requirement that urine specimens that yield positive,
adulterated, substituted, or invalid initial validity or drug test
results must be subject to confirmatory testing by the HHS-certified
laboratory, except for invalid specimens that cannot be tested. This
requirement would not be used because, under subpart M of part 26,
licensees or other entities would be required to use an HHS-certified
laboratory. For a laboratory to be HHS-certified, it must follow the
HHS Guidelines and include procedures that describe when a specimen
cannot be tested. Lastly, the Sec. 26.405(f) requirement that other
specimens that yield positive initial drug test results must be subject
to confirmatory testing by a laboratory that demonstrates compliance
with stringent quality control requirements that are comparable to
those required for certification by the HHS, would not be used because
subpart M of part 26 would require the use of an HHS-certified
laboratory.
Proposed Sec. 26.607(c)(4) would require the licensee or other
entity to contract with a primary and backup HHS-certified laboratory.
This provision would help ensure that specimens are processed and
tested to maintain FFD program effectiveness should the primary
laboratory be unable to perform specimen testing. This would help
maintain protections afforded to individuals subject to the FFD program
(e.g., should the donor or MRO request testing of the split specimen, a
different laboratory could be used). This requirement also would state
that the primary and backup laboratories must have a different
certifying scientist. Having a back-up HHS-certified laboratory and a
different certifying scientist would benefit the program and donor
because the drug testing instruments, technicians, and certifying
scientist would be independent of the
[[Page 86969]]
primary laboratory testing and review process. The back-up HHS-
certified laboratory may be of the same corporate entity as the primary
laboratory.
Proposed Sec. 26.607(c)(4) would also state that the laboratory
would be subject to inspection or audit by the licensee or other entity
and that records and documents must be provided and/or able to be
photocopied and removed from the premises to support the inspection or
audit. This requirement would be equivalent to current Sec. 26.41(d)
except that laboratories would not be able to limit the use and
dissemination of documents copied or taken from the laboratory by a
licensee or other entity. This is necessary to ensure the continuing
effectiveness of FFD programs, because NLCP findings and audit results
could adversely impact FFD program effectiveness. Pertinent information
includes and should not be limited to NLCP-identified weaknesses (e.g.,
custody and control, accessioning, instrumentation, procedures,
training, supervision, review of test results, and resolution of
previously identified corrective actions) that may impact the
effectiveness of FFD programs.
Proposed Sec. 26.607(d) would help protect the donor from mistakes
made during the drug and alcohol testing processes and help ensure FFD
program effectiveness. The rule would require the licensee or other
entity to protect the individual's privacy and the integrity of the
specimen and to implement quality controls to ensure that test results
are valid and attributable to the correct individual. This requirement
would be equivalent to the first sentence of current Sec. 26.405(e),
except that the word ``stringent'' was removed from the phrase
``stringent quality controls,'' because the word ``stringent'' is not
defined.
Proposed Sec. 26.607(e) would describe the requirements for
licensees and other entities that use offsite collection facilities.
Consistent with current Sec. 26.405(e), a licensee or other entity
would be able to conduct specimen collections and alcohol testing at a
local hospital or other facility. Unlike Sec. 26.405(e), proposed
Sec. 26.607(e) would not restrict licensees and other entities to use
hospitals and other facilities that meet the requirements in 49 CFR
part 40, ``Procedures for Transportation Workplace Drug and Alcohol
Testing Programs,'' because subpart M of part 26 is intended to provide
flexibilities beyond those in the current part 26 framework. Licensees
and other entities may use these Department of Transportation
requirements to inform their procedures under Sec. 26.606(b)(1) as
long as the procedures do not conflict with the requirements in part 26
or the HHS Guidelines.
Proposed Sec. 26.607(e) would also require licensees and other
entities to audit offsite collection facilities before their use and
biennially to confirm that the facility procedures are comparable to
those described in subpart E of part 26 or the HHS Guidelines for urine
and oral fluid. This proposed requirement is based on current Sec.
26.41(a) and (b). The Sec. 26.607(e) audit requirement would be a
program effectiveness consideration because offsite collection
facilities may not require vigilance of their collectors (e.g.,
identification of subversion attempts), diligence in the protection of
worker rights (e.g., privacy and specimen custody and control), or
procedural compliance.
The offsite facility used by a licensee or other entity under
proposed Sec. 26.607(e) would have to be licensed to conduct specimen
collections and perform alcohol testing, and be audited, by the State
or a State-designated entity. This requirement would help provide
assurance of adequate collection facility performance and may help
reduce the burden on the licensee or other entity and the collection
facility. Crediting a State audit (or State licensure, oversight, or
regulation) is established in Sec. Sec. 26.4(i)(4) and (j),
26.91(e)(5), 26.153(f)(1), and 26.183(a).
Proposed Sec. 26.607(f) would provide the requirements for initial
drug testing. This provision would be equivalent to Sec. 26.405(f)
except to account for the alternative biological specimens that may be
tested under subpart M of part 26. For the testing of all biological
specimens, the licensee or other entity under part 53 would be required
to use a device that employs an immunoassay screening technique, or an
alternative technology that the licensee or other entity has
incorporated into its FFD program through the Sec. 26.603(e) change
control process, that demonstrates compliance with the requirements of
the U.S. Food and Drug Administration (FDA) for commercial
distribution. Examples of alternative technologies include liquid or
gas chromatography and mass spectrometry. Licensees and other entities
would use the Sec. 26.603(e) change control process to evaluate and
document a change to their collection and analysis procedures to enable
the use of a better or perhaps more cost-effective collection and/or
testing technology. Another difference from Sec. 26.405(f) would be
changing the word ``urine'' in Sec. 26.405(f) to ``biological
specimens'' in Sec. 26.607(f). Lastly, proposed Sec. 26.607(f) would
include the phrase ``discrepant biological marker'' as a drug screening
result that must be analyzed by an HHS-certified laboratory and
evaluated by the MRO to help inform the MRO's determination of a
subversion attempt.
Proposed Sec. 26.607(g) would enable a part 53 licensee to use
oral fluid as a biological specimen for testing. This requirement would
be equivalent to Sec. 26.31(d)(5), which enables the MRO to conduct
drug and alcohol testing using alternative methods, and Sec. 26.405,
which does not preclude the use of oral fluid specimens for FFD
programs that implement subpart K of part 26 requirements. In order to
provide assurance that drug testing is effective and protects the
worker, Sec. 26.607(g) would require that the licensee's or other
entity's procedures incorporate the HHS Guidelines or the requirements
in part 26 for the conduct of urine or oral fluid testing.
The proposed Sec. 26.607(g) requires that the oral fluid
collection device must have received premarket approval from the FDA
and must not expire before laboratory testing. Also, the drugs, drug
metabolites, initial and confirmatory testing cutoffs, and biological
markers, if applicable, must be those established by HHS for oral fluid
drug testing and the alcohol cutoffs in part 26. If they are not
established by HHS or the NRC for the paneled drugs and drug
metabolites, then they would be determined and documented by a forensic
toxicologist review. This forensic toxicologist review would help
ensure that the device accurately tests for the drug, drug metabolite,
biological markers, adulterants, and/or alcohol and that the results
from the device are comparable to those established in the HHS
Guidelines for oral fluid testing.
Proposed Sec. 26.607(h)(1) and (2) would enable the use of a POCTA
device during the random and pre-access testing processes. These
requirements are adopted from Sec. 26.97, ``Collecting oral fluid
specimens for alcohol and drug testing,'' and Sec. 26.405(f), which
does not preclude the use of oral fluid testing. To use a POCTA device
for urine, oral fluid, or other biological indicators (breath, sweat,
etc.), a forensic toxicology review would be required to ensure that
the device is forensically effective. If the POCTA device is
forensically effective, then the donor would be reasonably protected
from a false positive test result, the licensee or other entity would
be reasonably protected from false negative test results, and the FFD
program would remain effective. For a POCTA device to be forensically
effective, the forensic toxicologist would need to document an
evaluation that the performance of the
[[Page 86970]]
POCTA device must be comparable to the requirements in Sec. 26.161(b)
for a urine specimen or the procedures in the HHS Guidelines for urine
or oral fluid, as implemented by the licensee or other entity through
its procedures.
The use of POCTA for oral fluid and urine specimens for the pre-
access and random testing processes would be acceptable because
individuals in the pre-access process would be subject to an oral fluid
or urine specimen collection and possible drug screening using a hair
specimen, which are both required to be sent to an HHS-certified
laboratory. For random testing, the individual would have also been
granted authorization under the AA and FFD requirements and have been
subject to behavioral observation and physical protection screening
(e.g., verification of identify, and screening for explosives and
contraband).
Proposed Sec. 26.607(h)(3) would require that procedures be
developed that ensure the effectiveness of the POCTA collection
process, assessment of the screening results, and prevention of
subversion attempts. This requirement would be equivalent to current
Sec. 26.403(b)(1) and would help ensure protections afforded to
individuals subject to the FFD program and program effectiveness. The
subpart M of part 26 framework enables the use of POCTA for random
screening of individuals for any part 53 facility, so the licensee or
other entity should exercise due diligence and implement risk
management strategies to ensure the efficacy of random screening and
its contribution to an effective FFD program.
Proposed Sec. 26.607(h)(4) would provide that an individual donor
who screens positive (or whose specimen is invalid or indicates a
discrepant biological marker or adulterant) is removed from all duties
and responsibilities making the donor subject to subpart M of part 26.
Under proposed Sec. 26.607(h)(4)(i), the donor then would be
immediately subject to a drug and alcohol test that provides quantified
confirmatory test results from which an FFD policy violation may be
issued. Similar to other requirements for specimen collections, except
for biological specimens analyzed by a passive detection system, the
licensee or other entity would be required to implement procedures that
ensure that all specimens collected are uniquely assigned to the donor
(i.e., procedures that provide for custody and control of the
specimen). If the individual shows signs of impairment during the POCTA
process, proposed Sec. 26.607(h)(4)(ii) would require the temporary
removal of the individual's authorization until the MRO reviews the
laboratory test result(s), and interviews the individual, and a
determination of fitness finds that authorization may be restored.
Section 26.607(h)(4) is equivalent to Sec. 26.77(b) and was informed
by the requirements in Sec. Sec. 26.419, 26.75(c) and (d), and
26.185(c).
Proposed Sec. 26.607(i) would enable the collection of hair
specimens for drug testing to supplement pre-access testing that uses
urine or oral fluid specimens. Hair testing would be a new feature in
the part 26 framework. The NRC proposes to permit the use of hair
testing for only Schedule I or II drugs or their metabolites to inform
a licensee's or other entity's determination whether the individual is
trustworthy and reliable. For example, if an individual stated no prior
use of illegal drugs or potentially addictive habits, a hair screening
test could be performed during the pre-access process to ascertain the
validity of the individual's statement. However, if the HHS-certified
laboratory communicates a laboratory-confirmed positive test result, an
FFD policy violation may not be administered. This laboratory
information must be treated as potentially disqualifying FFD
information, unless the individual subverts the screening process, in
which case a permanent denial of authorization must be issued under
proposed Sec. 26.610. To provide assurance of testing effectiveness
and protections afforded to individuals subject to the FFD program,
proposed Sec. 26.607(i) would require that an HHS-certified laboratory
must be used to analyze the hair specimen, a forensic toxicologist must
review the licensee's or other entity's hair screening process, the
test kit must be cleared by the FDA, and hair screening must be
conducted in accordance with the HHS Guidelines. The forensic
toxicologist review would be necessary if the panel of drug or drug
metabolites to be tested and their cutoffs are not established by HHS
or the NRC for hair.
Proposed Sec. 26.607(j) would allow the use of portal area
screening for drugs, alcohol, or both. This provision would result in a
substantial contribution to a licensee or other entity satisfying the
Sec. 26.23 performance objectives by helping ensure that 100 percent
of all individuals who arrive at the NRC-licensed facility to perform
or direct those duties and responsibilities or maintain those types of
access making them subject to the FFD program are fit for duty and
deterred from arriving onsite in a physiological condition that may be
adverse to safety and security. Additionally, screening could be
conducted when an individual exits the NRC-licensed facility to provide
assurance that substance abuse had not occurred on the site (see Sec.
26.23(d)). The screening device could be electronically linked to
temporarily prevent ingress or egress and could automatically inform
licensee- or other entity-designated officials of the portal area
alarm. The proposed requirement would enable the licensee or other
entity to use innovative technologies to maintain FFD program
effectiveness when their PMRP compels the licensee or other entity to
implement mitigative strategies to maintain program effectiveness. The
use of portal screening technologies may also represent cost savings
because, for NRC-licensed facilities that have small staff sizes or are
geographically remote, passive drug and alcohol screening technologies
could be an innovative alternative to a random testing program,
although the license or other entity would need to request and receive
an exemption.
Proposed Sec. 26.607(j) would also provide that if the portal area
screening instrument detects a substance that exceeds the instrument's
established setpoint, the individual would be tested with either a
collection kit that must be analyzed by an HHS-certified laboratory or
a POCTA. This situational screening would be equivalent to a for-cause
test. The requirements would not allow an individual to be rescreened
by the portal area screening instrument following an initial screening
detection that exceeded an established setpoint in order to prevent a
subversion attempt. Similar to other drug and alcohol testing
technologies enabled for use by subpart M of part 26, a forensic
toxicology review would be required before using passive screening
technology to help ensure the effectiveness of the instrument by
protecting against false positive or negative screening results, which
would place an unwarranted burden on the individual, licensee, or other
entity. These instruments and alcohol screening devices, already in the
marketplace, may also be used to determine true identity to facilitate
implementation of the FFD BOP, which may be very practicable at
facilities that operate with small staff sizes.
Proposed Sec. 26.607(k) would enable the use of a blood specimen
for drug, alcohol, or other testing for certain medical conditions as
determined by the licensee- or other entity-designated MRO. This
requirement would be equivalent to current Sec. 26.31(d)(5). The use
of a licensee- or other entity-designated MRO and not one designated by
a third party, such as an MRO
[[Page 86971]]
employed by an offsite specimen collection facility, is important
because the MRO must be familiar with the subpart M of part 26
requirements. To help ensure testing effectiveness and protect the
worker, the blood test would need to be conducted by a laboratory that
demonstrates compliance with quality control requirements that are
comparable to those required for certification by the HHS, such as a
hospital or clinic certified by the State, Commonwealth, or territory.
Proposed Sec. 26.607(l) would require licensee and other entities
to use a Federal custody-and-control form (CCF) approved by the OMB for
the collection and packaging of a hair, oral fluid, or urine specimen.
This proposed requirement is based on the CCF documentation
requirements in current subpart E of part 26 because subpart K of part
26 does not require the use of a CCF under Sec. 26.117(e).
Additionally, when using a POCTA device, the licensee or other entity
would be required to implement a licensee- or other entity-approved and
-maintained procedure that ensures the reliability of the tracking,
handling, and storage of a specimen from the point of specimen
collection to final disposition of the specimen and the reliability of
an identification system to uniquely assign the specimen to the donor.
Both requirements would help protect the worker by helping ensure chain
of custody and by contributing to program effectiveness.
Proposed Sec. 26.607(m) would establish requirements for the
licensee- or other entity-designated MRO. Section 26.607(m)(1) would be
equivalent to Sec. 26.405(g), however, the word ``designated'' would
be added to the first sentence to clarify that the MRO would be
designated by the licensee or other entity, and not by a third party.
As stated with regard to proposed Sec. 26.607(k), this change would
clarify that it is the licensee's or other entity's responsibility,
through their designated MRO, to determine whether an individual is fit
for duty and trustworthy and reliable. This would be consistent with
the description of FFD program personnel in current Sec. 26.31(b) and
help provide FFD program effectiveness and protections to individuals
subject to the FFD program. The paragraph was also modified from Sec.
26.405(g) to address the determinations of FFD policy violations and
fitness required by subpart H for a part 53 licensee or other entity
that implements the FFD program described in Sec. 26.605(b).
Proposed Sec. 26.607(m)(2) would help ensure that MRO reviews are
consistent with those MRO reviews conducted at other NRC-licensed
facilities subject to part 26 and that the MRO maintains knowledge of
drug collection, testing processes and procedures, and evaluation of
testing results.
The NRC also proposes that if an MRO performed the duties and
responsibilities in Sec. Sec. 26.185 and 26.187 for at least three
continuous years in the last 10 years prior to being hired or
contracted by the licensee or other entity, then the MRO would not need
to repeat the initial training and examination requirements. The basis
for 3 years is that the MRO would have experienced three annual cycles
of evaluating drug and alcohol test results, contributed to the FFD
annual report to the NRC, experienced a refueling or maintenance
outage, understood the duties and responsibilities of individuals
subject to the FFD program to make informed determinations of fitness,
demonstrated a safety culture that helps ensure FFD program
effectiveness, and been subject to NRC inspection. The basis for 10
years is the relatively long periods between significant changes to
part 26 and the HHS Guidelines.
Proposed Sec. 26.607(m)(3) would require that the MRO attend a
medical- or clinical-based training session on a triennial basis. This
proposal was developed from Section 13.1 of the HHS Guidelines for
urine and oral fluid with two substantial differences: the HHS
Guidelines state that ``requalification training,'' including an exam,
must be conducted ``at least every 5 years from initial
certification,'' whereas the proposed Sec. 26.607(m)(3) would require
a training session every three years. The proposed requirements are
justified because changes in societal drug use or forensic toxicology
could occur more frequently than every 5 years, which could compel MROs
to attend training in areas of forensic toxicology, determinations of
fitness, or other part 26 technical areas on a more frequent
periodicity than every 5 years to improve their knowledge and
expertise.
Proposed Sec. 26.607(m)(4) would require the MRO to evaluate drug
testing results by implementing the requirements in Sec. 26.185 or the
HHS Guidelines through the licensee's or other entity's procedures.
This requirement would help ensure FFD program effectiveness and
enhance consistency across the commercial nuclear industry for the
evaluation of drug testing results. This also would help protect
individuals because they would be subject to the same evaluation
criteria. If Sec. 26.185 provides insufficient information for an MRO
to make a determination on a drug testing result (including adulterant
and discrepant biological markers), the guidance issued by a State
agency in the state in which the NRC-licensed facility is located,
Federal agency, or nationally recognized MRO training and certification
organization may be used to inform an MRO determination. This provision
would ensure that the MRO has the flexibility to inform their
evaluation of the drug testing results and fitness determination, if
necessary, considering the drug- and alcohol-related flexibilities
afforded in subpart M of part 26.
The proposed requirement would also state that an MRO need not
review a confirmed alcohol positive test result determined by an EBT
device under Sec. 26.607(c)(3)(vi) and (vii), which are equivalent to
the current requirements in Sec. Sec. 26.101 and 26.103, respectively.
The results of an EBT device are precise and accurate enough to support
the issuance of an FFD policy violation without an MRO review of an EBT
test result if the instrument demonstrates compliance with the
requirements in Sec. 26.91. The NRC acknowledges that there are
physiological conditions that may cause an abnormally high blood
alcohol concentration, such as diabetes, acid reflux, gastroesophageal
reflux disease, and perhaps certain diets (high protein and low
carbohydrates). However, operating experience has not demonstrated a
compelling need to require an MRO review of all EBT test results. For
consistency, a licensee or other entity may elect to require its MRO to
review all EBT test results when a donor communicates a testing concern
or physiological condition. If the donor has a testing concern, the
occurrence could be appealed under the proposed Sec. 26.613. If the
donor presents a physical condition to the MRO that may have caused an
elevated EBT test result, the MRO may direct an alternative testing
process (see Sec. 26.607(m)(5)) should it be medically necessary.
Proposed Sec. 26.607(m)(5) would require the licensee- or other
entity-designated MRO to determine and approve the use of oral fluid or
urine as an alternative biological specimen when the donor cannot
provide a requested specimen for testing. This proposed requirement is
equivalent to Sec. 26.31(d)(5), which enables the use of an
alternative specimen collection if a medical condition makes the
collection of the biological specimen difficult. This determination and
the retest must be completed as soon as reasonably practicable and
documented to support recordkeeping, auditing, and NRC inspection.
[[Page 86972]]
Proposed Sec. 26.607(m)(6) would require that the MRO review all
specimens screened or tested associated with a drug-related FFD policy
violation. This includes POCTA, split specimens, and all specimens
taken to resolve a discrepant condition, such as a possible subversion
attempt, impairment without a known cause, or a donor-requested or MRO-
directed retest. To resolve a discrepant condition, the MRO is
authorized to test a specimen for a biological marker, adulterants, or
additional drugs. The broad scope of this MRO evaluation would be
necessary because of the variety of different screening and testing
methods that may have been associated with the FFD policy violation.
All information learned from the conduct of part 26 drug and alcohol
screening and testing should be used in the evaluation of an
individual's trustworthiness and reliability, issuance of a sanction,
and development of a follow-up treatment and testing plan, if
administered.
Proposed Sec. 26.607(n) is equivalent to current Sec. 26.31(d)(6)
and would establish limits on the screening and testing of biological
specimens. This is a protection consideration afforded to individuals
subject to the FFD program and was not provided in subpart K of part
26. This requirement states that specimens collected under NRC
regulations may only be designated or approved for screening and
testing as described in this part and may not be used to conduct any
other analysis or test without the written permission of the donor.
Analyses and tests that may not be conducted include, but are not
limited to, deoxyribonucleic acid (i.e., DNA) testing, serological
typing, or any other medical or genetic test used for diagnostic or
specimen identification purposes.
The NRC proposes to require that no biological specimens may be
passively sampled and analyzed in a manner different than described in
subpart M of part 26 to ensure workers are protected from non-
consensual passive screening. The subpart M framework enables passive
detection of drugs and alcohol, whereas passive detection is not
afforded in subparts A through I, N, and O of part 26.
Proposed Sec. 26.607(o) is equivalent to current Sec. Sec.
26.31(b)(1)(iii)(A) and 26.89 and would require that all specimen
collections be conducted by a licensee- or other entity-designated and
-trained individual. For subpart M of part 26, this would include
onsite specimen collections, except a collection by a portal area
screening instrument in Sec. 26.607(j).
Proposed Sec. 26.608 would require licensees and other entities to
provide FFD program training to individuals subject to the FFD program.
The proposed performance-based Sec. 26.608 requirement was developed
from the prescriptive training requirements in current Sec. 26.29 and
modeled on current Sec. 50.120 and the proposed requirements in
Sec. Sec. 53.725 and 53.830 because there is no training requirement
in subpart K of part 26.
Proposed Sec. 26.608(a)(1) would require an FFD training program
that includes the licensee's or other entity's FFD policies and
procedures, including fatigue management, and the individuals' FFD
program responsibilities. Individuals who collect specimens for testing
or screening must also be trained in specimen collector duties and
responsibilities, including, at a minimum, specimen collection, custody
and control, identification and response to subversion attempts, and
privacy. The fatigue management training must include the knowledge and
abilities described in Sec. 26.202(c). For individuals specified in
Sec. 26.4, a licensee or other entity of a commercial nuclear plant
would be required to use a SAT as defined in proposed in Sec. 53.725.
These requirements are based on requirements in Sec. 26.29(a)(2), (3),
(9), and (10).
Proposed Sec. 26.608(a)(2) would require training on the BOP. This
requirement would be based on Sec. Sec. 26.29(a)(8), (9), and (10) and
26.33. The proposal would require individuals to be trained in the
detection of behaviors or conditions related to not only illegal drugs,
as in the current Sec. 26.33 BOP requirements, but also illicit drugs
and substance abuse onsite and offsite. Also, in reference to
impairment from fatigue or any cause if left unattended, the phrase in
Sec. 26.33, ``may constitute a risk to public health and safety or the
common defense and security,'' would be replaced in Sec.
26.608(a)(2)(iii) with ``could result in inattentiveness or human
errors,'' because subpart M of part 26 is focused, in part, on ensuring
individuals are fit for duty to safely and competently perform or
direct the performance of assigned duties and responsibilities.
Proposed Sec. 26.608(a)(2)(iv) focuses on training to inform
individuals that they are responsible for their own conduct, as well as
observing others. Specifically, individuals would be trained to
recognize when they feel unable to safely and competently perform
assigned duties and responsibilities or act in a trustworthy and
reliable manner. The proposed training requirement and the proposed
reporting requirement in Sec. 26.606(a)(5) are in the interest of
safety and security because the individual is proactively announcing
that assistance may be necessary. This would be consistent with the
performance objectives in Sec. 26.23(b) and (c) where certain behavior
or stress conditions may be indicative of an individual not being fit
for duty, trustworthy, and reliable.
Proposed Sec. 26.608(a)(3) would help ensure that individuals
subject to the FFD program understand that FFD policy violations would
result in an FFD program sanction and that program information learned
or generated by FFD program implementation would be used to aide
licensee or other entity authorization determinations and be shared, as
requested, with other licensees or other entities subject to parts 26,
53, and 73. This proposed requirement is equivalent to Sec.
26.29(a)(1). Proposed Sec. 26.608(a)(3) would be a protection measure
afforded to individuals subject to the FFD program because they would
understand that licensees and other entities subject to parts 26, 53,
and 73 would be informed of, in part, an individual's character,
reputation, and ability to follow policies, procedures, and
instructions to safely and competently perform assigned duties and
responsibilities in a trustworthy and reliable manner. Fitness-for-
duty-related information would include drug and alcohol testing results
(not quantitative testing values), issuance of any sanctions, FFD-
determinations regarding trustworthiness and reliability, testing
programs, treatment, and other remedial or corrective action.
Proposed Sec. 26.608(b) would require individuals be trained and
receive a trainee assessment before pre-access testing and that
refresher training and trainee assessments be conducted periodically
thereafter. These requirements would be equivalent to Sec.
26.29(c)(1). However, Sec. 26.608(b) was developed from the SAT-based
training requirements in Sec. 50.120 and training elements from the
annual training requirements in Sec. 26.29(c)(2). The term ``systems
approach to training'' would have the meaning in proposed Sec.
53.725(c). A trainee assessment would be the same as in currently
required SAT-based training programs.
Proposed Sec. 26.608(c) would require licensees and other entities
to periodically evaluate their FFD training programs and revise them as
appropriate. This training focus is not required by subpart K of part
26 or Sec. 26.29 but is proposed to address the flexibilities afforded
in subpart M of
[[Page 86973]]
part 26. This section would be equivalent to Sec. 50.120(b)(3).
Proposed Sec. 26.609 would require the implementation of a BOP.
The proposed requirement would be equivalent to that in Sec. Sec.
26.33 and 26.407, ``Behavioral observation,'' and would apply during
construction, operation, and decommissioning, if applicable. Because
subpart M of part 26 would apply during decommissioning through a
licensee's IMP, proposed Sec. 26.609(a) and (b) were developed, in
part, from proposed Sec. 73.100(b)(9) and current Sec. Sec.
73.55(b)(9) and 73.56(f) to help ensure consistency in the conduct of
behavioral observation whether conducted for FFD or security purposes.
Under the FFD program, the purpose of the BOP would be to help
ensure that individuals subject to the FFD program are fit for duty and
trustworthy and reliable to perform or direct those duties and
responsibilities and maintain those types of access that make the
individual subject to the FFD program. This assurance is accomplished
by requiring each individual subject to subpart M of part 26 to be
subject to behavioral observation, and by requiring all individuals to
perform behavioral observation of others and report FFD concerns to the
licensee- or other entity-designated representative(s). The intent of
the BOP requirement is not to require that all individuals be observed
at all times by others; NRC-licensed operators, maintenance
professionals, security officers, and others routinely perform solo
operations periodically throughout the day. However, individuals must
be subject to observation while they are performing or directing the
performance of duties and responsibilities or maintaining the types of
access making them subject to the FFD program. Observing behavior only
at the beginning of a work shift is not sufficient to ascertain whether
an individual is fit for duty, trustworthy, and reliable. Controlled
substances may have a delayed effect between use (e.g., ingestion) and
the onset of physiological or psychological effects, and fatigue
accumulates with time. Behavior must be continually observed throughout
the work shift to detect any changes from baseline human performance
characteristics, including mental or physical health and mannerisms, or
any activities that may indicate that the individual is not trustworthy
and reliable.
Proposed Sec. 26.609(a) would differ from Sec. Sec. 26.33 and
26.407 in that it would place the responsibility for performing
behavioral observation on ``all individuals subject to this subpart,''
rather than only those ``individuals specified in Sec. 26.4(f) [who]
are constructing or directing the construction of safety- or security-
related SSCs'' in Sec. 26.407 or ``individuals who are trained under
Sec. 26.29 to detect behaviors'' in Sec. 26.33 to improve clarity.
Proposed Sec. 26.609(b) would require all individuals subject to
the FFD program to report to the licensee- or other entity-designated
representative any onsite or offsite behaviors or activities by
individuals subject to this part that may constitute an unreasonable
risk to the safety or security of the NRC-licensed facility or SNM or
may cause harm to others. The NRC proposes this description of
reportable conduct because an individual's activities (e.g., use of
illegal substances) and communications (e.g., hate speech or threats of
violence) offsite are a direct indication of the individual's fitness,
trustworthiness, and reliability and must be evaluated as to whether
authorization should be granted or maintained. Proposed Sec. 26.609(b)
would include a description of this conduct instead of the Sec. 26.33
undefined phrase, ``FFD concerns,'' to enhance the clarity of the
requirement. This proposed BOP reporting requirement would include any
information relating to character or reputation of the individual
indicating that the individual cannot be trusted or relied upon to
perform those duties and responsibilities or maintain access to NRC-
licensed facilities, SNM, or sensitive information. This would better
align with the proposed Sec. 73.120 BOP requirement, which states that
each person subject to behavioral observation must communicate to the
licensee or applicant observed behaviors or activities of individuals
that may constitute an unreasonable risk to the health and safety of
the public and common defense and security. Proposed Sec. 26.609(a)
and (b) were written broadly to include offsite conduct that the
reporting individual considers serious enough to call into question the
character or reputation of the subject individual.
Proposed Sec. 26.609(c) would require that licensees and other
entities perform behavioral observation visually, in-person, and, when
necessary, remotely by live video and audible streaming and capture.
This requirement was developed from the security observation
requirements in Sec. 73.55(e)(7)(i)(B) and (C), (h)(2)(v), and (i)(2)
and (i)(5)(ii). Conducting an in-person observation of another
individual is the preferred method to ascertain whether the observed
individual can safely and competently perform assigned duties and
responsibilities. When in-person observations are not feasible (e.g.,
during solo operations), the proposed requirement would enable the use
of video monitoring. This is addressed, for example, in proposed Sec.
26.609(d) regarding NRC-licensed operator manipulation of reactor
controls. Additionally, certain duties (such as maintenance activities
performed by a single worker outside of a control room) may not present
an opportunity for video monitoring; in these situations, behavioral
observation should be conducted on a sampling basis (i.e., a planned
observation of the work activity) as outlined in a licensee's or other
entity's FFD program.
In situations involving small staff sizes, facilities sited in
geographically remote locations, or both, additional observers would
enhance the effectiveness of a BOP. Technological developments in
automated safety and security systems may enable licensees or other
entities to reduce staff sizes to 10 to 40 percent of the staff size of
an LWR facility licensed under part 50 or 52. Smaller staff sizes may
translate into more solo operations, less teamwork, fewer peer checks,
or infrequent management oversight of field activities, leading to
fewer behavioral observations. Therefore, a licensee or other entity
would have fewer opportunities to observe whether individuals are fit
for duty. Enabling video and audible streaming and capture to enhance
the BOP would be consistent with the security-related behavioral
observation requirement in proposed Sec. 73.120(c)(2)(ii), which would
also enable video conferencing or other acceptable electronic means
promoting face-to-face interaction for those individuals working
remotely.
Proposed Sec. 26.609(d) would require that licensees or other
entities perform behavioral observation of NRC-licensed operators who
manipulate the controls of any commercial nuclear plant licensed under
part 53, remotely by live video and audible streaming capture for those
part 53 facilities where individual task loading does not allow for the
effective conduct of behavior observation in addition to assigned
operational tasks. The purpose of this paragraph would be similar to
that of proposed Sec. 26.609(c), where the possibility of in-person
observation is significantly diminished because of solo operations or
because the facility may only require a minimum staff size onsite.
Proposed Sec. 26.610 would be equivalent to Sec. 26.409,
``Sanctions,'' and would require the licensee or other entity to
establish sanctions for FFD
[[Page 86974]]
policy violations that, at a minimum, prohibit the individuals
specified in Sec. 26.4 from being assigned to perform or direct those
duties and responsibilities or maintaining authorization making them
subject to subpart M of part 26. To be consistent with Sec. 26.75,
``Sanctions,'' the severity of the sanction as described in Sec.
26.610 would escalate with the number of occurrences and severity of
the FFD policy violation. The sanction would be long enough to help
deter future FFD policy violations and facilitate counseling and
treatment before the licensee reinstates the individual's access to the
facility. The NRC proposes this requirement because the 14-day denial
described in Sec. 26.75 may not allow sufficient time for counseling
and treatment based on the particular FFD policy violation.
Equivalent to Sec. 26.75(c), proposed Sec. 26.610 would also
require a minimum 5-year denial of access to the NRC-licensed facility
for certain violations of the FFD policy within the protected area of a
commercial nuclear plant and by an individual or individuals who are
the operators of the conveyance to transport or use formula quantities
of strategic SNM. Equivalent to Sec. 26.75(b), proposed Sec. 26.610
would require a permanent denial of authorization be issued for any
subversion attempt.
Proposed Sec. 26.611 would protect information collected from FFD
program implementation and would be equivalent to current Sec. 26.411,
``Protection of information.'' The protected information would include,
but not be limited to, privacy and medical information. Section 26.611
would not include the Sec. 26.411 requirement that FFD programs must
maintain and use the personal information with the highest regard for
individual privacy because such a requirement would be unnecessary in
light of the proposed Sec. 26.611(a) requirement that licensees and
other entities must establish and maintain a system of files and
procedures to prevent unauthorized disclosure.
Proposed Sec. 26.611(b), although equivalent to Sec. 26.411(b),
would require licensees and other entities to have all individuals sign
a consent to be subject to the FFD program before subjecting the
individual to the FFD program (e.g., before being subject to a pre-
access test in Sec. 26.607(b)(1), unlike Sec. 26.411(b)). The purpose
of this proposal would be to enhance protections afforded to
individuals subject to the FFD program and their knowledge of, in part,
why they are subject to drug and alcohol testing, behavioral
observation, information collection, MRO reviews, and other FFD program
elements. Like the consent required by Sec. 26.411(b), the consent
would authorize disclosure of the collected information. Consent would
not be needed for disclosures to the individuals and entities specified
in Sec. 26.37(b)(1) through (b)(6), (b)(8), and persons deciding
matters under review in proposed Sec. 26.613, ``Appeals process.''
Proposed Sec. 26.613 would be equivalent to Sec. 26.413, ``Review
process.'' The proposed title was changed to an appeal process to
clarify that Sec. 26.613 would be the process implemented when an
individual elects to appeal a licensee or other entity determination
that the individual had violated the FFD policy. The proposal would
also require that the process include a schedule for the completion of
the review of the determination that the individual had violated the
FFD policy. The NRC proposes this requirement because operating
experience demonstrates that workers may not be protected from a
continuous review process that does not result in an outcome.
Proposed Sec. 26.615 would require licensees and other entities to
perform audits of the FFD program. The proposed section would be
equivalent to Sec. 26.415, ``Audits.'' Under proposed Sec. 26.615(a),
audits would be performed at a frequency that ensures the FFD program's
continuing effectiveness. This would be particularly important for FFD
program elements that are not part of the FFD PMRP required by Sec.
26.603(d). Corrective actions would be taken as soon as reasonably
practicable to resolve any problems identified and preclude recurrence.
Proposed Sec. 26.615(b) would require the subject matter, scope, and
frequency of audits be revised as necessary to improve or maintain
program performance based on findings resulting from licensee or other
entity implementation of its FFD PMRP. These requirements were
developed from appendix B to part 50, ``Quality Assurance Criteria for
Nuclear Power Plants and Fuel Reprocessing Plants''; criterion X,
``Inspection''; and criterion XVIII, ``Audits.''
Proposed Sec. 26.615(c) would be equivalent to Sec. 26.415(b) and
would enable licensees and other entities to conduct joint audits or
accept audits of C/Vs so long as the audit addresses the relevant
services of the C/Vs.
Proposed Sec. 26.615(d) would be equivalent to Sec. 26.415(c) by
establishing requirements for the auditing of HHS-certified
laboratories. Unlike Sec. 26.415(c), the proposal would not contain a
reference to the Department of Transportation drug and alcohol testing
requirements. This would broaden the regulatory flexibility afforded to
a licensee or other entity in that they may use an offsite collection
or testing facility that does not meet the Department of Transportation
requirements.
Proposed Sec. 26.615(d) would state that licensees and other
entities need not audit an HHS-certified laboratory if the licensee's
or other entity's panel of drugs and drug metabolites to be tested is
equivalent to the panel by which the laboratory is certified by HHS or
is subject to the standards and procedures for drug testing and
evaluation used by the laboratory under the HHS Guidelines. The NRC
would afford this flexibility because the NRC is aware that HHS desires
to streamline changes in its guidelines to its panel of drugs and drug
metabolites to be tested. Therefore, if a licensee or other entity
elects to implement the HHS Guidelines in its procedures and maintains
the minimum panel of drugs and drug metabolites to be tested as
required by subpart M of part 26, a licensee or other entity may still
use (and not audit) the HHS-certified laboratory because the Sec.
26.603(e) change control process would maintain FFD program
effectiveness.
To help ensure FFD program effectiveness, Sec. 26.615(d) would
also require that collection facility procedures are comparable to
those required in subpart E of part 26, including a proposed
requirement that the offsite facility's specimen collection and testing
procedures are audited on a biennial basis, which is also a protection
consideration afforded to individuals subject to the FFD program.
Conducting this audit on a biennial basis would be equivalent to that
required in Sec. 26.41(b) and would help ensure that the specimen
collection process at the facility remains effective.
Proposed Sec. 26.617 would establish recordkeeping and reporting
requirements equivalent to those in current Sec. 26.417. However,
Sec. 26.617 would require retention of records pertaining to
administration of the FFD program and FFD performance data required by
Sec. 26.717 until license termination, which is based on current Sec.
26.711(a) because Sec. 26.417 does not provide for a retention period.
Proposed Sec. 26.617(b)(1) would be identical to the reporting
requirements in Sec. 26.417(b)(1) regarding the licensee's or other
entity's FFD program.
Proposed Sec. 26.617(b)(2) would require the reporting of annual
(i.e., January through December) program performance information to the
NRC before March 1 of the following year. This reporting would be
equivalent to
[[Page 86975]]
the annual program performance requirement in Sec. 26.417(b)(1), and
the March 1 due date is based on the reporting deadline in Sec.
26.717(e). Licensees and other entities would be required to report FFD
performance information using new NRC Forms 893, ``Single FFD Policy
Violation Form,'' and 894, ``10 CFR part 26, subpart M, Annual
Reporting Form for FFD Performance Information.''
Proposed Sec. 26.617(c) would require that FFD-related information
be shared within the commercial nuclear industry when requested to
support authorization determinations. This requirement would help
individuals seeking employment by another NRC-licensed facility subject
to subpart C of part 26, complete their NRC-required sanctions and
licensee-administered or -directed drug and/or alcohol abuse treatment
plans before the restoration of authorization by a licensee or other
entity. Information sharing may also enhance FFD program effectiveness
because FFD-related lessons learned from, for example, substance
testing, subversion attempts, and laboratory and MRO performance must
be shared when requested.
Proposed Sec. 26.619 would require licensees or other entities to
establish a process to evaluate individuals when their fitness or
trustworthiness and reliability are in question. Section 26.619 would
be equivalent to Sec. 26.419, ``Suitability and fitness
determinations,'' but, unlike Sec. 26.419, would apply during the
construction and operation phases. Also, proposed Sec. 26.619 would
require that a suitability or fitness determination conducted for cause
be conducted face-to-face. This proposed requirement is based on
current Sec. 26.189(c); however, unlike Sec. 26.189(c), proposed
Sec. 26.619 would not prohibit augmenting determinations via
electronic means of communication. Instead, Sec. 26.619 would
explicitly permit determinations to be performed via electronic means,
so long as those determinations are supported by an appropriately
trained individual who is present in-person with the individual being
assessed.
In considering the current restriction on the use of electronic
means of communication for determinations of fitness conducted for
cause, the NRC finds that since publication of the 2008 part 26 final
rule, there have been developments in using electronic means of
communication (i.e., ``videoconferencing'') as an alternative to
conducting face-to-face interactions. To address these considerations,
the NRC contracted the Pacific Northwest National Laboratory (PNNL),
DOE, to study whether a medical and mental health assessment via
electronic communication could be an acceptable alternative to an in-
person, face-to-face assessment.\13\ Based on this study, if electronic
means were to be used to conduct a face-to-face assessment, an in-
person element would still be integral to the assessment process.
However, under certain circumstances, face-to-face determinations and
assessments conducted as part of an FFD program for an entity licensed
under part 53 (i.e., those determinations and assessments performed in
accordance with Sec. 26.619, Sec. 26.207, or Sec. 26.211) may be
augmented via electronic communications. Such remotely conducted
determinations and assessments would be required to be conducted with
someone who is present in-person with the individual being assessed and
who is trained in accordance with the requirements of either Sec.
26.29 and Sec. 26.203(c) or Sec. 26.608 and Sec. 26.202(c).
Permitting the use of electronic communications would help ensure FFD
program effectiveness, especially in instances where the part 53
commercial nuclear plant is sited in a geographically remote location
or when the facility has a small staff size.
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\13\ PNNL, Technical Letter Report, ``The Use of Electronic
Communications to Perform Determinations of Fitness,'' dated August
2017.
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D. Proposed Changes to Part 26, Subpart N
Proposed Sec. 26.709 would make the recordkeeping and reporting
requirements in subpart N of part 26 applicable to licensees and other
entities of facilities licensed under part 53 that elect not to
implement the requirements in subpart M of part 26 or elect to
implement the requirements in Sec. 26.605(b).
Proposed Sec. 26.711(c) and (d) would be amended to make these
requirements applicable to licensees or other entities described in
Sec. 26.3(f). Section 26.711(c) provides protection to individuals
subject to part 26 by enabling an individual's right to review FFD-
related information and correct any inaccurate or incomplete
information. Section 26.711(d) requires, in part, that any FFD-related
information shared with other licensees or other entities is correct
and complete.
E. Proposed Changes to Part 26, Subpart O
The vast majority of the proposed changes to part 26 would be new
or revised substantive provisions that would establish a regulatory
obligation or prohibition or would be conforming edits to reflect the
addition of part 53. The only new provision that would not be
substantive, such that violation of it would not result in a criminal
penalty, would be proposed Sec. 26.601. Therefore, the NRC proposes to
add Sec. 26.601 to the list of regulations in Sec. 26.825(b) to which
criminal sanctions do not apply.
10 CFR Part 50
A. Section 50.160: Emergency Preparedness for Small Modular Reactors,
Non-Light-Water Reactors, and Non-Power Production or Utilization
Facilities
This proposed rule would revise Sec. 50.160(b)(3) and (c)(2) to
make that section applicable to applicants and licensees under part 53.
Section 50.160 provides an alternative to other part 50 emergency
preparedness requirements focused on large light-water reactors to
provide an optional emergency preparedness framework specifically for
small modular reactors (SMRs) and other new technologies. These
alternative emergency preparedness requirements adopt a performance-
based, technology-inclusive, risk-informed, and consequence-oriented
approach. Commercial nuclear reactor applicants complying with Sec.
50.160 would be required to submit as part of the application the
analysis used to determine whether the criteria in Sec.
53.1109(g)(2)(i)(A) and (B) are met and, if they are met, the size of
the plume exposure pathway emergency planning zone (EPZ). An EPZ bounds
the area surrounding a facility within which detailed planning is
needed to implement predetermined, prompt protective actions. The
criterion in proposed Sec. 53.1109(g)(2)(i)(A) is that public dose, as
defined in Sec. 20.1003, is projected to exceed 10 mSv (1 rem) TEDE
over 96 hours from the release of radioactive materials from the
facility considering accident likelihood and source term, timing of the
accident sequence, and meteorology. The criterion in proposed Sec.
53.1109(g)(2)(i)(B) is that pre-determined, prompt protective measures
are necessary. These are the same criteria that are in Sec.
50.33(g)(2)(i)(A) and (B) and are used to assess the need for and size
of an EPZ in applications under parts 50 and 52.
Applicants choosing to comply with Sec. 50.160 must determine the
radiological releases from the facility that are evaluated in the
determination of the plume exposure pathway EPZ. Consistent with other
Federal guidelines such as the Federal Emergency Management Agency
``Radiological Emergency Preparedness Program Manual,'' issued in 2023,
and the
[[Page 86976]]
Environmental Protection Agency ``PAG Manual: Protective Action Guides
and Planning Guidance for Radiological Incidents,'' issued in 2017,
applicants should consider quantitative and qualitative information on
the potential radiological releases that make up the spectrum of
accidents used to develop the basis for the applicant's site-specific
EPZ. This information is derived from the licensing basis. The NRC
plans to update the risk-informed approach in RG 1.242 for part 53
while maintaining its flexibility for using information already
developed and available in licensing basis documents, including PRA
results, deterministic dose quantities, accident timing, target set
analyses, mitigation capabilities, and site-specific factors such as
meteorology.
In its safety analysis report, the applicant would describe the
LBEs relevant to the facility and would consider these LBEs as
candidates for the spectrum of accidents used to develop the site-
specific EPZ. The LBEs assessed include a wide range of events that are
appropriate for considering in the facility's emergency preparedness
and response planning. In addition, Sec. 50.160(b)(1)(iv)(A)(2)
requires licensees to be capable of implementing their approved
emergency response plan in conjunction with their safeguards
contingency plan. Radiological sabotage events are typically factored
into EPZ determinations by considering consequences to be bounded by
LBEs and by crediting protection against the DBT in reducing the
likelihood of a release.
The provisions in proposed Sec. 53.860(a) provide an alternative
to applicants and licensees by not requiring them to protect against
the DBT of radiological sabotage in accordance with Sec. Sec. 73.55
and 73.100 if they can demonstrate that the consequences from
unmitigated radiological sabotage events are below the safety criteria
in proposed Sec. 53.210. The deployment of some commercial nuclear
plants under part 53 may involve new scenarios where the source terms
and consequences of sabotage-related events are not bounded by the
consequences of the unlikely and very unlikely event sequences analyzed
under subpart C. Accordingly, the NRC plans to develop guidance for
part 53 applicants and licensees choosing to comply with the
alternative emergency preparedness requirements in Sec. 50.160 to
address this new class of reactors. In Section VI of this document, the
NRC is asking for stakeholder feedback on the clarity of the
regulations and guidance for various scenarios that might arise in
implementing graded approaches for security and emergency planning for
some commercial nuclear plant designs.
B. Appendix B to Part 50: Quality Assurance Criteria for Nuclear Power
Plants and Fuel Reprocessing Plants
Appendix B to part 50 would be amended to make it applicable to
applicants and licensees under part 53. This results in the need for
some revisions to recognize differences in terminology between parts 50
and 53. Namely, the term ``design bases,'' which is defined in Sec.
50.2, is not used in part 53. For this reason, text is added in both
Section III, ``Design Control,'' and Section IV, ``Procurement Document
Control,'' to refer to ``functional design criteria, as defined in
Sec. 53.020,'' as the part 53 equivalent of the term ``design bases.''
10 CFR Part 73
A. Section 73.100: Technology-Inclusive Requirements for Physical
Protection of Licensed Activities at Commercial Nuclear Plants Against
Radiological Sabotage
Proposed Sec. 73.100 would provide a performance-based regulatory
framework for the design, implementation, and maintenance of a physical
protection program and security organization for certain commercial
nuclear plants licensed under part 53. The current Sec. 73.55 physical
security requirements for nuclear power reactors licensed under part 50
and part 52 use a combination of performance criteria (e.g., Sec.
73.55(b)(1) through (3)) and numerous prescriptive requirements
developed to achieve performance objectives (e.g., Sec.
73.55(k)(5)(ii)). By contrast, in the proposed performance-based
approach to physical security for part 53, performance objectives and
requirements would be the primary bases for regulatory decision-making,
giving the licensee the flexibility to determine how to demonstrate
compliance with the established performance criteria for an effective
physical protection program. This proposed physical protection program
would provide an optional pathway for licensees that elect not to
demonstrate compliance with the provisions in Sec. 73.55 and do not
satisfy the criterion as described in proposed Sec. 53.860(a)(2). This
proposed physical protection program would provide that activities
involving SNM are not inimical to the common defense and security and
do not constitute an unreasonable risk to the public health and safety.
Section 73.100(a) would require each part 53 licensee that elects
to demonstrate compliance with this section rather than Sec. 73.55 to
implement the requirements therein through its physical security plan,
training and qualification plan, safeguards contingency plan, and
cybersecurity plan (referred to collectively hereafter as ``security
plans'') prior to initial fuel load into the reactor (or, for a fueled
manufactured reactor, before initiating the physical removal of any one
of the independent physical mechanisms to prevent criticality required
under Sec. 53.620(d)(1)). The security plans would need to identify,
describe, and account for site-specific conditions that affect the
licensee's capability to satisfy the requirements of Sec. 73.100.
Based on experience from recent new reactor licensing reviews, the NRC
recognizes that licensees may seek to receive unirradiated fuel onsite
before carrying out the security requirements in Sec. 73.100. However,
these security requirements would have to be implemented at some point
before reactor operation to address the increased risk arising from
irradiated fuel onsite. This proposed rule would make clear that part
53 applicants and licensees using Sec. 73.100 may bring unirradiated
nuclear fuel onsite and protect it in accordance with the NRC's
requirements for physical protection of SNM of moderate and low
strategic significance under Sec. 73.67 until initial fuel load into
the reactor (or, for a fueled manufactured reactor, until initiating
the physical removal of any one of the independent physical mechanisms
to prevent criticality required under Sec. 53.620(d)(1)).
Section 73.100(b) would outline the general performance objective
and design requirements of the licensee physical protection program. A
licensee's program would be required to provide protection against any
deliberate act within the DBT of radiological sabotage, including spent
fuel sabotage, which could directly or indirectly endanger the public
health and safety by exposure to radiation. The physical protection
program is supported by the AA program, cybersecurity program, and IMP
to demonstrate compliance with the general performance objective of
Sec. 73.100(b).
Section 73.100(b)(2) was developed, in part, from Sec.
73.55(b)(3). To satisfy the general performance objective of Sec.
73.100(b)(1), the physical protection program would need to protect
against the DBT of radiological sabotage. The existing fleet of LWR
satisfies this objective by preventing significant core
[[Page 86977]]
damage and spent fuel sabotage. Some non-LWR reactor licensees'
physical protection programs may be designed to prevent a significant
release of radionuclides from any source. Therefore, the proposed
performance objective would focus on radiological sabotage in general,
rather than a specific focus on core damage or spent fuel sabotage, to
be technology inclusive and allow for flexibility for different reactor
technologies.
Under the proposed Sec. 73.100(b)(2)(ii), licensees must provide
defense in depth in achieving performance requirements through the
integration of engineered systems, administrative controls, and
management measures. This requirement would apply defense-in-depth
concepts as part of the physical protection program to ensure the
capability to demonstrate compliance with the performance objective of
the proposed Sec. 73.100(b)(1) is maintained in the changing threat
environment. The defense-in-depth philosophy applies to measures
against intentional acts as required by Sec. 73.100(b), and the
designs of physical security systems should employ defense in depth
through systems diversity, independence, and separation under Sec.
73.100(b)(2). The most common defense-in-depth measures apply concepts
of redundancy, diversity, independence, and safety margin to ensure
systems reliability and availability. The defense-in-depth philosophy
applies to the design of a physical protection program, which
integrates engineered controls and administrative controls, to provide
protection against the DBT for radiological sabotage.
Section 73.100(b)(3) would require the physical protection program
to be designed and implemented to achieve and maintain the reliability
and availability of SSCs required for demonstrating compliance with
specified performance requirements. These physical protection
performance requirements were informed by Sec. 73.55(b) and the
Commission's Advanced Reactor Policy Statement.
The performance objective of protecting against the DBT of
radiological sabotage is achieved by the design and implementation of
the physical protection program, maintained at all times, with the
following required performance capabilities proposed in the provisions
in Sec. 73.100(b)(3): intrusion detection, intrusion assessment,
security communication, security response, protecting against land and
waterborne vehicle bomb assaults, and access control portals. The
physical protection program must maintain the reliability and
availability of SSCs relied upon for demonstrating compliance with the
performance requirements. The terms ``reliability and availability''
are intended to describe defense in depth in a performance-based manner
and would be critical elements for demonstrating compliance with the
proposed requirement for protection against the DBT of radiological
sabotage as described in the proposed Sec. 73.100(b)(2).
The first element, ``intrusion detection,'' would be provided
through the use of detection equipment, patrols, access controls, and
other program elements and would provide notification to the licensee
that a potential threat is present and where the threat is located.
The second element, ``intrusion assessment,'' would provide a
mechanism through which the licensee would identify the nature of the
threat detected. This would be accomplished through the use of video
equipment, patrols, and other program elements that would provide the
licensee with timely information about the threat for use in
determining how to respond.
The third element, ``security communication,'' would provide a
mechanism through which the licensee would communicate the necessary
information to the response force to ensure effectiveness of the
physical protection program. This would be accomplished through the
redundant, independent, and diverse design of physical security and/or
plant SSCs relied on for onsite and offsite security communications.
The continuity and integrity of communications should account for the
DBT's ability to affect the reliability and availability of
communications.
The fourth element, ``security response,'' would provide a
mechanism through which the licensee would be capable of timely
security response to interdict and neutralize threats up to and
including the DBT of radiological sabotage. The security response may
include the use of onsite armed responders, law enforcement responders
(local, State, or Federal), or other offsite armed responders (e.g.,
licensee proprietary or contract security personnel who are positioned
offsite), or a combination thereof, as appropriate.\14\ The licensee
must provide protection against any element of the DBT, to include
those that do not rise to the full capability of the DBT. Structures,
systems, and components relied on to provide delay functions must be
designed to provide for timely response to adversary attacks with
adequate defense in depth. Delay would allow the licensee to take
necessary actions to counter any attempt by the threat to advance
towards the protected target or target set element. The overall
response objective would be to place the threat in a condition from
which the threat no longer has the potential for, or capability of,
doing harm to the protected target.
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\14\ The NRC's security regulations for commercial nuclear power
reactors have historically considered onsite armed responders to be
the only acceptable method for interdicting and neutralizing threats
up to and including the DBT of radiological sabotage. The proposed
rule would permit advanced power reactor licensees to use any
interdiction and neutralization method, which would be an extension
of the Commission's position in SRM-SECY-17-0100, ``Security
Baseline Inspection Program Assessment Results and Recommendations
for Program Efficiencies,'' dated October 8, 2018, and SRM-SECY-20-
0070, ``Technical Evaluation of the Security Bounding Time Concept
for Operating Nuclear Power Plants,'' dated June 6, 2024. Under the
proposed rule, a licensee would retain the responsibility to detect,
assess, interdict, and neutralize threats up to and including the
DBT of radiological sabotage, but would be able to rely on law
enforcement or other offsite armed responders as a method for
fulfilling the required interdiction and neutralization
capabilities. For licensees that choose to rely on law enforcement
to fulfill these capabilities, the proposed rule would not create
any NRC regulatory jurisdiction over, or requirements for, law
enforcement. In SRM-SECY-23-0021, ``Proposed Rule: Risk-Informed,
Technology-Inclusive Regulatory Framework for Advanced Reactors (RIN
3150-AK31),'' dated March 4, 2024, the Commission approved a similar
approach to defend against radiological sabotage.
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The fifth element, ``protecting against land and waterborne vehicle
bomb assaults,'' would provide a mechanism through which the licensee
would be capable of protecting the plant against the DBT vehicle bomb
assault. The methods that are relied on to protect against a DBT land
vehicle and waterborne vehicle bomb assault must be designed to protect
the reactor building, structures containing safety or security related
systems, and components from explosive effects.
The sixth element, ``access control portals,'' would provide a
mechanism through which the licensee would be capable of detecting and
denying unauthorized access to persons and pass-through of contraband
materials (e.g., weapons, incendiaries, explosives) to protected areas.
Integrity of the access control system is maintained through licensee
oversight and ensures that attempts to circumvent or bypass the
established process will be detected and access denied.
The proposed performance requirements would permit the applicant or
licensee to determine how to design the physical protection program to
protect the plant against the DBT of radiological sabotage without
[[Page 86978]]
prescriptive requirements such as those currently found in Sec. 73.55.
DG-5076, ``Guidance for Technology Inclusive Requirements for Physical
Protection of Licensed Activities at Commercial Nuclear Plants,'' has
been developed by the NRC to describe one acceptable approach to
demonstrate compliance with requirements proposed in Sec. 73.100.
Section 73.100(b)(4) would require the licensee to identify target
sets in accordance with Sec. 73.55(f). For non-LWR and SMRs, target
sets would be defined in DG-5071, ``Target Set Identification and
Development for Nuclear Power Plants,'' as the minimum combination of
equipment, operator actions, and/or structures that, if all are
prevented from performing their intended safety function or prevented
from being accomplished, barring extraordinary actions by plant
operations, would likely result in a significant release of
radionuclides from any source (e.g., a release to the environment
exceeding that analyzed in the DBA licensing basis).
Section 73.100(b)(5) would require that each licensee perform a
site-specific analysis for the purpose of identifying and analyzing
site-specific conditions that affect the design of the onsite physical
protection program.
Section 73.100(b)(6) would require licensees to implement a
performance evaluation program, which would ensure that a licensee will
periodically test and evaluate the effectiveness of the physical
protection program to protect against the DBT. This program would
ensure that licensees are able to demonstrate that the physical
protection program satisfies the response requirements of Sec. 73.100
and that the site's protective strategy effectively protects against
the DBT. Licensee performance evaluations would include methods to
assess, test, and challenge the integration of the physical protection
programs functions and demonstrate the effectiveness of security plans,
licensee protective strategy, and implementing procedures in accordance
with Sec. 73.100(g).
Section 73.100(b)(7) would require licensees to implement an AA
program in accordance with Sec. 73.56. Section 73.100(b)(8) would
require licensees to establish, maintain, and implement protection
against a cyberattack based on either the proposed cybersecurity
program described in Sec. 73.110 or the program described in existing
Sec. 73.54.
Section 73.100(b)(9) would require an IMP that monitors the initial
and continuing trustworthiness and reliability of individuals granted
or retaining unescorted access or unescorted AA to a protected or vital
area. The IMP must also implement defense-in-depth methodologies to
minimize the potential for an insider (active, passive, or both) to
adversely affect the licensee's capability to protect against
radiological sabotage. Because no one element of the AA program, FFD
program, cybersecurity program, or physical protection program, would,
by itself, provide the level of protection against the insider
necessary to demonstrate compliance with the performance objective of
the proposed Sec. 73.100(b), the effective integration of these
programs is a necessary requirement to achieve defense in depth against
the potential insider.
Section 73.100(b)(10) would require that the licensee have the
capability to track, trend, correct, and prevent recurrence of failures
and deficiencies in the implementation of the requirements of this
section. Section 73.100(b)(11) would require the coordination of the
security plans and associated procedures with other onsite plans to
manage the safety and security interface during normal or emergency
operations.
Section 73.100(c) was developed from Sec. 73.55(c)(7), ``Security
implementing procedures,'' and Sec. 73.55(d), ``Security
organization,'' and would outline the requirements for the composition,
equipping, and training of the security organization. The purpose of
the security organization is to effectively implement the physical
protection program. Individuals assigned to perform physical protection
or contingency response duties must be trained, equipped, and qualified
to perform assigned duties and responsibilities.
Section 73.100(d) would establish a performance requirement for
searches of personnel, vehicles, and materials for the protection
against radiological sabotage. The requirement describes broad
categories of material (explosives, firearms, incendiary devices, etc.)
to be detected and prevented from entry into the protected area;
specific items that will be prohibited would not be prescribed in the
regulation but will be stated in the licensee security plans with
detailed descriptions being identified in implementation procedures.
Section 73.100(e) would require a training and qualification
program, described in the training and qualification plan, that ensures
personnel are able to effectively perform their assigned security-
related job duties. This high-level requirement would allow flexibility
in how the licensee chooses to train its security personnel. One method
for accomplishing this requirement would be to provide a training and
qualification program that would be equivalent to appendix B to part
73.
Section 73.100(f) would require periodic security reviews of the
physical protection program to ensure effective implementation of the
program by independent individuals. The evaluation process would
provide a systematized approach for assessing the physical protection
program as a basis for further development and improvement. Program
reviews should be designed to ensure that the physical protection
program maintains effectiveness and demonstrates compliance with NRC
requirements. Section 73.100(f)(1) was developed from Sec. 73.55(m)
and would require review of each element of the physical protection
program. Section 73.100(f)(2) would require licensees to perform self-
assessments of physical protection program functions to ensure that the
capability to detect, assess, interdict, and neutralize the DBT of
radiological sabotage is maintained. Section 73.100(f)(3) would require
an audit of the effectiveness of the physical protection program;
security plans; implementing procedures; cybersecurity programs;
management of the safety/security interface activities; the testing,
maintenance, and calibration program; and response commitments by
local, State, and Federal law enforcement authorities. Section
73.100(f)(4) would require that results and recommendations, management
findings, and any actions taken be documented and maintained to be
available for inspection by the NRC. These reviews are independent of
the ongoing performance evaluations described in Sec. 73.100(b)(6) and
(g).
Section 73.100(g) would require that licensee performance
evaluations, described in Sec. 73.100(b)(6), include methods
appropriate and necessary to assess, test, and challenge the
integration of the physical protection program's functions to protect
against the DBT. The performance evaluations must also address the
licensee's measures to protect against cyberattacks, in accordance with
the required cybersecurity plan, and engineered systems designed to
protect against the DBT standalone ground vehicle bomb attack.
Section 73.100(h) would establish performance requirements for
maintaining security SSCs relied on to perform security functions to
protect against the DBT. It would require that corrective actions and
compensatory measures be taken by a licensee in response to a
degradation of security
[[Page 86979]]
equipment or failure of the equipment to perform its intended
functions. The licensee would be required to maintain the SSCs
described in its design and licensing basis to ensure that they are
reliable and available.
Section 73.100(i) would establish requirements for the suspension
of security measures in response to emergency and extraordinary
conditions. The requirements of this paragraph, which were developed
from Sec. 73.55(p), would be intended to provide flexibility to a
licensee for taking reasonable actions that depart from a security plan
in an emergency when such actions are immediately needed to protect the
public health and safety and no action consistent with license
conditions and TS that can provide adequate or equivalent protection is
immediately apparent in accordance with proposed Sec. 53.740(h).
Section 73.100(j) would establish requirements regarding the
inspection, retention and maintenance of records required to be kept by
the NRC regulations, orders, or license conditions. These proposed
requirements are developed from Sec. 73.55(q).
B. Section 73.110: Technology-Inclusive Requirements for Protection of
Digital Computer and Communication Systems and Networks
Section 53.860 would require that a licensee establish, implement,
and maintain a cybersecurity program in accordance with Sec. 73.54 or
Sec. 73.110. Section 73.110 would establish requirements for the
development and maintenance of a cybersecurity program for commercial
nuclear plants licensed under part 53. This proposed section would
implement a graded approach to determine the level of cybersecurity
protection required for digital computers, communication systems, and
networks. The proposed new section is informed by: (1) the operating
experience from power reactors and fuel cycle facilities; and (2) the
existing Sec. 73.54 framework, which addresses some of the basic
issues for cybersecurity regardless of the type of reactor. Differences
between the Sec. 73.54 requirements and those proposed in Sec. 73.110
are primarily based on the implementation of a consequence-based
approach to cybersecurity that provides flexibility to accommodate the
wide range of reactor technologies to be assessed by the NRC. A graded
approach based on consequences is intended to account for the differing
risk levels among reactor technologies. Specifically, the proposed new
section would require licensees to demonstrate protection against
cyberattacks in a manner that is commensurate with the potential
consequences from those attacks.
Under proposed Sec. 73.110(a), licensees would need to ensure that
digital computer and communications systems are adequately protected
against a potential cyberattack that would result in: (1) a scenario
where the cyberattack leads to offsite radiation doses that would
endanger public health and safety (i.e., the resulting consequence
exceeds the reference dose values in Sec. 53.210); or (2) a scenario
where the cyberattack adversely impacts the physical security digital
assets used by the licensee to prevent unauthorized removal of material
or radiological sabotage. Security digital assets would include those
used for nuclear MC&A.
The proposed Sec. 73.110(b) would require licensees to protect the
communication system and networks associated with the functions
described in Sec. 73.110(a)(1) and (a)(2) from cyberattacks. To
accomplish this, the licensee would establish, implement, and maintain
a cybersecurity program for protecting digital assets within the scope
of Sec. 73.110 that would make use of risk insights, including threat
information, and would consider the resulting level of consequences of
the threats. If the outcome of the assessment by the licensee under
Sec. 73.110(b)(1) revealed that a potential cyberattack would not
compromise any digital assets that support safety and security
functions, and thus would not result in the consequences listed in
Sec. 73.110(a) (e.g., would not exceed the reference dose values),
then only a narrow set of the cybersecurity program requirements in
Sec. 73.110(d) and (e) would apply. For example, the licensee would
only need to develop a cybersecurity program that implements the
requirements dealing with:
Analyzing modifications of any asset before implementation
to see if they demonstrate compliance with the potential consequences
in Sec. 73.110(a);
Ensuring employees and contractors are aware of
cybersecurity requirements and have some level of cybersecurity
training;
Evaluating and managing cybersecurity risks to the plant;
Reviewing the cybersecurity plan for any required changes;
and,
Retaining records of the cybersecurity plan along with any
plan changes.
Section 73.110(c) through (e) were developed from Sec.
73.54(a)(2), and (c) through (h), respectively.
The proposed requirements would address the need for the licensee
to develop a cybersecurity program that implements a defense-in-depth
protective strategy as required by proposed section Sec. 73.110(d)(2).
A defense-in-depth protective strategy for cybersecurity is represented
by collections of complementary and redundant security controls that
establish multiple layers of protection to safeguard critical digital
assets. Under a defense-in-depth protective strategy, the failure of a
single protective strategy or security control should not result in the
compromise of safety and security functions.
C. Section 73.120: Access Authorization Program for Commercial Nuclear
Plants
Section 73.120 would address AA for certain commercial nuclear
plants licensed under part 53. The proposed language in Sec. 73.120
would provide an alternate approach to the existing framework for AA
under Sec. Sec. 73.55, 73.56, and 73.57, commensurate with risk and
consequences to public health and safety. It would be available to part
53 applicants and licensees who demonstrate in an analysis that the
offsite consequences of a DBE satisfy the criterion defined in Sec.
53.860(a)(2)(i) (i.e., would not exceed the offsite dose values in
Sec. 53.210(b)). The proposed requirements in Sec. 73.120 would be
similar to the existing AA program elements for those NRC licensed
facilities issued additional security measures (ASMs) orders and for
materials licensees under Sec. 37.21. Applicants not satisfying the
criterion would need to establish, implement, and maintain a full AA
program, including an IMP, in accordance with Sec. 73.56.
Proposed Sec. 73.120(a) would be based on an applicant satisfying
the eligibility criterion in Sec. 53.860(a)(2)(i). Section 73.120(b)
would identify the categories of individuals who would be subject to an
AA program in accordance with this section. The applicability statement
in Sec. 73.120(b)(1)(i) would encompass individuals whom the licensee
intends to grant unescorted access to the facilities' most sensitive
areas, consistent with Sec. 73.56(b)(1)(i) for power reactors and the
ASM orders and license conditions issued to any NRC licensed facility
or material licensee. Sections 73.120(b)(1)(ii) through (iv) would be
consistent with Sec. 73.56(b)(1)(ii) through (iv), respectively. The
program would include individuals who may be onsite or offsite (e.g.,
remote operators or information technology staff) and have virtual
access to important plant operational and communication systems based
upon assigned duties and
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responsibilities. An individual who has remote access to plant
equipment and communication systems may have trusted privileges greater
than the personnel at the plant site. Section 73.120(b)(1)(iii) would
state that offsite law enforcement personnel on official duty would not
be subject to the licensee AA program.
Section 73.120(c) would provide general performance objectives and
requirements largely consistent with the AA program requirements for
nuclear power reactors under Sec. 73.56 and would provide licensees
and applicants the flexibility in establishing their AA program to
demonstrate compliance with various performance objectives.
Section 73.120(c)(1) would include background investigation
requirements consistent with Sec. 37.25, as well as ASMs and license
conditions that are applied to non-power reactor licensees. Background
investigations include important elements to establish the
trustworthiness and reliability of an individual, such that they do not
constitute an unreasonable risk to public health and safety or the
common defense and security. These include the following: (1) personal
history disclosure, (2) verification of true identity, (3) employment
history evaluation, (4) unemployment/military service/education, (5)
credit history evaluation, (6) character and reputation evaluation, and
(7) Federal Bureau of Investigation criminal history record check.
Section Sec. 73.120(c)(2) would establish behavioral observation
requirements, which are an awareness initiative for recognizing
behaviors adverse to the safe operation and security of the facility
through observing the behavior of others in the workplace and reporting
aberrant behavior or changes in behavior that might reflect negatively
on an individual's trustworthiness or reliability. Maintaining
behavioral observation would assist and/or improve worker safety and
reduce the risk of an insider threat. This proposed requirement in
Sec. 73.120(c)(2) would be a scaled version of the full BOP required
under Sec. 73.56(f).
Section Sec. 73.120(c)(2) would provide licensees greater
flexibility to implement behavioral observation options for individuals
granted unescorted access to the commercial nuclear plant's protected
area. Such options on reporting questionable behavior may include a
program similar to the Department of Homeland Security's program, ``If
you see something, say something,'' or to a corporate behavioral
awareness program. Commensurate with the potential lower safety and
security risks of a commercial nuclear plant that meets the criterion
in Sec. 53.860(a)(2)(i), Sec. 73.120(c)(2) would not require the
establishment of a comprehensive training program for behavioral
observation (i.e., initial and refresher training including knowledge
checks) as required for power reactors under Sec. 73.56 and part 26.
Under Sec. 73.120(c)(2)(ii), behavioral observation would be able to
be performed in-person or remotely by video, and identified behavior of
concern would need to be reported to plant supervision. The remote
access alternative to face-to-face interactions provides substantial
flexibility for licensees and applicants. Any video conferencing or
other acceptable electronic means promoting face-to-face interaction
for those individuals working remotely would demonstrate compliance
with this regulation.
Section 73.120(c)(3) captures and maintains the self-reporting of
legal actions as an essential performance element to enhance the
licensee's behavioral observation initiative similar to the current
requirements under Sec. 73.56(g), assuring that personnel who are
granted and who maintain unescorted access are trustworthy and
reliable.
Section 73.120(c)(4) would provide a scalable approach for granting
and maintaining unescorted access. One component not included from
Sec. 73.56 is the need for a psychological assessment and reassessment
under Sec. 73.56(e) for granting unescorted access and Sec.
73.56(i)(v)(B) for individuals who perform one or more of the job
functions described in Sec. 73.120(b)(1)(ii) for maintaining
unescorted access. Moreover, the requirement would permit criminal
history updates to be completed within 10 years of the last review,
compared to the three- or five-year reinvestigation periodicity for
personnel at an operating commercial nuclear plant. In addition, no
credit check re-evaluation would be required for these individuals.
The continued need to maintain unescorted access would be evaluated
on an annual basis by the reviewing official. Guidance in DG-5074,
``Access Authorization Program for Commercial Nuclear Plants,'' would
specify that this evaluation should be based on a compilation of
personnel interactions as described in the licensee's or applicant's
policy and procedures for behavioral observation and the maintenance of
an approved AA list.
Section 73.120(c)(5) would require licensees and applicants to
determine when a person no longer requires the need for unescorted
access or no longer satisfies the AA requirement found within this
section. Guidance in DG-5074 would further explain that licensees have
the flexibility to terminate unescorted access to specific areas of the
site if individuals lack the continued need for that access to perform
their duties and responsibilities.
Section 73.120(c)(6) would be consistent with the purpose of Sec.
37.23(e) and would include the individual's right to correct and
complete information as required under Sec. 37.23(g). The section
would include a requirement for designating a reviewing official. The
language would provide clarity regarding the roles and responsibility
of a reviewing official, who would be the only individual authorized to
make unescorted access determinations.
Section 73.120(c)(7) would align with the corresponding
requirements under Sec. 37.23(f), and Sec. 73.120(c)(8) would align
with the corresponding requirements under Sec. 37.31. These
requirements would encompass the roles and responsibilities for
licensees, applicants, and if applicable, the contractor/vendors to
establish, implement, and maintain a system of files and records to
ensure personal information is not disclosed to unauthorized persons.
Section 73.120(c)(9) would align with the requirements of Sec.
37.33. Section 73.120(c)(10) would require licensees, applicants, and
contractors or vendors to maintain the records that are required by the
regulations in this section and retain them for a period of 3 years
after the record is superseded or no longer needed. The record
retention period of three years would be consistent with Sec.
37.23(h), contrasting with the five-year retention period under Sec.
73.56(o). Records maintained in any database(s) would need to be
available for NRC review, consistent with the requirements found under
Sec. 73.56(o)(6)(ii).
VI. Specific Requests for Comments
The NRC is seeking advice and recommendations from the public on
this proposed rule. We are particularly interested in comments and
supporting rationale from the public on the following:
Part 26--Fitness for Duty Program
1. The proposed rule under Sec. 26.603(c) would enable a licensee
or other entity to implement an FFD program under proposed Sec.
26.604, ``FFD program requirements for facilities that satisfy the
Sec. 26.603(c) criterion,'' if the
[[Page 86981]]
licensee or other entity performs a site-specific analysis to
demonstrate that the facility and its operation satisfy the criterion
in Sec. 53.860(a)(2).
Should the NRC consider replacing its proposed Sec. 26.603(c)
criterion referencing Sec. 53.860(a)(2) with an alternative
requirement that if the commercial nuclear plant is of the class
described in Sec. 53.800, ``Facility licensees for self-reliant-
mitigation facilities,'' and either Sec. 53.800(a)(1) or (2) is
satisfied, then drug and alcohol testing would not be required? This
proposal would align the Sec. 26.603(c) criterion with that proposed
in the NRC-licensed operator regulatory framework of part 53. Please
provide your considerations and rationale for your recommendation.
Should the NRC also consider making a conforming change to the
proposed Sec. 73.120 criterion used for the AA program? Please provide
your considerations and rationale for your recommendation.
Part 26--Technology-Inclusive Approaches to Fatigue Management
Requirements Applicable to Unit Outages
In establishing the outage minimum days off requirement of Sec.
26.205(d)(4), the NRC's objective was to ensure that individuals
performing the duties described in Sec. 26.4(a)(1) through (a)(4) have
sufficient periodic long-duration breaks to prevent cumulative fatigue
from degrading their ability to safely and competently perform their
duties. In addition to the science of fatigue management, the NRC
considered several factors in establishing the existing requirements.
These additional factors were practical and safety considerations
associated with the management of refueling outages for large LWRs,
including the following: (1) the typical duration and frequency of
outages; (2) the availability of contract personnel to perform the
work; (3) the risk presented by the outage work while the reactor is
shut down; and (4) the controls applied to the work that may limit the
potential for latent errors to challenge reactor safety when the
reactor is returned to power. The details of such considerations may
differ for new reactor technologies or designs. Such considerations may
not be relevant for some reactor designs (e.g., reactors capable of on-
line refueling) and there may be additional, more pertinent factors to
consider for other designs.
The NRC is seeking stakeholder input on whether alternative fatigue
management requirements applicable to outages should be adopted to
support technology-inclusive approaches that would be appropriate to
support the licensing and regulation of future commercial nuclear
plants. Please provide your considerations and rationale for your
recommendation.
Part 26--Draft Regulatory Guidance Approach for Fatigue Management
In support of this proposed rule, the NRC has issued DG-5078,
``Fatigue Management for Nuclear Power Plant Personnel at Commercial
Nuclear Plants Licensed Under 10 CFR part 53.'' This DG describes
methods the NRC staff considers acceptable for addressing certain
aspects of FFD programs at commercial nuclear facilities licensed under
part53.
The NRC staff also intends to eventually transition this draft
guide into an update to RG 5.73, ``Fatigue Management for Nuclear Power
Plant Personnel,'' or the development of a new RG. At this point, NRC
staff is considering four options for future RG development:
Option 1: Amend the existing RG. The NRC may develop an
updated version of RG 5.73 that continues to endorse (with
clarifications, additions, and exceptions) the guidance contained in
NEI 06-11, ``Managing Personnel Fatigue at Nuclear Power Reactor
Sites,'' Revision 1, and incorporates the topics discussed within DG-
5078 as new NRC staff positions in section C of RG 5.73.
Option 2: Issue a new RG specific to part 53 licensees.
The NRC may develop an entirely new RG applicable specifically to
facilities licensed under part 53. This new RG would capture the
guidance contained in DG-5078 and incorporate existing guidance (e.g.,
selected guidance in RG 5.73 and NEI 06-11) that is considered to be
technology inclusive in nature. The existing guidance (i.e., RG 5.73)
would remain in place as the guidance for facilities licensed under
parts 50 and 52.
Option 3: Review and potentially endorse new or revised
industry-developed guidance. The NRC may engage with the industry
regarding a potential update to industry guidance document NEI 06-11 or
the development of new, separate industry-developed guidance specific
to facilities licensed under part 53. The NRC would then review the new
or revised industry-developed guidance within the NRC's RG process,
which includes opportunities for public participation. New or revised
industry-developed guidance could incorporate DG-5078 or propose
alternatives for the NRC to consider.
Option 4: Develop a comprehensive revision of the existing
RG. The NRC may develop a more comprehensive revision of RG 5.73 that
would explicitly detail all NRC positions reflected in the existing RG
(including those endorsed positions currently contained in NEI 06-11,
Revision 1), along with the guidance of DG-5078. Such a revision would
thereby be a ``stand-alone'' document, without reference to or explicit
endorsement of separate, industry-developed guidance.
The NRC is seeking stakeholder input regarding which of the four
options listed above would be optimal (or whether there are other
options that the NRC should consider). Please provide your
considerations and rationale for your recommendation.
Part 53--Overall Organization
Part 53 is structured as one framework with subparts providing
technical, licensing, and administrative requirements for the various
stages of the life cycle of a commercial nuclear plant. The
organization of part 53 in this manner puts a complete set of
requirements for each stage of the life cycle in a separate subpart
with additional subparts for licensing and administrative requirements.
The NRC is seeking comment on the proposed organization of the
requirements in part 53 and possible improvements to how specific
requirements (e.g., examples of which specific sections) could be
consolidated or otherwise reorganized to make the rule clearer or more
concise.
There are numerous references in proposed part 53 to other NRC
regulations. Examples of such references include those in proposed
Sec. 53.610 to NRC regulations related to radiation protection (part
20), FFD (part 26), physical security (part 73), and MC&A (10 CFR part
74, ``Material Control and Accounting of Special Nuclear Material'')
for facilities receiving byproduct or SNMs.
The NRC is seeking comment on whether such references to other
regulations in various sections in the proposed part 53 provide
benefits to applicants and licensees, or to other stakeholders seeking
to understand the regulatory framework under part 53, or whether such
references could be removed to reduce the length of part 53.
Part 53, Subpart B--Comprehensive Risk Metrics
The NRC is proposing to require the use of comprehensive risk
metrics and associated risk performance objectives as one of several
performance standards in part 53. Comprehensive risk metrics could
include a risk metric or set of risk metrics that approximate the total
overall risk from the facility to the
[[Page 86982]]
extent practicable. Associated risk performance objectives are
preestablished values indicative of the comprehensive risk metrics that
are used during risk-informed decision-making to gauge plant safety.
Specifically, comprehensive risk metrics and associated risk
performance objectives would provide one element of the safety criteria
for LBEs other than DBAs in the proposed Sec. 53.220. Comprehensive
risk metrics, in the form of the IEFR and the ILCFR, and associated
risk performance objectives, in the form of the QHOs of
5x10-7 per year and 2x10-6 per year,
respectively, were similarly used in the LMP methodology to ensure that
other evaluation criteria were conservatively defined and as a tool for
focusing attention on matters important to managing the risks posed by
nuclear power plants. The use of such comprehensive risk metrics and
associated risk performance objectives in an integrated risk-informed
decision-making process is similar to that used in RG 1.174, ``An
Approach for Using Probabilistic Risk Assessment in Risk-Informed
Decisions on Plant-Specific Changes to the Licensing Basis,'' Revision
3.
The NRC is seeking comment on the use of comprehensive risk metrics
and associated risk performance objectives in part 53 as one of several
performance standards. The IEFR and ILCFR and the QHOs represent
comprehensive risk metrics and associated risk performance objectives
that the NRC has used for decades in a variety of capacities. What
other performance standards could be used to address the comprehensive
risks posed by proposed commercial nuclear plants? Please provide your
considerations and rationale for your recommendation.
If an applicant proposes a novel approach to comprehensive plant
risk and the NRC approves the approach, should the resulting NRC-
approved comprehensive plant risk metrics and associated risk
performance objectives be codified or otherwise memorialized over time
and, if so, how?
Part 53, Subpart B--Defense in Depth
Proposed Sec. 53.250 would establish requirements based on the
longstanding NRC philosophy of providing defense in depth to address
uncertainties concerning the design, operation, and performance of
commercial nuclear plants during LBEs.
The NRC is seeking comment on the inclusion of the proposed
requirements to assess and provide defense in depth. The NRC is also
seeking comment on whether to include specific provisions in Sec.
53.250 and subpart B to more explicitly address the possible role of
inherent characteristics of some SSCs in preventing or mitigating
unplanned events. The proposed Sec. 53.250 is worded to preclude
relying on a single engineered design feature to address the range of
LBEs other than DBAs, which could possibly allow crediting inherent
characteristics without further lines of defense. How could possible
inherent characteristics of SSCs be considered in the proposed
requirements in Sec. 53.250 or in any alternative requirements for
defense in depth provided in response to this item? Please provide your
considerations and rationale for your recommendation.
Part 53, Subpart C--Probabilistic Risk Assessment
Current consensus PRA standards provide processes for appropriately
defining the scope of a PRA and determining applicability of supporting
requirements to suit the specific needs of a given applicant under
proposed part 53. In addition to assessing other aspects of PRA
acceptability such as PRA peer reviews, NRC determinations of the
acceptability of such PRAs would assess the appropriateness of the
applicant-defined scope as part of determining the applicability of a
consensus PRA standard supporting an application. This approach is
consistent with the current state of practice and offers appropriate
flexibility for PRAs to be developed and assessed based on the
application they are used to support, which includes consideration of
how PRA results and insights are relied upon, together with factors
such as safety margin, simplicity of design, and treatment of
uncertainty.
The NRC is seeking comment on what additional guidance, if any, is
needed regarding PRA acceptability for Part 53 applicants and
licensees.
Part 53, Subparts C and D--Earthquake Engineering
Proposed Sec. 53.480 would establish requirements related to
seismic design considerations. This proposed section is intended to
provide a clear connection between siting activities and seismic design
activities and to support various approaches to presenting seismic
hazards and addressing those hazards in designs. The proposed
requirements are intended to provide sufficient flexibility to allow
approaches like those currently in parts 50 and 100 or approaches that
might be endorsed by the NRC in the future that could incorporate more
risk insights from PRAs.
The NRC is seeking comment on whether the proposed requirements for
earthquake engineering provide appropriate flexibility in addressing
seismic risks while also ensuring that the regulations continue to
adequately address seismic hazards. Please provide your considerations
and rationale for your recommendation.
Part 53, Subpart E--Construction and Manufacturing
1. Proposed Sec. 53.610(b)(1)(iii) would require procedures that
describe how construction will be controlled so as not to impact other
features important to the design (e.g., dewatering, slope stability,
backfill, compaction, and seepage).
The NRC is seeking comment on whether such specific requirements
are useful or whether these requirements could be met through other
requirements proposed in part 53 or already present in other relevant
regulations (e.g., quality assurance requirements in appendix B to part
50).
Part 53, Subparts E and H--Manufacturing Licenses
1. The proposed requirements governing manufacturing are set forth
in subpart E, and the proposed requirements governing the licensing
processes are contained in subpart H. Some of the proposed
requirements, including provisions related to the loading of
unirradiated fuel into a manufactured reactor, are intended to cover a
factory-fabrication model that has been suggested for some micro-
reactor designs. However, as written, the proposed provisions are not
limited to any size or type of reactor.
The NRC is seeking comment on whether the proposed regulations are
sufficient to govern various scenarios for the possible manufacturing
and deployment of manufactured reactors.
If a comment indicates that the proposed regulations are not
sufficient, please describe the reasons why, including, if applicable,
any plausible scenario for which the commenter believes the proposed
regulations are not sufficient.
2. The proposed regulations in subpart H allow holders of or
applicants for a COL to reference an ML but do not include such a
provision for the holder of or applicant for a CP or OL. This proposed
change from the current relationship between subparts in part 52 and
the part 50 licensing process was made to simplify the provisions in
the proposed part 53 for licensing and deploying manufactured reactors.
The NRC seeks comment on whether part 53 should include provisions
for an applicant for or a holder of a CP or an OL to reference an ML
and, if so, how this should be done.
[[Page 86983]]
3. Proposed Sec. 53.1295 states that the holder of an ML could not
begin manufacture of a manufactured reactor less than 6 months before
the expiration of the license. This limitation is similar to the
current restriction in Sec. 52.177, which states that the manufacture
of a reactor cannot begin less than 3 years before the expiration of
the license. The restriction was revised from 3 years in part 52 to 6
months in the proposed part 53 in recognition of the likely use of MLs
for a factory-fabrication model for micro-reactors.
The NRC seeks comment on whether it is necessary or appropriate to
revise the 3-year restriction in part 52 on when manufacturing
activities could begin in relation to license expiration and, if so,
what that restriction should be.
4. Proposed Sec. 53.1288 provides the finality provisions for MLs
and includes, as does existing Sec. 52.171, limitations on the NRC's
imposition of new requirements on either the design or the requirements
for the manufacture of a manufactured reactor. No MLs have been issued
under part 52 and there is no practical experience with the proposed
finality sections. While the implications of the finality provisions
related to the design of a manufactured reactor can reasonably be
inferred from experience with DCs and COLs, there is no experience or
available guidance regarding finality for ``requirements for the
manufacture of the manufactured reactor.''
The NRC is seeking comment on the proposed finality provisions for
MLs and specifically if and how finality for manufacturing processes
might be requested and used.
5. The NRC is seeking comment on the proposed regulations for the
loading of fresh (unirradiated) fuel into a manufactured reactor for
subsequent transport to a site for which the Commission has issued a
COL that authorizes construction and operation of a commercial nuclear
plant using the manufactured reactor. The proposed regulation includes
provisions for loading of fuel into manufactured reactors at a
manufacturing facility prior to transporting the fueled reactor to its
deployment site, as suggested by some stakeholders. The NRC has
historically viewed reactor operation as including fuel load, and
existing NRC regulations reflect this view. While the Act authorizes
the NRC to issue licenses to manufacture production or utilization
facilities, it does not contain specific provisions on fueling or
operating facilities licensed under an ML, and existing ML regulations
under part 52 do not include provisions for fuel load.
The proposed rule addresses this matter by allowing an applicant to
combine an ML with a part 70 license, which would authorize possession
of a manufactured reactor in which the licensee has loaded unirradiated
fuel provided at least two independent criticality prevention
mechanisms are in place, each of which is sufficient to prevent
criticality assuming optimum neutron moderation and neutron reflection
conditions. This requirement would limit the possibility of creating
fission products and allow the control of SNM, so that the loading of
the fuel into a manufactured reactor could be governed primarily via a
part 70 license and associated regulations (including those in subpart
H of part 70).
A specific topic on which the NRC is seeking comment is on the
potential benefits of and issues with including the requirements of
subpart H of part 70 within the proposed regulations for loading fuel
into manufactured reactors at the manufacturing facility. For example,
should the NRC include a threshold for including the requirements of
subpart H of part 70 and, if so, what factors and decision criteria
should be considered in such a threshold? If a comment indicates that
the proposed regulations are not sufficient, please describe the
reasons why, including the plausible scenarios for which the proposed
regulations would not work or could be made to work better.
6. Section 170, ``Indemnification and Limitation of Liability,'' of
the Act states that each license under section 103 shall have as a
condition of the license a requirement that the licensee have and
maintain financial protection of such type and in such amounts as the
NRC shall require.
The NRC is seeking comment on whether the proposed regulations
should include amounts of required financial protections for MLs for
fueled manufactured reactors, and, if so, what would be appropriate
amounts of required financial protection.
7. Some stakeholders have suggested that a fueled manufactured
reactor with appropriate protections against criticality should not be
categorized as a utilization facility under NRC regulations or Section
11cc. of the Act.
The NRC is seeking comment on possible approaches where the NRC
could find that a fueled manufactured reactor would not be a
utilization facility, the basis for such a finding, and the potential
benefits of and potential issues with such a finding.
8. Proposed requirement Sec. 53.620(d)(2)(i) would require a
security program, including a physical security plan, for any ML
authorizing possession of a manufactured reactor into which fuel has
been loaded at the manufacturing facility. Currently, requirements in
Sec. 73.67(c)(1) only require that a physical security plan be
submitted for those licensees who possess, use, transport, or deliver
to a carrier for transport SNM of moderate strategic significance, or
10 kg or more of SNM of low strategic significance.
The NRC is seeking comment on whether the proposed requirement: (1)
should be specific to the facility type (i.e., manufacturing facility)
or be specific to the category of material being used at the facility;
(2) should apply to all manufacturing plants, including those at which
licensees may only possess SNM of low strategic significance (i.e.,
category III), or only those facilities for which an applicant must
submit a physical security plan per Sec. 73.67(c)(1); or (3) should
include more specific requirements on the supplemental security
measures that may be needed for licensees possessing SNM of moderate
strategic significance (i.e., category II)?
9. Proposed requirement Sec. 53.620(d)(2)(i) would require a
cybersecurity program. The proposed general cybersecurity performance
requirements would be to provide reasonable assurance that a
cyberattack could not adversely impact the functions performed by
digital assets used by the licensee for implementing the physical
security, radiation monitoring, and criticality requirements.
The NRC is seeking comment on the following: (1) to what extent
stakeholders envision physical security controls, radiation monitoring,
and criticality controls at a manufacturing facility being digital; (2)
to what extent should the ML holder be required to protect digital
computer and communications systems that impact safety and security
functions from a cyberattack at a manufacturing facility authorized to
load fuel; and (3) whether the rule provides sufficient clarity on the
cybersecurity measures needed for license issuance or if additional
detail should be included either in the rule or in guidance?
10. Proposed requirement Sec. 53.620(d)(2)(i)(B) would require
that the physical security program be designed to prevent unintended
and uncontrolled criticality events. This would include criticality
events that are initiated maliciously.
The NRC is seeking comment on whether the ML holder should be
required to design its security program to protect against radiological
sabotage
[[Page 86984]]
(i.e., an unintended criticality event leading to unacceptable
radiological consequences), in addition to theft and diversion. For
example, should the NRC establish security requirements to prevent an
adversary, including an insider, from tampering with the reactor at a
manufacturing facility or during transport in such a way as to cause an
inadvertent criticality event? If so, should the NRC consider factors
such as the category of fuel and the number of reactors at a factory
that can simultaneously be loaded with fuel in establishing the
security requirements?
11. Proposed requirement Sec. 53.620(d)(2)(i) would require an ML
holder to meet the performance objectives in Sec. 73.67. Requirements
Sec. 73.67(e) and Sec. 73.67(g) include provisions for security of
category II and category III quantities of SNM, respectively, during
transportation.
The NRC is seeking comment on the extent to which the ML should
require ASMs (i.e., security measures above those required by Sec.
73.67(e) and Sec. 73.67(g)) for transportation of a fueled reactor to
its place of operation. What should those measures be?
12. Proposed requirement Sec. 53.620(d)(2)(i) would require an ML
holder to meet the performance objectives of Sec. 73.67. For licensees
utilizing a category II quantity of SNM, the requirement in Sec.
73.67(d)(4) would have the ML holder conduct a screening to confirm the
identity of an individual prior to granting unescorted access to the
controlled access area where the material is used or stored. The
purpose of this requirement is to both confirm the identity of the
individual and support a determination that the individual is
trustworthy and reliable.
The NRC is seeking comment on whether the ML requirements should
include ASMs (i.e., measures beyond those required by Sec.
73.67(d)(4)) in order to provide reasonable assurance of identity
confirmation and trustworthiness and reliability.
13. The NRC is seeking comment on whether provisions regulating the
testing of fueled manufactured reactors in the manufacturing facility
should be included in part 53 and, if so, what would be practical for
the holder of an ML while also providing adequate protection of public
health and safety. One possibility could be COLs that would be issued
to the holders of an ML to cover low power (e.g., <5% rated thermal
power) nuclear physics testing of fueled manufactured reactors within
the manufacturing facility prior to the manufactured reactors being
transported to and incorporated into a commercial nuclear plant for the
purpose of energy production. The NRC recognizes configuration changes
are needed to perform nuclear physics testing and is seeking comment on
what requirements should apply to the manufactured reactors and the
manufacturing facility during such testing (e.g., limiting power
levels). If a comment indicates that the regulations should address
limited operations at manufacturing facilities, please describe the
likely scenarios that would need to be addressed and suggest what would
be appropriate requirements for such scenarios.
While an ML holder could accomplish nuclear physics testing by
applying for a COL under the proposed subpart H of part 53,
stakeholders have indicated that many of the requirements would likely
be unnecessary, given the reduced risk profile posed by such
activities. Therefore, the NRC is seeking comment on what requirements
in subpart H of part 53 should apply to applicants for a COL who would
perform testing of fueled manufactured reactors at the manufacturing
plant. Examples of proposed requirements that might be relaxed or
modified for applications for low power testing at manufacturing plants
include those related to selection of LBEs to reflect limited inventory
of radionuclides and decay heat, aircraft impact assessments, and
earthquake engineering.
Additionally, the NRC is seeking comment on whether several other
requirements in part 53 could be modified for applications for a low
power testing COL at a manufacturing facility. For example, the NRC is
seeking comment on how portions of the ML facility used to support
testing should fall within the requirements for construction activities
under Sec. 53.610; whether Sec. Sec. 53.710 and 53.715 (SSC
configuration control) must be implemented to ensure portions of the ML
facility relied on to limit potential radiological consequences from
LBEs are available to perform their safety functions; and whether the
requirements of Sec. 53.730 could be modified to reflect the
conditions of low power physics testing. If a comment indicates that
some design and analysis requirements and related application
requirements in subpart H of the proposed part 53 are not needed for
the testing of fueled manufactured reactors, please provide a rationale
supporting your comment and, if applicable, what alternate requirements
would be appropriate.
Moreover, the licensing mechanism for the facility could present
unique challenges. One option could be to issue a low power testing COL
for each fueled manufactured reactor to be tested. This would comport
with the agency's practice of issuing one license per reactor but could
prove prohibitive from a cost standpoint and may provide very little
safety benefit if all manufactured reactors are the same.
Alternatively, one low power testing COL could be issued for the
portions of the ML facility used to test the fueled manufactured
reactors and allow multiple fueled manufactured reactors to be
completed and tested over the course of the ML. Under this approach,
any ITAAC related to testing of the fueled manufactured reactors would
need to be closed after they were manufactured but prior to testing,
and the NRC would issue a notice of intended operation and provide the
public an opportunity to request a hearing on whether each fueled
manufactured reactor as constructed complies, or on completion will
comply, with the acceptance criteria of the license. The NRC is seeking
comment on the potential benefits and issues with having a COL for each
fueled manufactured reactor to be tested versus having a COL cover the
testing of multiple fueled manufactured reactors. If a comment
indicates a preference for a particular approach, please provide a
rationale supporting the comment and describe the specific scenarios
that the regulations need to address.
Part 53, Subpart F--Staffing and Generally Licensed Reactor Operators
Under the Act Sections 106 and 107, the NRC is proposing to group
commercial reactors into classes upon the basis of the similarity of
operating and technical characteristics of the facilities, and then to
prescribe uniform conditions for licensing individuals as operators of
any of the various classes; determine the qualifications of such
individuals; and, for certain classes of commercial reactors, issue
general licenses (i.e., licenses for which no application is needed) to
such individuals allowing the individuals to operate the commercial
reactor.
1. Categories of Individuals Who May Manipulate Facility Controls:
The NRC is proposing requirements that would allow the manipulation of
the controls of certain facilities by GLROs in lieu of specifically
licensed reactor operators and senior reactor operators. Reactor
operators and senior reactor operators are the only categories of
individuals currently allowed to be licensed to manipulate the controls
of utilization facilities under part 55.
The NRC is interested in public perspectives on this proposed
addition of the GLRO category, particularly in light of new reactor
technologies and concepts of operations.
[[Page 86985]]
2. Criteria for GLRO Staffing: The NRC is proposing criteria under
which facilities would be staffed by GLROs in lieu of specifically
licensed reactor operators and senior reactor operators. These criteria
establish a new class of self-reliant-mitigation facilities, as defined
in part 53, for which distinct GLRO licensing and staffing requirements
would apply.
The NRC is soliciting public feedback regarding whether these
proposed criteria are appropriate and what, if any, alternative
criteria should be considered. Please provide your considerations and
rationale for your answer.
3. Medical Requirements for GLROs: Based on the proposed criteria
that a self-reliant-mitigation facility, as defined in part 53, must
meet, the NRC is proposing not to subject GLROs to requirements for
medical fitness and medical examination. This is in contrast with the
proposed requirements associated with specifically licensed reactor
operators and senior reactor operators, as well as the existing
requirements for reactor operators and senior reactor operators under
part 55.
The NRC is soliciting public feedback regarding whether GLROs
should be subject to medical fitness and/or medical examination
requirements like reactor operators and senior reactor operators.
Please provide your considerations and rationale for your answer.
4. Onshift Engineering Expertise: The NRC is proposing to require
that engineering expertise be accounted for within facility staffing
plans. This proposed requirement would be in lieu of the traditional
position of the Shift Technical Advisor. The NRC is further proposing
that individuals providing such engineering expertise would need, among
other things, to possess either a qualifying 4-year degree or licensure
as a Professional Engineer.
The NRC is interested in feedback from the public regarding the
appropriateness of this requirement, including any alternatives that
should be considered. Please provide your considerations and rationale
for your answer.
5. Use of Simulation Facilities as HFE Testbeds: The NRC is
proposing to establish regulations pertaining to the use of simulation
facilities within the context of the licensing programs both for
specifically licensed reactor operators and senior reactor operators as
well as for GLROs. However, these regulations, as currently proposed,
do not address the use of simulation facilities within the context of
serving as testbeds for HFE-related analyses and assessments. Rather,
the NRC currently envisions that the use of simulation facilities as
HFE testbeds is more appropriately addressed via guidance documents.
The NRC is soliciting public feedback regarding whether simulation
facility requirements should also address the use of simulation
facilities as HFE testbeds. Please provide your considerations and
rationale for your answer.
Part 53, Subpart F--Emergency Preparedness and Security Programs
1. The proposed framework for part 53 would incorporate the changes
to NRC regulations from the final rulemaking on ``Emergency
Preparedness for Small Modular Reactors and Other New Technologies''
(the EP for SMR/ONT rule) by including references to Sec. 50.160,
``Emergency preparedness for small modular reactors, non-light-water
reactors, and non-power production or utilization facilities,'' and by
making conforming changes within Sec. 50.160. The proposed framework
for part 53 would also introduce a graded approach to physical
protection requirements that includes the criterion in Sec.
53.860(a)(2)(i) to establish a class of licensees that would not be
required to protect against the design-basis threat (DBT) of
radiological sabotage. The NRC is soliciting public comment relating to
these topics, which could include ways that graded approaches for both
emergency preparedness and security programs might be assessed and
considered during the licensing process.
The NRC is seeking comment on the sufficiency and clarity of
requirements in proposed part 53 related to the assessments needed to
support graded emergency planning and security. If a comment indicates
that there is an issue with the sufficiency or clarity of the proposed
regulations, please describe the reasons why, including, if applicable,
any scenario for which the proposed regulations are not sufficient and
possible ways to clarify the requirements. The NRC is specifically
seeking comment on possible challenges arising from the interactions
between the proposed regulations and related assessments for grading
the requirements for emergency planning and security.
2. The NRC is preparing various guidance documents to support this
rulemaking and other ongoing or recently completed rulemakings related
to emergency preparedness and security. DG-5076, ``Guidance for
Technology-Inclusive Requirements for Physical Protection of Licensed
Activities at Commercial Nuclear Plants,'' has been issued along with
this proposed rulemaking and public comments are requested via this
notice on that draft guidance. The NRC is also planning to issue a
draft revision of RG 1.242, ``Performance-Based Emergency Preparedness
for Small Modular Reactors, Non-Light-Water Reactors, and Non-Power
Production or Utilization Facilities,'' for public comment. The planned
revision to RG 1.242 would add guidance for part 53 applicants and
licensees.
In the staff requirements memorandum to SECY-23-0021, the
Commission directed the NRC staff to address the consideration of
security-related events for an advanced reactor that addresses security
through design and engineered safety features when it harmonizes this
rulemaking with the EP for SMR/ONT rule. In the EP for SMR/ONT rule,
the NRC established an alternative performance-based and risk-informed
approach for emergency planning, including determining the need for and
size of an emergency planning zone (EPZ) to support predetermined,
prompt protective actions. The NRC has incorporated the relevant rule
language from the EP for SMR/ONT rule into this proposed rule and is
seeking stakeholder feedback as to whether additional rule language
changes or additional guidance would be beneficial.
In light of the Commission direction and the above considerations,
the NRC is assessing how best to address the treatment of security-
related events in emergency planning, including in the determination of
EPZ size, for reactors licensed under part 53. Part 53 is introducing
an alternative approach to meeting security regulations that should be
taken into consideration under Sec. 50.160. Stakeholders are
encouraged to take a holistic view of the various activities and
opportunities to provide comments on this rulemaking and related
guidance supporting this rulemaking (e.g., DG-5076 on physical
protection requirements, future revisions to RG 1.242). In developing
comments, the NRC urges stakeholders to consider various scenarios that
might arise when implementing graded approaches for security and
emergency planning for various reactor designs. Scenarios could include
the following:
the potential consequences from security events up to and
including the DBT of radiological sabotage are bounded by unlikely and
very unlikely event sequences such that security events do not need
separate analyses in the EPZ size determination;
[[Page 86986]]
the potential consequences from security events up to and
including the DBT are not bounded by unlikely and very unlikely event
sequences but could otherwise support a reduced EPZ size consistent
with considerations discussed in RG 1.242 and NUREG-0396, ``Planning
Basis for the Development of State and Local Government Radiological
Emergency Response Plans in Support of Light Water Nuclear Power
Plants''; or
the potential consequences from security events up to and
including the DBT are not bounded by unlikely and very unlikely event
sequences and warrant consideration of increasing the size of the EPZ.
The NRC is interested in comments on the need for additional rule
language or guidance to address graded approaches for emergency
planning and security programs under the scenarios described above for
part 53 applicants and licensees. Please address within the comments
any technical, policy, or legal issues that are associated with your
suggestions.
Part 53, Subpart F--Integrity Assessment Program Requirements
Decades of operating experience with LWRs suggests that phenomena
such as environmentally assisted fatigue and chemical interactions
could impact certain SSCs during the life of a commercial nuclear
plant. Under the existing regulatory framework, historically, some of
these phenomena were not addressed during early licensing reviews but
were identified and addressed later when significant safety issues
arose (e.g., see numerous generic letters, bulletins, orders, and
development and implementation of vessel integrity and materials
reliability programs) or a licensee voluntarily pursued renewal of an
OL under part 54. The NRC is proposing to include a new set of
programmatic requirements for an Integrity Assessment Program that
would ensure these phenomena are addressed early in the life of a
commercial nuclear plant licensed under part 53. The requirements would
be provided in Sec. 53.870.
The NRC is seeking comment on whether the proposed requirements
under the Integrity Assessment Program appropriately complement design
requirements to address concerns regarding aging, cyclic or transient
load limits, and degradation mechanisms related to chemical
interactions, operating temperatures, effects of irradiation, and other
environmental factors. In addition, the NRC is interested in views on
whether, and if so how, degradation mechanisms are or could be
addressed in other programs.
Part 53, Subpart G--Decommissioning
1. On March 3, 2022, the NRC published the proposed rule entitled
``Regulatory Improvements for Production and Utilization Facilities
Transitioning to Decommissioning'' (87 FR 12254). This rulemaking would
amend the NRC's current regulations to provide an appropriate
regulatory framework for nuclear power reactors transitioning from
operations to decommissioning. The rulemaking would address lessons
learned from licensees that have completed or are currently in the
decommissioning process. The NRC staff sent a draft final rule to the
Commission for its consideration on January 31, 2024, in SECY-24-0011,
``Final Rule: Regulatory Improvements for Production and Utilization
Facilities Transitioning to Decommissioning (3150-AJ59; NRC-2015-
0070).''
What aspects of this draft final rule, if any, should be
incorporated in a part 53 final rule and why?
2. Proposed Sec. 53.1060(b) in subpart G would require that, ``No
later than 30 days after the Commission publishes notice in the Federal
Register under Sec. 53.1452(a), the licensee must submit a report
containing a certification that financial assurance for decommissioning
is being provided in an amount specified in the licensee's most recent
updated certification, including a copy of the financial instrument
obtained to satisfy Sec. 53.1040.'' This is similar to the current
requirement in Sec. 50.75(e)(3) for part 52 COL holders. The NRC is
seeking comment on whether commercial nuclear plant COL holders under
part 53 should have the same requirement as COL holders under part 52
to demonstrate that they have financial assurance in place no later
than 30 days after the Commission issues the notice of intended
operation under Sec. 53.1452. Please provide your considerations and
rationale for your answer.
Part 53, Subpart H--Licenses To Construct and Operate Commercial
Nuclear Plants of Identical Design at Multiple Sites
In addition to including provisions in part 53, subpart H, for
referencing ESPs, standard design approvals, and design certifications
in applications for commercial nuclear plants, the proposed Sec.
53.1470 provides optional requirements related to the submittal and NRC
review of CP, OL, and COL applications to construct and operate
commercial nuclear plants of identical design at multiple sites,
similar to requirements found in appendix N in both 10 CFR parts 50 and
52. This section would set out the particular requirements and
provisions applicable to situations in which applications for CPs and
subsequent OLs, or COLs, under this part, are filed by one or more
applicants for licenses to construct and operate nuclear power reactors
of identical design (``common design'') to be located at multiple
sites. Hearings for applications filed under appendix N in both parts
50 and 52 are governed by subpart D of part 2, as would be the case for
future part 53 applications under proposed Sec. 53.1470.
Under the proposed requirements in this section, each application
is to be treated as a separate application, with the exception of the
common design, and so would require separate applications, separate
determinations of sufficiency for docketing, separate notices of
docketing, and so forth. Proposed Sec. 53.1470 would also require that
each application list all the applications that are to be treated
together to ensure that the NRC is clearly informed of the intentions
of all applicants. Ordinarily, the NRC would publish in the Federal
Register a separate notification of docketing for each application, so
that delays in the docketing of one application would not delay the
docketing and subsequent technical review of other applications.
However, if circumstances allow (e.g., sufficiency review for multiple
applications are completed simultaneously), the NRC could publish a
single notice of docketing for multiple applications.
With regard to how the NRC would fulfill its obligations under the
National Environmental Policy Act of 1969, as amended, the NRC staff
would prepare a separate environmental document for each application,
but the NRC could conduct joint scoping on environmental issues related
to the common design. If the applications reference a standard design
certification or the use of a manufactured reactor, then the
environmental document would need to incorporate by reference the
environmental assessment (EA) prepared for either the design
certification or the ML, as applicable. In addition, Sec. 53.1470
would require the ACRS to report on each of the applications, as would
be required by provisions in subpart H of part 53. Each ACRS report
would be limited to the safety matters which are not relevant to the
common design. In addition, the ACRS would need to issue a report on
the safety of the common design--except for those matters relevant to
the
[[Page 86987]]
safety of a referenced design certification or manufactured reactor.
Given this synopsis of how the requirements in proposed Sec.
53.1470 would be implemented as currently written, the NRC is seeking
comment on whether there are opportunities to allow added flexibility
for applicants under these provisions. This could include consideration
of whether applications for which the ``common design'' is not
completely identical could be evaluated under this provision and, if
so, what the process would be for determining the appropriateness of a
common review. In addition, the NRC is interested in feedback about the
pros and cons of requiring that applications under these proposed
provisions be submitted at the same time versus allowing them to be
submitted on a staggered basis.
Part 53, Subparts H and I--Probabilistic Risk Assessment Information
Proposed Sec. 53.1239(a)(18) in subpart H and the related
references to this proposed requirement for the holders of OLs and COLs
would require a description of the PRA required by Sec. 53.450(a), and
its results to be included in FSARs. However, guidance documents may
further clarify the division of PRA-related information needed to be in
the FSAR, in other possible licensing basis documents, and controlled
as plant records subject to inspections and audits. For example, a
possible approach could be to include a summary of the PRA results in
the FSAR and control that information under Sec. 53.1545 and create a
separate document related to the broader PRA analyses and related
processes as a program document under Sec. 53.1560. The program
document would provide more detail than the summaries in the FSAR but
still be a much-condensed source of information in comparison to the
documentation of the PRA. This possible approach would reflect the role
of the PRA in the licensing process under part 53 and in maintaining
margins to the safety and evaluation criteria in subparts B and C but
may allow a more appropriate evaluation process to address the
particulars and complexities of the PRA-related documents.
The NRC is seeking comment on the appropriate placement of PRA-
related information among various licensing basis documents and plant
records. In addition to the placement of PRA-related information, the
NRC is seeking comment on the appropriate control of that information
and on the routine submittal of updates to the NRC. Please provide your
considerations and rationale for your answer.
Part 53, Subparts H and I--Changes to Manufacturing Licenses
Proposed Sec. 53.1530 would not allow the holder of an ML or the
holder of a COL using a manufactured reactor to make changes to the
design of the manufactured reactor without requesting a license
amendment from the NRC. The proposed requirements do not include a
specific mention of the manufacturing processes for which the NRC could
possibly provide finality under proposed Sec. 53.1288.
The NRC is seeking comment on the appropriate change control
provisions for MLs, including whether criteria could be developed to
determine when a license amendment request would not be required and
whether those criteria should address changes in manufacturing
processes as well as changes in the design. Please provide your
considerations and rationale for your recommendation.
Financial Qualifications
Utility new reactor applicants are exempt under Sec. 50.33(f) from
financial qualification reviews because they are generically presumed
to be financially qualified for operations. In contrast, merchant power
plant new reactor applicants are required under Sec. 50.33(f)(2) to
submit information that demonstrates they possess or have reasonable
assurance of obtaining the funds necessary to cover estimated
construction and operating costs for the period of the license. A
``merchant power plant new reactor applicant'' is a non-rate-regulated
entity (e.g., a nonutility) that engages in the business of production,
manufacturing, generating, buying, aggregating, marketing, or brokering
electricity for sale at wholesale or for retail sale to the public.
Over the past decade, the agency has heard some concerns about the
challenges that merchant power plant applicants face in demonstrating
compliance with the current financial qualification requirements.
Does this standard continue to pose challenges for merchant power
plant applicants? If so, please provide a detailed explanation of these
challenges.
Should part 53 have the same financial qualification requirements
as parts 50 and 52? Why or why not?
Are there categories of merchant new reactor applicants for which a
part 70 ``appears to be financially qualified'' standard would be more
appropriate? \15\ If so, please explain what types of applicants should
be able to use the part 70 financial qualification standard and what
distinguishes these applicants from ones that should not be able to use
this standard.
---------------------------------------------------------------------------
\15\ Section 70.23(a)(5).
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If a part 70 financial qualification standard were to apply to a
category of merchant new reactor applicants, should it also apply to
pre-construction license transfer applications for these reactors? Why
or why not?
Is there another standard the agency should consider for financial
qualification of merchant new reactor applicants? Commenters are
encouraged to provide specific suggestions and the basis for those
suggestions.
Part 73, Section 73.100--Physical Security
The proposed Sec. 73.100 would identify the proposed performance-
based physical security requirements with which future commercial power
reactor applicants or licensees' physical protection programs would
need to demonstrate compliance, without prescribing the specific
methods that must be used to satisfy them. Applicants and licensees
would have increased flexibility regarding the modern technologies and
methods that they could use. Implementing guidance in DG-5076 (proposed
RG 5.97), ``Guidance for Technology Inclusive Requirements for Physical
Protection of Licensed Activities at Commercial Nuclear Plants,'' would
be available to assist applicants and licensees. For example, DG-5076
provides detailed guidance, including performance standard
recommendations, on the probability of detection and alternative
sources of power for exterior intrusion detection systems (subsection
4.1.1.1.A), interior intrusion detection (subsection 4.1.1.1.B),
intrusion assessment (subsection 4.1.1.2.A), security response/
neutralization subsection (4.1.1.4.A), security communication
(subsection 4.1.1.3.A), and security delay (subsection 4.1.1.4.C).
Does the NRC's proposed approach in Sec. 73.100 provide a
sufficient level of detail to be readily understood and easily applied
to the licensing and oversight of new and advanced power reactors, or
should the NRC consider moving some objective and measurable security
performance standard recommendations from the draft implementing
guidance in DG-5076 into proposed Sec. 73.100? If so, which objective
and measurable security performance standard recommendations should be
moved from DG-5076 to Sec. 73.100? Please provide the basis for your
response.
[[Page 86988]]
Part 73, Section 73.110--Cybersecurity
The proposed Sec. 73.110 would require licensees to demonstrate
protection against cyberattacks in a manner that is commensurate with
the potential consequences from those attacks, without prescribing the
specific methods that must be used to demonstrate protection. Under
proposed Sec. 73.110(a), licensees would need to ensure that digital
computer and communications systems are adequately protected against a
potential cyberattack that would, for example, result in adverse
impacts to the physical security digital assets used by the licensee to
prevent unauthorized removal of material per Sec. 53.860(a).
Protecting against such a potential cyberattack would involve requiring
cybersecurity for SNM at a commercial nuclear reactor licensed under
part 53. Applicants and licensees would have increased flexibility
regarding the modern technologies and methods that they could use for
protecting against such a potential cyberattack. Detailed implementing
guidance in DG-5075 (proposed RG 5.96), ``Establishing Cybersecurity
Programs for Commercial Nuclear Plants licensed under 10 CFR part 53,''
would be available to assist applicants and licensees. For example, DG-
5075 provides guidance on the implementation of security by design
features (e.g., facility design) for negating the potential
consequences from such a potential cyberattack.
If a cyberattack were to compromise the availability, integrity, or
confidentiality of data or systems associated with security systems/
measures for the protection of SNM at a commercial nuclear reactor
licensed under part 53, do the potential consequences warrant requiring
cybersecurity for such material? Please provide the basis for your
response including a detailed explanation of challenges, if any, posed
by requiring cybersecurity for SNM at a commercial nuclear reactor
licensed under part 53.
Recent Legislation
On July 9, 2024, the President signed into law the Accelerating
Deployment of Versatile, Advanced Nuclear for Clean Energy Act of 2024,
also referred to as the ADVANCE Act. Section 203, ``Licensing
Considerations Relating to Use of Nuclear Energy for Nonelectric
Applications,'' and Section 208, ``Regulatory Requirements for Micro-
Reactors,'' of the ADVANCE Act specifically mention the technology-
inclusive regulatory framework to be established under section
103(a)(4) of NEIMA as a potential vehicle to be considered for the
report to Congress required under section 203 and a potential vehicle
to implement strategies and guidance for the licensing and regulation
of micro-reactors required under section 208. This proposed rulemaking
is, in part, how the NRC is implementing section 103(a)(4) of NEIMA.
The NRC is seeking comment on how part 53 could be revised to
better enable its potential use to implement the ADVANCE Act.
Specifically, Section 208 of the ADVANCE Act requires the NRC to
develop and implement ``risk-informed and performance-based strategies
and guidance'' in several areas for the licensing and regulation of
micro-reactors, including with respect to ``licensing mobile
deployment.'' The ADVANCE Act requires the NRC to consider ``the unique
characteristics of micro-reactors,'' including physical size, design
simplicity, and source term; opportunities to incorporate specific
improvements related to streamlining the review process; and other
policy and licensing issues. With regard to implementation, the ADVANCE
Act provides the NRC with three options. The NRC may implement the
developed strategies and guidance, as appropriate, via (1) the existing
regulatory framework, (2) the Part 53 rulemaking, or (3) a pending or
new rulemaking. Given the language included in Section 208, the NRC is
seeking comment on how part 53 could be revised to better address the
ADVANCE Act's requirements related to strategies and guidance for
micro-reactors.
VII. Section-by-Section Analysis
The following paragraphs describe the specific changes proposed by
this rulemaking.
Sec. 1.43 Office of Nuclear Reactor Regulation
This proposed rule would revise Sec. 1.43(a)(2) to extend the
authority of the Office of Nuclear Reactor Regulation to regulate
source, byproduct, and SNM at facilities licensed under part 53.
Sec. 2.1 Scope
This proposed rule would revise Sec. 2.1(e) to apply to standard
design approvals under part 53.
Sec. 2.4 Definitions
This proposed rule would revise Sec. 2.4 to update the definition
of ``Contested proceeding'' to include NRC enforcement actions against
applicants for a standard DC under part 53. It would also update the
definition of ``Facility'' to encompass utilization facilities as
defined in Sec. 53.020 (there are no production facilities under part
53).
Sec. 2.100 Scope of Subpart
This proposed rule would revise Sec. 2.100 to extend the scope of
subpart A to licenses and standard design approvals issued under
Sec. Sec. 53.1200 through 53.1221.
Sec. 2.101 Filing of Application
This proposed rule would revise Sec. 2.101 to be applicable to
part 53 applicants in addition to part 50 and 52 applicants by adding
references to part 53 in paragraphs (a)(3)(i), (a)(5), and (a)(9).
Sec. 2.104 Notice of Hearing
This proposed rule would extend the hearing notice requirement in
Sec. 2.104(a) to applications concerning facilities covered under part
53. Footnote 1 to Sec. 2.104 would be revised in a corresponding
manner.
Sec. 2.105 Notice of Proposed Action
This proposed rule would revise Sec. 2.105 to extend the
requirement in Sec. 2.104 to publish a notice of intended operation or
a notice of proposed action, as applicable, to part 53 applicants in
addition to part 50 and 52 applicants by adding corresponding
references to part 53 in paragraphs (a), (a)(4), (a)(10), (a)(12),
(a)(13), and (b)(3).
Sec. 2.106 Notice of Issuance
This proposed rule would revise Sec. 2.106 to extend the issuance
notice requirement to applications concerning facilities covered under
part 53 through updated references in paragraphs (a)(2) and (3), and
(b)(2).
Sec. 2.109 Effect of Timely Renewal Application
This proposed rule would revise Sec. 2.109 to add references to
part 53 in paragraphs (b), (c), and (d) regarding the timing of license
renewal applications.
Sec. 2.110 Filing and Administrative Action on Submittals for Standard
Design Approval or Early Review of Site Suitability Issues
This proposed rule would revise Sec. 2.110 to include references
to part 53 in paragraphs (a)(1) and (b).
Sec. 2.202 Orders
This proposed rule would revise Sec. 2.202(e) to add references to
part 53 regarding the requirements to be followed for orders involving
the modification of a license, COL, ESP, standard DC rule, standard
design approval, or ML.
[[Page 86989]]
Sec. 2.309 Hearing Requests, Petitions To Intervene, Requirements for
Standing, and Contentions
This proposed rule would revise Sec. 2.309 to include references
to part 53 in paragraphs (a), (f)(1)(i), (f)(1)(vi) and (vii), (g),
(h)(2), (i)(2), and (j) regarding a request for hearing under Sec.
53.1452.
Sec. 2.310 Selection of Hearing Procedures
This proposed rule would revise Sec. 2.310 by revising paragraph
(a), the introductory text for paragraph (h), and paragraphs (i) and
(j) to incorporate references to part 53 regarding hearing procedures.
Sec. 2.329 Prehearing Conference
This proposed rule would revise Sec. 2.329(a) to extend the timing
requirements for prehearing conferences involving CPs and licenses
under part 53.
Sec. 2.339 Expedited Decision-Making Procedure
This proposed rule would revise Sec. 2.339(d) to include
references to part 53 regarding expedited decision-making procedures.
Sec. 2.340 Initial Decision in Certain Contested Proceedings;
Immediate Effectiveness of Initial Decisions; Issuance of
Authorizations, Permits and Licenses
This proposed rule would revise Sec. 2.340 regarding initial
decisions of a presiding officer in certain contested proceedings, the
effective date of those decisions, and the issuance of authorizations,
permits, and licenses, by incorporating references to part 53 in
paragraphs (b), (c), (d), (f), (i), and (j).
Sec. 2.341 Review of Decisions and Actions of a Presiding Officer
This proposed rule would revise Sec. 2.341(a)(1) to include an
updated reference to part 53 regarding the allowance of a period of
interim operation.
Sec. 2.400 Scope of Subpart
This proposed rule would revise Sec. 2.400 to extend the scope of
subpart D of part 2 to include part 53 applicants for licenses to
construct or operate nuclear power reactors of identical design at
multiple sites.
Sec. 2.401 Notice of Hearing on Construction Permit or Combined
License Applications Pursuant to Appendix N of 10 CFR Parts 50, 52, or
53
This proposed rule would revise the section heading and Sec. 2.401
to extend the hearing notice requirement to applications concerning
facilities covered under part 53.
Sec. 2.402 Separate Hearings on Separate Issues; Consolidation of
Proceedings
This proposed rule would revise Sec. 2.402(a) to apply provisions
regarding separate hearings and the consolidation of proceedings to
part 53 applicants.
Sec. 2.403 Notice of Proposed Action on Applications for Operating
Licenses Pursuant To Appendix N of 10 CFR Part 50
This proposed rule would revise Sec. 2.403 to require the
Commission to publish a notification of proposed action in the Federal
Register after applications under part 53 are docketed.
Sec. 2.404 Hearings on Applications for Operating Licenses Pursuant to
Appendix N of 10 CFR Part 50
This proposed rule would revise Sec. 2.404 to apply to
applications for an OL under part 53.
Sec. 2.405 Initial Decisions in Consolidated Hearings
This proposed rule would revise Sec. 2.405 to be applicable to
CPs, full-power OLs, and COLs under part 53.
Sec. 2.406 Finality of Decisions on Separate Issues
This proposed rule would revise Sec. 2.406 to be applicable to
proceedings conducted pursuant to part 53.
Sec. 2.500 Scope of Subpart
This proposed rule would revise Sec. 2.500 to extend the
provisions of subpart E of part 2 to include applications for a license
to manufacture nuclear power reactors under part 53.
Sec. 2.501 Notice of Hearing on Application Under Subpart F of 10 CFR
Part 52 or 53 for a License To Manufacture Nuclear Power Reactors
This proposed rule would revise the section heading and Sec.
2.501(a) by extending its provisions to applications for a license to
manufacture nuclear power reactors under part 53.
Sec. 2.643 Acceptance and Docketing of Application for Limited Work
Authorization
This proposed rule would revise Sec. 2.643(b) regarding the
acceptance and docketing of an application for a CP for a utilization
facility of the type specified in part 53.
Sec. 2.645 Notice of Hearing
This proposed rule would revise Sec. 2.645(a) to incorporate a
reference to part 53.
Sec. 2.649 Partial Decisions on Limited Work Authorization
This proposed rule would revise Sec. 2.649 to extend its
provisions to LWAs issued under part 53.
Sec. 2.800 Scope and Applicability
This proposed rule would revise Sec. 2.800 by revising paragraphs
(c) and (d) to incorporate references to part 53 regarding the scope
and applicability of the rulemaking procedures contained in this
subpart.
Sec. 2.801 Initiation of Rulemaking
This proposed rule would revise Sec. 2.801 to include a reference
to part 53.
Sec. 2.813 Written Communications
This proposed rule would revise Sec. 2.813(a) to apply general
requirements for correspondence with the Commission to communications
concerning part 53, in addition to parts 50, 52, and 100.
Sec. 2.1103 Scope of Subpart K
This proposed rule would revise the first sentence of Sec. 2.1103
to extend the provisions of subpart K of part 2 to licenses under part
53 to expand the spent fuel capacity at the site of a civilian nuclear
power plant.
Sec. 2.1202 Authority and Role of NRC Staff
This proposed rule would amend Sec. 2.1202 by revising paragraphs
(a)(1) through (3), and (a)(6) to include references to part 53.
Sec. 2.1301 Public Notice of Receipt of a License Transfer Application
This proposed rule would revise Sec. 2.1301(b) to include a
corresponding reference to license transfers under part 53 in addition
to parts 50 and 52.
Sec. 2.1403 Authority and Role of the NRC Staff
This proposed role would update Sec. 2.1403 to specify that
``significant hazards considerations'' has the same meaning as defined
in part 53.
Sec. 2.1500 Purpose and Scope
This proposed rule would revise Sec. 2.1500 to extend the scope of
subpart O of part 2 to DC rulemaking hearings under part 53.
Sec. 2.1502 Commission Decision To Hold Legislative Hearing
This proposed rule would revise Sec. 2.1502, paragraphs (a) and
(b)(1) to
[[Page 86990]]
incorporate references to part 53 regarding the Commission's decision
to hold a DC rulemaking.
Sec. 10.1 Purpose
This proposed rule would revise Sec. 10.1(a)(3) to include a
reference to part 53.
Sec. 10.2 Scope
This proposed rule would revise Sec. 10.2(b) to extend the scope
of subpart A to applicants and holders of licenses, certificates, and
standard design approvals under part 53 in addition to part 52.
Sec. 11.7 Definitions
This proposed rule would revise Sec. 11.7 such that terms defined
in part 53 have the same meaning when used in part 11.
Sec. 19.2 Scope
This proposed rule would revise Sec. 19.2(a) to include references
to part 53.
Sec. 19.3 Definitions
This proposed rule would revise the definitions of ``License'' and
``Regulated entities'' in Sec. 19.3 to incorporate references to part
53.
Sec. 19.11 Posting of Notices to Workers
This proposed rule would amend Sec. 19.11 by revising paragraphs
(a), (b), and (e)(1) to apply to applicants and holders of licenses,
permits, standard design approvals, and standard DCs under part 53 in
addition to part 52.
Sec. 19.14 Presence of Representatives of Licensees and Regulated
Entities, and Workers During Inspections
This proposed rule would revise Sec. 19.14(a) to apply to
applicants and holders of a license, standard design approval, ESP, or
standard DC under part 53 in addition to part 52.
Sec. 19.20 Employee Protection
This proposed rule would revise Sec. 19.20 to include a reference
to protected activities under part 53.
Sec. 20.1002 Scope
This proposed rule would revise the first sentence of 10 CFR part
20, ``Standards for Protection Against Radiation,'' Sec. 20.1002 to
extend the scope of part 20 to apply to persons licensed by the
Commission to receive, use, transfer, or dispose of byproduct, source,
or SNM or to operate a production or utilization facility under part
53.
Sec. 20.1003 Definitions
This proposed rule would revise Sec. 20.1003 to update the
definition of ``License'' to include those issued under part 53.
Sec. 20.1101 Radiation Protection Programs
This proposed rule would revise Sec. 20.1101(d) to exclude
licensees subject to Sec. 53.260 from its requirements.
Sec. 20.1401 General Provisions and Scope
This proposed rule would revise Sec. 20.1401, paragraphs (a) and
(c) to extend the scope of subpart E of part 20 to apply to the
decommissioning of facilities licensed under part 53 and the release of
part of a facility or site for unrestricted use in accordance with
Sec. 53.1080.
Sec. 20.1403 Criteria for License Termination Under Restricted
Conditions
This proposed rule would revise Sec. 20.1403(d) to include
decommissioning plans under part 53.
Sec. 20.1404 Alternate Criteria for License Termination
This proposed rule would revise Sec. 20.1404(a)(4) to include a
reference to part 53 regarding alternate criteria for license
termination.
Sec. 20.1406 Minimization of Contamination
This proposed rule would revise Sec. 20.1406(a) to include
references to applicants for licenses other than ESPs or MLs under part
53. It would also revise Sec. 20.1406(b) to include references to
standard DCs and standard design approvals under part 53 in addition to
part 52.
Sec. 20.1501 General
This proposed rule would revise Sec. 20.1501(b) regarding the
requirement for retention of records from surveys describing the
location and amount of subsurface residual radioactivity at a site to
include a reference to the retention requirements under part 53.
Sec. 20.1905 Exemptions to Labeling Requirements
This proposed rule would revise Sec. 20.1905(g) to apply to
facilities licensed under part 53 in addition to parts 50 and 52
regarding exemptions to labeling requirements.
Sec. 20.2004 Treatment or Disposal by Incineration
This proposed rule would revise Sec. 20.2004(b)(1) to include
references to part 53 regarding the treatment or disposal of waste oil
by incineration.
Sec. 20.2201 Reports of Theft or Loss of Licensed Material
This proposed rule would revise Sec. 20.2201 to include references
to part 53 in paragraphs (a)(2)(i), (b)(2)(i) and (c) regarding
requirements for reports of theft or loss of licensed material.
Sec. 20.2202 Notification of Incidents
This proposed rule would revise Sec. 20.2202(d)(1) to add
references to part 53 regarding reports to the NRC Operations Center.
Sec. 20.2203 Reports of Exposures, Radiation Levels, and
Concentrations of Radioactive Material Exceeding the Constraints or
Limits
This proposed rule would revise Sec. 20.2203(c) to refer to
procedures under part 53 for reporting occurrences of exposures,
radiation levels, and concentrations of radioactive material exceeding
the constraints or limits.
Sec. 20.2206 Reports of Individual Monitoring
This proposed rule would revise Sec. 20.2206(a)(1) to include a
reference to part 53.
Sec. 21.2 Scope
This proposed rule would revise Sec. 21.2, paragraphs (a), (b),
and (c) to include references to part 53 regarding the scope and
applicability of part 21 requirements.
Sec. 21.3 Definitions
This proposed rule, in Sec. 21.3 would revise the definitions of
``Basic component,'' ``Commercial grade item,'' ``Critical
characteristics,'' ``Dedicating entity,'' ``Dedication,'' ``Defect,''
and ``Substantial safety hazard'' with references to part 53.
Sec. 21.21 Notification of Failure To Comply or Existence of a Defect
and Its Evaluation
This proposed rule would revise Sec. 21.21, by incorporating
references to part 53, to update the requirements for notifying the
Commission of a failure to comply or defect in paragraphs (a)(3) and
(d)(1).
Sec. 21.51 Maintenance and Inspection of Records
This proposed rule would revise Sec. 21.51(a)(4) and (5) to apply
to applicants for standard DC and applicants or holders of a standard
design approval under part 53, in addition to part 52, regarding the
retention of records.
[[Page 86991]]
Sec. 21.61 Failure To Notify
This proposed rule would revise Sec. 21.61(b) to include
references to part 53 licensees and applicants regarding failure to
provide the notice required in Sec. 21.21.
Sec. 25.5 Definitions
This proposed rule would update the definition of ``License'' to
include those issued under part 53.
Sec. 25.17 Approval for Processing Applicants for Access Authorization
This proposed rule would revise Sec. 25.17(a) to add a reference
to part 53 regarding AAs for individuals who need access to classified
information in connection with activities under part 53.
Sec. 25.35 Classified Visits
This proposed rule would update Sec. 25.35(a) to apply the
requirements for classified visits to licensees, certificate holders,
and applicants under part 53 in addition to part 52.
Sec. 26.3 Scope
This proposed rule would amend Sec. 26.3 by revising paragraph (d)
and adding new paragraph (f) which would establish the phase of
construction or operation by which applicants and licensees under part
53 would be required to comply with subpart M of part 26, or all of the
requirements of part 26 except subparts K and M.
Sec. 26.4 FFD Program Applicability to Categories of Individuals
This proposed rule would revise paragraphs (a), (b), (c), (e), (f),
(g), and (h) of Sec. 26.4 to include references to part 53 and
provisions for implementing an FFD program under subpart M.
Sec. 26.5 Definitions
This proposed rule would amend Sec. 26.5 by adding definitions for
``Biological marker,'' ``Change,'' ``Illicit substance,'' ``Reduction
in FFD program effectiveness,'' and ``Special Nuclear Material.'' It
would also revise definitions of ``Constructing or construction
activities,'' ``Contractor/vendor (C/V),'' ``Other entity,''
``Questionable validity,'' ``Reviewing official,'' ``Safety-related
structures, systems, and components (SSCs),'' ``Security-related
SSCs,'' and ``Unit outage'' within this section.
Sec. 26.8 Information Collection Requirements: OMB Approval
This proposed rule would revise Sec. 26.8(b) with the new
information collection requirements contained in proposed Sec. Sec.
26.202, 26.603, 26.604, 26.605, 26.606, 26.607, 26.608, 26.609, 26.611,
26.613, 26.617, and 26.619.
Sec. 26.21 Fitness-for-Duty Program
This proposed rule would revise Sec. 26.21 to include a reference
to Sec. 26.3(f).
Sec. 26.51 Applicability
This proposed rule would revise Sec. 26.51 to extend the
requirements of subpart C of part 26 to licensees and other entities
identified in Sec. 26.3(f) that do not implement the requirements of
subpart M of part 26, as well as licensees and other entities that
implement the requirements of Sec. 26.605.
Sec. 26.53 General Provisions
This proposed rule would revise Sec. 26.53 paragraphs (e), (g),
(h), and (i) to include references to Sec. 26.3(f).
Sec. 26.63 Suitable Inquiry
This proposed rule would revise Sec. 26.63(d) with a reference to
Sec. 26.3(f).
Sec. 26.73 Applicability
This proposed rule would revise Sec. 26.73 to extend the
requirements of subpart D of part 26 to licensees and other entities
identified in Sec. 26.3(f) that do not implement the requirements of
subpart M of part 26, as well as licensees and other entities that
implement the requirements of Sec. 26.605(b).
Sec. 26.81 Purpose and Applicability
This proposed rule would revise Sec. 26.81 to extend the
requirements of subpart E of part 26 to licensees and other entities
identified in Sec. 26.3(f) that do not implement the requirements of
subpart M of part 26, as well as licensees and other entities that
implement the requirements of Sec. 26.605.
Sec. 26.201 Applicability
This proposed rule would revise Sec. 26.201 to include references
to the proposed provisions in Sec. Sec. 26.3(f) and 26.202, as well as
revise the applicability of requirements in subpart I of part 26.
Sec. 26.202 General Provisions for Facilities Licensed Under Part 53
This proposed rule would add new Sec. 26.202, which would require
applicable licensees under part 53 to incorporate a policy for fatigue
management into their FFD program in accordance with the provisions of
this section.
Sec. 26.205 Work Hours
This proposed rule would revise paragraphs (d)(7)(iii) and (d)(8)
of Sec. 26.205 to incorporate references to Sec. Sec. 26.606 and
26.202(a) and (b).
Sec. 26.207 Waivers and Exceptions
This proposed rule would revise Sec. 26.207(a)(1)(ii) to include
references to Sec. Sec. 26.608 and 26.202(c) and to include provisions
for implementing certain face-to-face supervisor assessments using
electronic communications.
Sec. 26.211 Fatigue Assessments
This proposed rule would revise Sec. 26.211, paragraphs (a)(1),
(a)(3), and (b) to incorporate references to Sec. Sec. 26.202(c),
26.607(b), 26.608, and 26.619 and to include provisions for
implementing certain face-to-face assessments using electronic
communications.
Subpart M--Fitness for Duty Programs for Facilities Licensed Under Part
53
This proposed rule would add new Subpart M of part 26 containing
Sec. Sec. 26.601, 26.603, 26.604 through 26.611, 26.613, 26.615,
26.617, and 26.619, which adds an optional technology-inclusive, risk-
informed, and performance-based approach for the application of drug
and alcohol testing and fatigue management requirements for facilities
licensed under part 53.
Sec. 26.601 Applicability
This proposed rule would add Sec. 26.601, which would allow a
licensee or other entity in Sec. 26.3(f) to establish an FFD program
in accordance with the requirements of subpart M of part 26.
Sec. 26.603 General Provisions
This proposed rule would add Sec. 26.603, which would establish
the general requirements for implementing an FFD program under subpart
M of part 26.
Sec. 26.604 FFD Program Requirements for Facilities That Satisfy the
Sec. 26.603(c) Criterion
This proposed rule would add Sec. 26.604, which would establish
the FFD program elements for a licensee or other entity whose
facilities and operations demonstrate compliance with the criterion in
Sec. 26.603(c).
Sec. 26.605 FFD Program Requirements for Facilities That Do Not
Implement Sec. 26.604
This proposed rule would add Sec. 26.605, which would establish
the FFD program elements for a licensee or other entity that does not
demonstrate compliance with the criterion in Sec. 26.603(c), or
otherwise chooses to maintain an FFD program under this section.
[[Page 86992]]
Sec. 26.606 Written Policies and Procedures
This proposed rule would add Sec. 26.606, which would require
licensees and other entities that implement an FFD program under
subpart M of part 26 to develop a written FFD policy statement and
provide it to all individuals subject to the FFD program, and to
establish, implement, and maintain written procedures addressing the
topics outlined in this section.
Sec. 26.607 Drug and Alcohol Testing
This proposed rule would add Sec. 26.607, which would establish
requirements for licensees and other entities performing drug and
alcohol testing as part of an FFD program under subpart M of part 26.
Sec. 26.608 FFD Program Training
This proposed rule would add Sec. 26.608, which would require
individuals who are subject to the FFD program under subpart M of part
26 to receive periodic training on FFD policies and procedures,
including their duties and responsibilities under the BOP.
Sec. 26.609 Behavioral Observation
This proposed rule would add Sec. 26.609, which would establish
the requirements for a BOP under subpart M of part 26.
Sec. 26.610 Sanctions
This proposed rule would add Sec. 26.610, which would require
licensees and other entities implementing an FFD program under subpart
M of part 26 to establish sanctions for FFD policy violations.
Sec. 26.611 Protection of Information
This proposed rule would add Sec. 26.611, which would require
licensees and other entities implementing an FFD program under subpart
M of part 26 to establish a system to protect personal information
against unauthorized disclosure.
Sec. 26.613 Appeals Process
This proposed rule would add Sec. 26.613, which would require
licensees and other entities that implement an FFD program under
subpart M of part 26 to establish procedures for an individual to
appeal a policy violation determination.
Sec. 26.615 Audits
This proposed rule would add Sec. 26.615, which would establish
provisions for licensees and other entities that implement an FFD
program under subpart M of part 26 to conduct audits to monitor the
effectiveness of FFD program elements.
Sec. 26.617 Recordkeeping and Reporting
This proposed rule would add Sec. 26.617, which would require
licensees or other entities implementing an FFD program under subpart M
of part 26 to retain records pertaining to the administration of the
program and to make reports in accordance with the requirements of this
section.
Sec. 26.619 Suitability and Fitness Determinations
This proposed rule would add Sec. 26.619, which would require
licensees and other entities that implement FFD programs to develop,
implement, and maintain procedures to assess whether individuals are
fit to perform the duties that make them subject to the FFD program.
Sec. 26.709 Applicability
This proposed rule would designate the current paragraph as new
paragraph (a), and it would be revised to reference paragraphs (a)
through (d) of Sec. 26.3. It would also add paragraph (b) to Sec.
26.709, which would extend the requirements of subpart N of part 26 to
licensees and other entities identified in Sec. 26.3(f) that do not
implement the requirements of subpart M of part 26, as well as
licensees and other entities that implement the requirements of Sec.
26.605(b).
Sec. 26.711 General Provisions
This proposed rule would revise Sec. 26.711(c) and (d) to
incorporate a reference to Sec. 26.3(f).
Sec. 26.825 Criminal Penalties
This proposed rule would revise Sec. 26.825(b) to include a
reference to the proposed Sec. 26.601.
Sec. 30.4 Definitions
This proposed rule would revise the definition for ``Utilization
facility'' in Sec. 30.4 to include utilization facilities defined in
the regulations under part 53 in addition to part 50.
Sec. 30.50 Reporting Requirements
This proposed rule would revise Sec. 30.50(c)(3) to include
references to part 53 in addition to part 50.
Sec. 40.60 Reporting Requirements
This proposed rule would revise Sec. 40.60(c)(3) to include
references to part 53 in addition to part 50 regarding reporting
requirements.
Sec. 50.47 Emergency Plans
This proposed rule would revise Sec. 50.47(a)(1) and (e) with
appropriate references to part 53.
Sec. 50.54 Conditions of Licenses
This proposed rule would revise Sec. 50.54(q)(2), (q)(4), and
(gg)(1) with appropriate references to part 53.
Sec. 50.160 Emergency Preparedness for Small Modular Reactors, Non-
Light-Water Reactors, and Non-Power Production or Utilization
Facilities
This proposed rule would revise Sec. 50.160(b)(3) and (c)(2) with
the appropriate references to part 53.
Appendix B to 10 CFR Part 50--Quality Assurance Criteria for Nuclear
Power Plants and Fuel Reprocessing Plants
This proposed rule would revise appendix B to part 50 by revising
the introduction and specific criteria to incorporate the appropriate
references and terminology for part 53.
Sec. 51.20 Criteria for and Identification of Licensing and Regulatory
Actions Requiring Environmental Impact Statements
This proposed rule would revise Sec. 51.20(b)(1) and (2) to
require an EIS prior to the issuance of a CP, LWA, or ESP under part
53, or the issuance to renewal of a full power or design capacity
license to operate a nuclear power reactor, testing facility, or fuel
reprocessing plant under part 53.
Sec. 51.22 Criterion for Categorical Exclusion; Identification of
Licensing and Regulatory Actions Eligible for Categorical Exclusion or
Otherwise Not Requiring Environmental Review
This proposed rule would revise Sec. 51.22 to include
corresponding references to part 53 in paragraphs (c)(3), (c)(9),
(c)(12), (c)(17), (c)(22) and (23).
Sec. 51.26 Requirement To Publish Notice of Intent and Conduct Scoping
Process
This proposed rule would revise Sec. 51.26(d) to add a reference
to part 53.
Sec. 51.30 Environmental Assessment
This proposed rule would revise the introductory text to paragraph
(a) and revise paragraphs (d) and (e) of Sec. 51.30 to incorporate the
appropriate references to part 53 regarding EAs.
Sec. 51.31 Determinations Based on Environmental Assessment
This proposed rule would revise Sec. 51.31(a) to include a
reference to part 53.
[[Page 86993]]
Sec. 51.32 Finding of No Significant Impact
This proposed rule would revise Sec. 51.32(b)(1) and (3), finding
there is no significant environmental impact associated with the
issuance of standard DCs and MLs under part 53.
Sec. 51.49 Environmental Report-Limited Work Authorization
This proposed rule would revise the introductory text of Sec.
51.49(c) to require applicants for an ESP under part 53 requesting a
LWA to include the environmental report required by Sec. 51.50(b).
Sec. 51.50 Environmental Report--Construction Permit, Early Site
Permit, or Combined License Stage
This proposed rule would revise Sec. 51.50, paragraphs (a),
(b)(4), and the introductory text for paragraph (c) to incorporate the
appropriate references to part 53.
Sec. 51.53 Postconstruction Environmental Reports
This proposed rule would revise Sec. 51.53(d) to include the
appropriate references to part 53 regarding a license termination plan
or decommissioning plan and related requirements for postconstruction
environmental reports.
Sec. 51.54 Environmental Report--Manufacturing License
This proposed rule would update Sec. 51.54(a) to require
applicants for MLs under part 53 to submit an environmental report with
the application.
Sec. 51.55 Environmental Report--Standard Design Certification
This proposed rule would update Sec. 51.55(a) to require
applicants for a standard DC under part 53 to submit an environmental
report with the application.
Sec. 51.58 Environmental Report--Number of Copies; Distribution
This proposed rule would revise Sec. 51.58(b) to incorporate the
appropriate references to part 53.
Sec. 51.77 Distribution of Draft Environmental Impact Statement
This proposed rule would revise the introductory text for Sec.
51.77(a) to add a reference to part 53.
Sec. 51.92 Supplement to the Final Environmental Impact Statement
This proposed rule would revise Sec. 51.92(b) to apply to COL
applications referencing an ESP under part 53.
Sec. 51.95 Postconstruction Environmental Impact Statements
This proposed rule would revise the introductory text for Sec.
51.95(c) to include a reference to part 53 regarding the Commission's
obligations to prepare an EIS following the renewal of an operating or
COL for a nuclear plant under part 53.
Sec. 51.101 Limitations on Actions
This proposed rule would revise Sec. 51.101(a)(2) to include the
corresponding references to part 53 where appropriate.
Sec. 51.103 Record of Decision--General
This proposed rule would update Sec. 51.103(a)(6) to apply to the
issuance of a LWA in connection with a CP or COL under part 53.
Sec. 51.105 Public Hearings in Proceedings for Issuance of
Construction Permits or Early Site Permits; Limited Work Authorizations
This proposed rule would revise Sec. 51.105(c)(1) to include the
appropriate reference to LWAs under part 53 for CPs or ESPs.
Sec. 51.107 Public Hearings in Proceedings for Issuance of Combined
Licenses; Limited Work Authorizations
This proposed rule would amend Sec. 51.107 by revising the
introductory text for paragraphs (a) and (b) and updating paragraph
(d)(1) to include the appropriate corresponding references to part 53.
Sec. 51.108 Public Hearings on Commission Findings That Inspections,
Tests, Analyses, and Acceptance Criteria of Combined Licenses Are Met
This proposed rule would revise Sec. 51.108 to incorporate the
appropriate references to part 53.
10 CFR part 53--Risk-Informed, Technology-Inclusive Regulatory
Framework for Commercial Nuclear Plants
This proposed rule would add a new part to 10 CFR Chapter I,
designated as Part 53 including Sec. Sec. 53.000 through 53.9010.
Sec. 53.000 Purpose
This proposed rule would add Sec. 53.000 which provides an
optional technology-inclusive, performance-based framework for the
issuance, amendment, renewal, and termination of licenses, permits,
certifications, and approvals for commercial nuclear plants licensed
under section 103 of the Atomic Energy Act of 1954, as amended.
Subpart A--General Provisions
This proposed rule would add subpart A, to establish a set of
general provisions, which apply to all applicants and licensees under
part 53.
Sec. 53.015 Scope
This proposed rule would add Sec. 53.015, which would extend the
provisions of subpart A to all applicants and licensees under part 53.
Sec. 53.020 Definitions
This proposed rule would add Sec. 53.020, which would define key
terms in part 53.
Sec. 53.040 Written Communications
This proposed rule would add Sec. 53.040, which would govern how
applicants and licensees submit written communications to the NRC,
including applications, submissions related to the security plans,
emergency plan, and quality assurance, certifications of permanent
cessation of operations and permanent fuel removal, and other
submittals required under part 53.
Sec. 53.050 Deliberate Misconduct
This proposed rule would add Sec. 53.050, which would prohibit
licensees or applicants, contractors and subcontractors, or employees
of those entities from deliberately violating NRC rules, regulations,
or orders, or the terms, conditions, and limitations of a part 53
license. This proposed rule would also prohibit deliberate submissions
of incomplete or inaccurate information. Violations would be subject to
enforcement actions under subpart B of part 2.
Sec. 53.060 Employee Protection
This proposed rule would add Sec. 53.060, which would prohibit
applicants and licensees from discriminating against employees for
engaging in the protected activities listed in this section and provide
remedial procedures for employees who believe they are the subjects of
discrimination.
Sec. 53.070 Completeness and Accuracy of Information
This proposed rule would add Sec. 53.070, which would require
licensees and applicants under part 53 to provide complete and accurate
information in accordance with all applicable laws, Commission
regulations, and the terms and conditions of their license. This
proposed rule would also require licensees to notify the Commission
within two days of identifying information with material implications
[[Page 86994]]
for public health and safety or common defense and security.
Sec. 53.080 Specific Exemptions
This proposed rule would add Sec. 53.080, which would establish
the special circumstances under which the Commission could grant
exemptions to part 53 licensees and the Commission's criteria for
making such a determination.
Sec. 53.090 Standards for Review
This proposed rule would add Sec. 53.090 to establish the
standards that the Commission would consider when determining whether
to issue a permit or license under part 53.
Sec. 53.100 Jurisdictional Limits
This proposed rule would add Sec. 53.100, which would provide that
permits, licenses, standard design approvals, and standard DCs are
solely issued for activities within the jurisdiction of the United
States.
Sec. 53.110 Attacks and Destructive Acts
This proposed rule would add Sec. 53.110, which would exempt
licensees or applicants under part 53 from providing design features to
protect against attacks or destructive acts directed at the facility by
United States adversaries.
Sec. 53.115 Rights Related to Special Nuclear Material
This proposed rule would add Sec. 53.115, which would establish
provisions regarding the rights to SNM under a part 53 license.
Sec. 53.117 License Suspension and Rights of Recapture
This proposed rule would add Sec. 53.117, which would provide that
the Commission may suspend licenses and recapture material or control
of a facility in a state of war or national emergency declared by
Congress.
Sec. 53.120 Information Collection Requirements: OMB Approval
This proposed rule would add Sec. 53.120, which would establish
requirements for information collection requirements and Office of
Budget and Management approval.
Subpart B--Technology-Inclusive Safety Requirements
This proposed rule would add subpart B, to establish a set of
technology-inclusive performance standards that would be used
throughout part 53 to determine appropriate regulatory controls for
SSCs, human actions, and programs.
Sec. 53.210 Safety Criteria for Design-Basis Accidents
This proposed rule would add Sec. 53.210 to set dose values to
ensure that plants are designed to limit the public's radiation
exposure in the event of a DBA.
Sec. 53.220 Safety Criteria for Licensing-Basis Events Other Than
Design-Basis Accidents
This proposed rule would add Sec. 53.220 to require plants to
implement a combination of design features and programmatic controls to
control risks to the public in the event of a LBE other than a DBA.
Sec. 53.230 Safety Functions
This proposed rule would add Sec. 53.230, which specifies that
limiting the release of radioactive materials from the facility is the
primary safety function of a commercial nuclear plant, and that
additional safety functions must be defined to support the retention of
radioactive materials during LBEs.
Sec. 53.240 Licensing-Basis Events
This proposed rule would add Sec. 53.240 to require commercial
nuclear plants to conduct an analysis of LBEs to confirm that design
features and programmatic controls satisfy the safety criteria under
Sec. Sec. 53.210 and 53.220, or alternatively, under Sec. 53.470.
Sec. 53.250 Defense in Depth
This proposed rule would add Sec. 53.250 to establish a
performance-based, defense-in-depth approach to address uncertainties
about the effectiveness and reliability of plant SSCs, personnel, and
programmatic controls.
Sec. 53.260 Normal Operations
This proposed rule would add Sec. 53.260, requiring holders of
licenses to operate commercial nuclear plants to control public doses
and dose rates in unrestricted areas to meet the requirements in part
20, during normal plant operation.
Sec. 53.270 Protection of Plant Workers
This proposed rule would add Sec. 53.270, requiring holders of
licenses to operate commercial nuclear plants to control occupational
doses to meet the requirements in part 20.
Subpart C--Design and Analysis Requirements
This proposed rule would add subpart C, which requires the
implementation of certain design features and the performance of risk
assessments and analyses to demonstrate compliance with the safety
criteria and safety functions in subpart B.
Sec. 53.400 Design Features for Licensing-Basis Events
This proposed rule would add Sec. 53.400, which would require
design features that satisfy the safety criteria defined in Sec.
53.210 and Sec. 53.220 or Sec. 53.470 and fulfill the safety
functions identified in Sec. 53.230 during LBEs.
Sec. 53.410 Functional Design Criteria for Design-Basis Accidents
This proposed rule would add Sec. 53.410, which would stipulate
that functional design criteria must be defined for each design feature
required by Sec. 53.400 to demonstrate compliance with the safety
criteria defined in Sec. 53.210 for DBAs.
Sec. 53.415 Protection Against External Hazards
This proposed rule would add Sec. 53.415, which would require SR
SSCs to be designed to withstand the effects of natural phenomena and
constructed hazards while performing the intended safety functions.
Sec. 53.420 Functional Design Criteria for Licensing-Basis Events
Other Than Design-Basis Accidents
This proposed rule would add Sec. 53.420, which would require
functional design criteria to be defined for each design feature
required by Sec. 53.400 to demonstrate compliance with the safety
criteria defined in Sec. 53.220 for LBEs other than DBAs.
Sec. 53.425 Design Features and Functional Design Criteria for Normal
Operations
This proposed rule would add Sec. 53.425, which would require
commercial nuclear plants to implement design features and define
functional design criteria sufficient to demonstrate compliance with
Sec. 53.850 and show through functional design criteria that design
features and corresponding programmatic controls control wastes, as
required under part 20.
Sec. 53.430 Design Features and Functional Design Criteria for
Protection of Plant Workers
This proposed rule would add Sec. 53.430, which would require
commercial nuclear plants to implement design features and define
functional design criteria sufficient to demonstrate compliance with
Sec. 53.270.
Sec. 53.440 Design Requirements
This proposed rule would add Sec. 53.440, which would establish
various
[[Page 86995]]
design feature requirements, including protection against fires and
explosions, criticality accidents, and the impact of a large commercial
aircraft.
Sec. 53.450 Analysis Requirements
This proposed rule would add Sec. 53.450, which would require
commercial nuclear plants to perform PRAs in combination with other
analytical methods to identify and assess risks and determine
compliance with the safety criteria in subpart B. In addition, Sec.
53.450 would require analysis of DBAs and other analyses to assess the
adequacy of protections against fire, aircraft impact, and the release
of effluents.
Sec. 53.460 Safety Categorization and Special Treatments
This proposed rule would add Sec. 53.460 to address the safety
classification of SSCs and determine appropriate special treatments.
Sec. 53.470 Maintaining Analytical Safety Margins Used To Justify
Operational Flexibilities
This proposed rule would add Sec. 53.470 to permit applicants and
licensees to implement more restrictive criteria than that defined in
Sec. Sec. 53.220 and 53.450(e) to support operational flexibilities.
Sec. 53.480 Earthquake Engineering
This proposed rule would add Sec. 53.480 to provide overall
seismic design considerations based on the safety criteria in subpart B
and siting requirements in subpart D to ensure that SSCs are able to
withstand the effects of earthquakes without loss of capability to
fulfill safety functions.
Subpart D--Siting Requirements
This proposed rule would add subpart D, which would address
requirements associated with the siting of commercial nuclear
facilities under part 53, including considerations of external hazards
and potential adverse impacts on the surrounding population.
Sec. 53.500 General Siting and Siting Assessment
This proposed rule would add Sec. 53.500, which would require a
siting assessment for each commercial nuclear plant to ensure that
design features and programmatic controls are sufficient to address
LBEs and mitigate potential adverse impacts of the plant on the
surrounding environs.
Sec. 53.510 External Hazards
This proposed rule would add Sec. 53.510, which would require
site-specific assessments, including an evaluation of geological and
seismic siting factors, to identify and characterize the external
hazard level for a range of natural and constructed hazards.
Sec. 53.520 Site Characteristics
This proposed rule would add Sec. 53.520, which would require the
design and analyses conducted under subpart C to consider how site
characteristics may contribute to LBEs.
Sec. 53.530 Population-Related Considerations
This proposed rule would add Sec. 53.530, which would establish
requirements related to the facility's exclusion area, low-population
zone, and population center distance.
Sec. 53.540 Siting Interfaces
This proposed rule would add Sec. 53.540, which would require that
external hazards and site characteristics must be accounted for in the
design features, programmatic controls, and supporting analyses used to
demonstrate compliance with the safety criteria in Sec. Sec. 53.210
and 53.220.
Subpart E--Construction and Manufacturing Requirements
This proposed rule would add subpart E, which would establish
requirements for the construction and manufacture of commercial nuclear
plants.
Sec. 53.600 Construction and Manufacturing--Scope and Purpose
This proposed rule would add Sec. 53.600, which would indicate
that this subpart applies to construction and manufacturing activities
authorized by a CP, COL, ML, or LWA issued under this part.
Sec. 53.605 Reporting of Defects and Noncompliance
This proposed rule would add Sec. 53.605, which would describe the
procedures, notification requirements, and records retention
requirements that each CP, ML, and COL is subject to with respect to
reporting of defects and noncompliance.
Sec. 53.610 Construction
This proposed rule adds Sec. 53.610 to address the management and
control of the construction of a commercial nuclear plant, including
specific requirements for procedures and quality assurance, control of
radioactive materials, and post construction inspections.
Sec. 53.620 Manufacturing
This proposed rule would add Sec. 53.620, which would ensure that
the holders of an ML under part 53 develop plans, programs, and
organizational units to manage and control manufacturing activities,
and would establish requirements for the loading of fuel into a
manufactured reactor for subsequent transport to a commercial nuclear
plant and operation pursuant to a COL.
Subpart F--Requirements for Operation
This proposed rule would add subpart F, which would establish
regulatory requirements to ensure that the safety criteria in subpart B
are satisfied whenever a commercial nuclear plant licensed under part
53 is operational. This includes periods of normal operation and
unplanned events.
Sec. 53.700 Operational Objectives
This proposed rule would add Sec. 53.700, which would establish
general operational objectives to ensure that licensees under part 53
have implemented and maintained the SSCs necessary to demonstrate
compliance with the safety functions identified in subpart B for
addressing normal operations and responding to LBEs.
Sec. 53.710 Maintaining Capabilities and Availability of Structures,
Systems, and Components
This proposed rule would add Sec. 53.710, which would require
licensees under part 53 to demonstrate compliance with the safety
criteria in subpart B by establishing TS for all SR SSCs and developing
documents and procedures for all NSRSS SSCs.
Sec. 53.715 Maintenance, Repair, and Inspection Programs
This proposed rule would add Sec. 53.715, which would require
licensees to develop, implement, and maintain programs to assess and
manage any risks posed by maintenance activities and to evaluate the
efficacy of performance, condition monitoring, and maintenance
activities.
Sec. 53.720 Response to Seismic Events
This proposed rule would add Sec. 53.720, which would establish
requirements for licensees to respond to a seismic event during the
operating phase of the life cycle of a commercial nuclear plant.
Sec. 53.725 General Staffing, Training, Personnel Qualifications, and
Human Factors Requirements
This proposed rule would add Sec. 53.725, which would provide an
[[Page 86996]]
overview of the staffing, training, personnel qualifications, and human
factors requirements established in Sec. Sec. 53.725 through 53.830
and would provide definitions of ``Automation,'' ``Auxiliary
operator,'' ``Controls,'' ``Generally licensed reactor operator,''
``Load following,'' ``Operator,'' ``Performance testing,'' ``Reference
plant,'' ``Self-reliant mitigation facility,'' ``Senior operator,''
``Simulation facility,'' and ``Systems approach to training.'' Proposed
Sec. Sec. 53.725 through 53.830 would apply to applicants for or
holders of OLs or COLs under part 53.
Sec. 53.726 Communications
This proposed rule would add Sec. 53.726, which would contain
communications requirements applicable to sections Sec. Sec. 53.725
through 53.830. It also contains requirements to notify the Commission
within 30 days should a specifically licensed operator or senior
operator be reassigned, terminated, or suffer permanent disability or
illness.
Sec. 53.728 Completeness and Accuracy of Information
This proposed rule would add Sec. 53.728, which would require
submitted information to be complete and accurate in all material
respects.
Sec. 53.730 Defining, Fulfilling, and Maintaining the Role of
Personnel in Ensuring Safe Operations
This proposed rule would add Sec. 53.730, which would establish
technical requirements for applicants or holders of OLs or COLs within
the areas of HFE, human-system interface design, concept of operations,
functional requirements analysis, function allocation, operating
experience, procedures, staffing, operator training, operator
examinations, and operator proficiency.
Sec. 53.735 General Exemptions
This proposed rule would add Sec. 53.735, which would establish
general exemptions for licensed operators.
Sec. 53.740 Facility Licensee Requirements--General
This proposed rule would add Sec. 53.740, which would establish
staffing requirements for interaction-dependent-mitigation facilities
and self-reliant mitigation facilities.
Sec. 53.745 Operator License Requirements
This proposed rule would add Sec. 53.745, which would require
individuals to be licensed to perform certain functions.
Sec. 53.760 Operator Licensing
This proposed rule would add Sec. 53.760, which would address the
applicability of the requirements of Sec. Sec. 53.760 through 53.795
for specifically licensed operators and senior operators.
Sec. 53.765 Medical Requirements
This proposed rule would add Sec. 53.765, which would establish
medical requirements for specifically licensed operators and senior
operators.
Sec. 53.770 Incapacitation Because of Disability or Illness
This proposed rule would add Sec. 53.770, which would establish
requirements to address permanent medical conditions for specifically
licensed operators and senior operators.
Sec. 53.775 Applications for Operators and Senior Operators
This proposed rule would add Sec. 53.775, which would establish
the application process and requirements for individuals applying for
specific operator and senior operator licenses.
Sec. 53.780 Training, Examination, and Proficiency Program
This proposed rule would add Sec. 53.780, which would contain the
requirements associated with specifically licensed operator and senior
operator initial training, initial examinations, requalification
training, requalification examinations, examination integrity,
simulation facilities, waivers, and proficiency.
Sec. 53.785 Conditions of Operator and Senior Operator Licenses
This proposed rule would add Sec. 53.785, which would establish
conditions for specific operator and senior operator licenses.
Sec. 53.790 Issuance, Modification, and Revocation of Operator and
Senior Operator Licenses
This proposed rule would add Sec. 53.790, which would contain
requirements associated with the issuance, modification, or revocation
of specific operator and senior operator licenses.
Sec. 53.795 Expiration and Renewal of Operator and Senior Operator
Licenses
This proposed rule would add Sec. 53.795, which would contain
requirements associated with the expiration and renewal of specific
operator and senior operator licenses.
Sec. 53.800 Facility Licensees for Self-Reliant-Mitigation Facilities
This proposed rule would add Sec. 53.800, which would establish
the technical criteria by which commercial nuclear plants under part 53
are determined to be of the self-reliant mitigation class of facilities
that would be staffed by GLROs in lieu of specifically licensed
operators and senior operators.
Sec. 53.805 Facility Licensee Requirements Related to Generally
Licensed Reactor Operators
This proposed rule would add Sec. 53.805, which would establish
requirements that apply to the facility licensee at those facilities
staffed by GLROs.
Sec. 53.810 Generally Licensed Reactor Operators
This proposed rule would add Sec. 53.810, which would issue and
describe the general license for GLROs that manipulate the controls of
a self-reliant mitigation facility.
Sec. 53.815 Generally Licensed Reactor Operator Training, Examination,
and Proficiency Programs
This proposed rule would add Sec. 53.815, which would contain the
requirements for GLRO initial training, initial examinations,
continuing training, requalification examinations, examination
integrity, simulation facilities, examination waivers, and proficiency.
Sec. 53.820 Cessation of Individual Applicability
This proposed rule would add Sec. 53.820, which would address the
requirements by which the general license for GLROs would cease to be
applicable on an individual basis.
Sec. 53.830 Training and Qualification of Commercial Nuclear Plant
Personnel
This proposed rule would add Sec. 53.830, which would address
training and qualification requirements for supervisors, technicians,
and other appropriate operating personnel at commercial nuclear plants.
Sec. 53.845 Programs
This proposed rule would add Sec. 53.845, which would require
licensees under part 53 to establish programs that include, but are not
limited to, radiation protection, emergency preparedness, security,
quality assurance, integrity assessment, fire protection, ISI and IST,
and facility safety, to ensure that the safety criteria and functions
in subpart
[[Page 86997]]
B are maintained during normal operations and LBEs.
Sec. 53.850 Radiation Protection
This proposed rule would add Sec. 53.850, which would require
licensees under part 53 to implement and maintain programs and
processes to limit and monitor radioactive plant effluents and limit
the exposure of plant personnel and the public.
Sec. 53.855 Emergency Preparedness
This proposed rule would add Sec. 53.855, which would require
licensees under this part to have an emergency response plan for
radiological emergencies.
Sec. 53.860 Security Programs
This proposed rule would add Sec. 53.860, which would require
licensees under part 53 to develop, implement, and maintain programs
for physical security, FFD, AA, cybersecurity, and information
security.
Sec. 53.865 Quality Assurance
This proposed rule would add Sec. 53.865, which would require
licensees under part 53 to establish a quality assurance program that
includes a written manual to ensure activities are conducted in
accordance with codes and standards found acceptable by the NRC.
Sec. 53.870 Integrity Assessment Programs
This proposed rule would add Sec. 53.870, which would require
licensees under part 53 to establish an integrity assessment program to
ensure that the plant continues to fulfill safety criteria and
functional design criteria as it ages.
Sec. 53.875 Fire Protection
This proposed rule would add Sec. 53.875, which would require
licensees under part 53 to establish a fire protection plan and
describe the necessary elements that the plan must incorporate.
Sec. 53.880 Inservice Inspection and Inservice Testing
This proposed rule would add Sec. 53.880, which would require
licensees under part 53 to develop and implement a program for ISI and
IST in accordance with the requirements of this section.
Sec. 53.910 Procedures and Guidelines
This proposed rule would add Sec. 53.910, which would require
licensees under part 53 to develop, maintain, and implement procedures
and guidelines that address normal plant operations and responses to
unplanned events.
Subpart G--Decommissioning Requirements
This proposed rule would add subpart G, to establish
decommissioning requirements for applicants for or holders of an OL or
COL under part 53.
Sec. 53.1000 Scope and Purpose
This proposed rule would add Sec. 53.1000, which would establish
the scope of the decommissioning requirements for applicants and
licensees under part 53 and describe the contents of subpart G of part
53.
Sec. 53.1010 Financial Assurance for Decommissioning
This proposed rule would add Sec. 53.1010, which would establish
the requirement that applicants for an OL or COL under part 53 provide
reasonable assurance that funds will be available for the
decommissioning process. This section would describe the requirements
associated with the required plan and an associated decommissioning
report that ensures and documents that adequate funding for
decommissioning will be available.
Sec. 53.1020 Cost Estimates for Decommissioning
This proposed rule would add Sec. 53.1020, which would require
site-specific cost estimates for decommissioning and establish the
aspects that must be included in the estimate.
Sec. 53.1030 Annual Adjustments to Cost Estimates for Decommissioning
This proposed rule would add Sec. 53.1030, which would require
that holders of an OL or COL under part 53 annually adjust their cost
estimate for decommissioning to account for escalation in labor,
energy, and waste burial costs. This section would allow licensees to
elect either a site-specific adjustment factor or a generic adjustment
factor.
Sec. 53.1040 Methods for Providing Financial Assurance for
Decommissioning
This proposed rule would add Sec. 53.1040, which would establish
suitable methods that holders of an OL or COL under part 53 may use to
provide financial assurance for decommissioning to the NRC.
Sec. 53.1045 Limitations on the Use of Decommissioning Trust Funds
This proposed rule would add Sec. 53.1045, which would establish
requirements for decommissioning trust funds under part 53, including
criteria for using decommissioning trust funds and required terms.
Sec. 53.1050 NRC Oversight
This proposed rule would add Sec. 53.1050, which would outline the
steps the NRC may take to ensure adequate accumulation of
decommissioning funds.
Sec. 53.1060 Reporting and Recordkeeping Requirements
This proposed rule would add Sec. 53.1060, which would contain
reporting and recordkeeping requirements related to decommissioning for
each holder of an OL or COL under part 53. This section would outline
requirements for documents such as: certification of decommissioning
funding, decommissioning cost estimates and copies of financial
instruments, licensee records of information important to safe and
effective decommissioning, post-shutdown decommissioning activities
report, financial assurance reports, and reports on the status of
funding for managing irradiated fuel.
Sec. 53.1070 Termination of License
This proposed rule would add Sec. 53.1070, which would establish
procedures for decommissioning and license termination applicable to
licensees under part 53 that have determined to permanently cease
operations.
Sec. 53.1075 Program Requirements During Decommissioning
This proposed rule would add Sec. 53.1075, which would require
licensees under part 53 to establish and maintain a decommissioning
fire protection program to prevent, detect, and control fires, and
ensure that the risk of fire induced radiological hazards are minimized
through the various stages of facility decommissioning.
Sec. 53.1080 Release of Part of a Commercial Nuclear Plant or Site for
Unrestricted Use
This proposed rule would add Sec. 53.1080, which would establish
licensee procedures for requesting and NRC procedures for approving
partial release of a commercial nuclear plant or site for unrestricted
use prior to receiving approval of a license termination plan from the
Commission under part 53.
Subpart H--Licenses, Certifications, and Approvals
This proposed rule would add subpart H, which would govern the
process of applying for, amending, renewing, or
[[Page 86998]]
terminating a LWA, ESP, standard design approval, standard DC, ML, CP,
OL, or COL under part 53.
Sec. 53.1100 Filling of Application for Licenses, Certifications, or
Approvals; Oath or Affirmation
This proposed rule would add Sec. 53.1100, which would establish
requirements for applicants seeking a standard design approval,
standard DC, license, or permit under part 53 to submit an application.
Sec. 53.1101 Requirement for License
This proposed rule would add Sec. 53.1101, which would prohibit
any use of a utilization facility except as authorized by a license
issued by the NRC or by an exception as described in Sec. 53.1120.
Sec. 53.1103 Combining Applications and Licenses
This proposed rule would add Sec. 53.1103, which would permit
applicants under part 53 seeking multiple licenses to submit a single
application, and the Commission to issue a single license for
activities that would otherwise be licensed separately.
Sec. 53.1106 Elimination of Repetition
This proposed rule would add Sec. 53.1106, which would allow
applicants under part 53 to reference information contained in previous
documents filed with the Commission so long as those references are
clear and specific.
Sec. 53.1109 Contents of Applications; General Information
This proposed rule would add Sec. 53.1109, which would establish
the general content to be included in applications made under part 53,
including but not limited to the identifying information of the
applicant and the radiological emergency response plans of government
entities within the plume exposure pathway EPZ.
Sec. 53.1112 Environmental Conditions
This proposed rule would add Sec. 53.1112, which would allow the
Commission to attach conditions to CPs, ESPs, and licenses issued under
part 53 to address environmental issues during construction, operation,
or decommissioning. These conditions will be derived from the
information contained in the environmental report submitted as part of
the application for a permit or license.
Sec. 53.1115 Agreement Limiting Access to Classified Information
This proposed rule would add Sec. 53.1115, which would require
applicants to agree in writing, prior to receiving a license or
standard design approval under part 53, to restrict any facilities, or
any individuals with access to plant facilities, from possessing
Restricted Data or classified National Security Information until they
have received the appropriate authorization.
Sec. 53.1118 Ineligibility of Certain Applicants
This proposed rule would add Sec. 53.1118, which would prevent
citizens, nationals, or agents of a foreign country or corporations
owned, controlled, or dominated by a foreign entity from applying for
or obtaining a license under part 53.
Sec. 53.1120 Exceptions and Exemptions From Licensing Requirements
This proposed rule would add Sec. 53.1120, which would establish
the activities that are exempt from licensing requirements.
Sec. 53.1121 Public Inspection of Applications
This proposed rule would add Sec. 53.1121, which would allow
applicant submissions to be made publicly available under the
provisions of part 2.
Sec. 53.1124 Relationship Between Sections
This proposed rule would add Sec. 53.1124, which would outline the
relationship between LWAs, ESPs, standard design approvals, standard
DCs, MLs, CPs, OLs, and COLs under part 53.
Sec. 53.1130 Limited Work Authorizations
This proposed rule would add Sec. 53.1130, which would establish
requirements for requesting an LWA and grounds for the Commission to
issue an LWA. It would also contain details about the effect of an LWA
and the implementation of a redress plan.
Sec. 53.1140 Early Site Permits
This proposed rule would add Sec. 53.1140, which would provide an
overview of the requirements regarding applications for and the
issuance of ESPs under part 53.
Sec. 53.1143 Filing of Applications
This proposed rule would add Sec. 53.1143, which would enable an
applicant under part 53 to apply for an ESP, regardless of whether they
have filed an application for a CP or COL for that site.
Sec. 53.1144 Contents of Applications for Early Site Permits; General
Information
This proposed rule would add Sec. 53.1144, which would require
applications for ESPs to include the information required by Sec.
53.1109(a) through (d) and (j).
Sec. 53.1146 Contents of Applications for Early Site Permits;
Technical Information
This proposed rule would add Sec. 53.1146, which would require
applicants for ESPs to submit technical information, including but not
limited to a Site Safety Analysis Report and emergency plans.
Sec. 53.1149 Review of Applications
This proposed rule would add Sec. 53.1149, which would establish
standards for review of applications for ESPs under part 53, including
requirements for the Commission to prepare an EIS and assess the
adequacy of protective actions in the event of a radiological
emergency. It would also require the administrative review of
applications and hearings to follow the procedural requirements of part
2.
Sec. 53.1155 Referral to the Advisory Committee on Reactor Safeguards
This proposed rule would add Sec. 53.1115, which would require the
ACRS to review SR content in the application for an ESP under part 53.
Sec. 53.1158 Issuance of Early Site Permit
This proposed rule would add Sec. 53.1158, which would establish
the conditions under which the Commission may issue an ESP under part
53, as well as the information, terms, and conditions to be included in
the permit.
Sec. 53.1161 Extent of Activities Permitted
This proposed rule would add Sec. 53.1161, which would require
that a valid ESP only be used for the purpose of site redress, unless
the site is referenced in an application for a CP or COL under part 53.
Sec. 53.1164 Duration of Permit
This proposed rule would add Sec. 53.1164, which would govern the
conditions under which an ESP remains valid following the date of
issuance.
Sec. 53.1167 Limited Work Authorization After Issuance of Early Site
Permit
This proposed rule would add Sec. 53.1167, which would permit the
[[Page 86999]]
holder of an ESP to request a LWA under Sec. 53.1130.
Sec. 53.1170 Transfer of Early Site Permit
This proposed rule would add Sec. 53.1170, which would govern the
transfer of an ESP in accordance with Sec. 53.1570.
Sec. 53.1173 Application for Renewal
This proposed rule would add Sec. 53.1173, which would establish
the conditions and procedures for renewing an ESP under part 53.
Sec. 53.1176 Criteria for Renewal
This proposed rule would add Sec. 53.1176, which would establish
the criteria that the Commission may use to grant a renewal of an ESP
under part 53.
Sec. 53.1179 Duration of Renewal
This proposed rule would add Sec. 53.1179, which would govern the
duration of a renewed ESP under part 53.
Sec. 53.1182 Use of Site for Other Purposes
This proposed rule would add Sec. 53.1182, which would govern
acceptable uses of the site for purposes other than those described in
the permit.
Sec. 53.1188 Finality of Early Site Permit Determinations
This proposed rule would add Sec. 53.1188, which would address the
finality of ESP determinations under part 53.
Sec. 53.1200 Standard Design Approvals
This proposed rule would add Sec. 53.1200, which would address the
procedures for filing an application for a standard design approval
under part 53, the process of review by NRC staff, and referral to the
ACRS of standard designs.
Sec. 53.1203 Filing of Applications
This proposed rule would add Sec. 53.1203, which would enable
applicants to submit a final design for the entire facility, or major
portions, to the NRC staff for review.
Sec. 53.1206 Contents of Applications for Standard Design Approvals;
General Information
This proposed rule would add Sec. 53.1206, which would require
applications for a standard design approval under part 53 to contain
the information required by Sec. 53.1109(a) through (c) and (j).
Sec. 53.1209 Contents of Applications for Standard Design Approvals;
Technical Information
This proposed rule would add Sec. 53.1209, which would require the
inclusion of certain technical information, including a FSAR, site
parameters, and design information, when an applicant seeks review of
major portions of a standard design.
Sec. 53.1210 Contents of Applications for Standard Design Approvals;
Other Application Content
This proposed rule would add Sec. 53.1210, which would require
applications for standard design approvals under part 53 to include a
description of the availability controls used to satisfy the safety
criteria of Sec. 53.220, the program to protect Safeguards Information
against unauthorized disclosure, evidence that safety questions
associated with SSCs have been resolved, and a description of how
design features fulfill design criteria.
Sec. 53.1212 Standards for Review of Applications
This proposed rule would add Sec. 53.1212, which would require
applications for standard design approval to be reviewed under the
standards in parts 20, 53, and 73.
Sec. 53.1215 Referral to the Advisory Committee on Reactor Safeguards
This proposed rule would add Sec. 53.1215, which would require the
ACRS to report on any portions of the application for a standard design
approval under part 53 concerning safety.
Sec. 53.1218 Staff Approval of Design
This proposed rule would add Sec. 53.1218, which would require the
NRC staff to make a determination on the acceptability of the design,
publish its decision in the Federal Register, and issue a report
analyzing the design that is available at https://nrc.gov. Additionally,
the rule would establish the conditions under which a design approval
under part 53 remains valid.
Sec. 53.1221 Finality of Standard Design Approvals; Information
Requests
This proposed rule would add Sec. 53.1221, which would require NRC
staff and the ACRS to rely upon an approved design in their review of
any standard DC, ML, or individual facility license application under
part 53 that references the standard design approval. The proposed rule
would also govern requirements for issuing information requests.
Sec. 53.1230 Standard Design Certifications
This proposed rule would add Sec. 53.1230, which would provide an
overview of the requirements and procedures that govern the issuance of
standard DCs under part 53.
Sec. 53.1233 Filing of Applications
This proposed rule would add Sec. 53.1233, which would enable an
application for DC to be filed, regardless of whether an application
for a CP, COL, or ML has been filed, provided it complies with the
filing requirements in Sec. 53.040 and Sec. Sec. 2.811 through 2.819.
Sec. 53.1236 Contents of Applications for Standard Design
Certifications; General Information
This proposed rule would add Sec. 53.1236, which would require an
application for a standard DC under part 53 to contain all of the
information required by Sec. 53.1109(a) through (c) and (j).
Sec. 53.1239 Contents of Applications for Standard Design
Certifications; Technical Information
This proposed rule would add Sec. 53.1239, which would require
applicants for a standard DC under part 53 to submit a FSAR that
includes technical design information at a level of detail sufficient
to enable the Commission to make a safety determination.
Sec. 53.1241 Contents of Applications for Standard Design
Certifications; Other Application Content
This proposed rule would add Sec. 53.1241, which would require
applications for standard DCs under part 53 to include an environmental
report, as well as a description of the availability controls used to
satisfy the safety criteria of Sec. 53.220, proposed ITAAC, the
program to protect Safeguards Information against unauthorized
disclosure, evidence that safety questions associated with SSCs have
been resolved, and a description of how design features fulfill design
criteria.
Sec. 53.1242 Review of Applications
This proposed rule would add Sec. 53.1242, which would require
applications for standard DCs to be reviewed for compliance with the
standards in parts 20, 51, 53, and 73. It would also establish
procedural requirements for reviewing applications and holding hearings
in accordance with subpart H of part 2.
[[Page 87000]]
Sec. 53.1245 Referral to the Advisory Committee on Reactor Safeguards
This proposed rule would add Sec. 53.1245, which would require the
ACRS to report on any portions of the application for a standard DC
under part 53 concerning safety.
Sec. 53.1248 Issuance of Standard Design Certification
This proposed rule would add Sec. 53.1248, which would establish
the conditions under which the Commission may issue a DC rule that
specifies the site parameters, design characteristics, and any
additional terms and conditions of the DC rule.
Sec. 53.1251 Duration of Certification
This proposed rule would add Sec. 53.1251, which would set the
conditions under which a standard DC remains valid.
Sec. 53.1254 Application for Renewal
This proposed rule would add Sec. 53.1254, which would establish
the conditions and procedures for renewing a standard DC under part 53.
Sec. 53.1257 Criteria for Renewal
This proposed rule would add Sec. 53.1257, which would enable the
Commission to issue a rule granting the renewal of a standard DC under
part 53, impose additional requirements, and grant amendment requests.
Sec. 53.1260 Duration of Renewal
This proposed rule would add Sec. 53.1260, which would provide
that a renewal of a standard DC under part 53 is valid for not less
than 10 years, nor more than 15 years.
Sec. 53.1263 Finality of Standard Design Certifications
This proposed rule would add Sec. 53.1263, which would establish
limited conditions under which the Commission may initiate a rulemaking
to modify, rescind, or impose new requirements on a standard DC rule
under part 53. It would also address requests for an exemption from
elements of the certification information, and require that applicants
for a CP, COL, or ML that references a DC rule make information
normally contained in engineering documents available for audit.
Sec. 53.1270 Manufacturing Licenses
This proposed rule would add Sec. 53.1270, which would provide an
overview of the requirements and procedures for applying for and
issuing an ML under part 53.
Sec. 53.1273 Filing of Applications
This proposed rule would add Sec. 53.1273, which would establish
the requirements to apply for an ML under part 53.
Sec. 53.1276 Contents of Applications for Manufacturing Licenses;
General Information
This proposed rule would add Sec. 53.1276, which would require
applicants for an ML under part 53 to include the information contained
in Sec. 53.1109(a) through (e) and (j).
Sec. 53.1279 Contents of Applications for Manufacturing Licenses;
Technical Information
This proposed rule would add Sec. 53.1279, which would require an
applicant for an ML under part 53 to include certain technical
information in a FSAR, including but not limited to information about
site parameters, design information, manufacturing information, and
information related to the potential fueling and ultimate deployment of
a completed manufactured reactor.
Sec. 53.1282 Contents of Applications for Manufacturing Licenses;
Other Application Content
This proposed rule would add Sec. 53.1282, which would require
applicants for an ML under part 53 to include in their application the
proposed ITAAC, an environmental report, a description of the program
to protect Safeguards Information against unauthorized disclosure, and
a description of how design features fulfill design criteria. It would
also include content requirements for the ITAAC and environmental
reports in applications that reference a standard DC.
Sec. 53.1285 Review of Applications
This proposed rule would add Sec. 53.1285, which would require
applications for MLs under part 53 to be reviewed for compliance with
applicable standards and establish procedural requirements for
reviewing applicants and holding hearings in accordance with part 2.
Sec. 53.1286 Referral to the Advisory Committee on Reactor Safeguards
This proposed rule would add Sec. 53.1286, which would require the
ACRS to report on any portions of the application for an ML under part
53 concerning safety.
Sec. 53.1287 Issuance of Manufacturing Licenses
This proposed rule would add Sec. 53.1287, which would establish
the conditions under which the Commission may issue an ML under part
53.
Sec. 53.1288 Finality of Manufacturing Licenses
This proposed rule would add Sec. 53.1288, which would address the
limited circumstances in which the Commission may modify, rescind, or
impose new requirements following the issuance of an ML under part 53.
It would also address requests for a departure from the specifications
of the license.
Sec. 53.1291 Duration of Manufacturing Licenses
This proposed rule would add Sec. 53.1291, which would govern the
expiration of an ML, which is valid for no less than 5, nor more than
15 years from the date of issuance.
Sec. 53.1293 Transfer of Manufacturing Licenses
This proposed rule would add Sec. 53.1293, which would provide
that an ML under part 53 may be transferred in accordance with Sec.
53.1570.
Sec. 53.1295 Renewal of Manufacturing Licenses
This proposed rule would add Sec. 53.1295, which would establish
the procedures for applicants to apply for and the Commission to grant
a renewal of an ML under part 53.
Sec. 53.1300 Construction Permits
This proposed rule would add Sec. 53.1300, which would provide an
overview of the requirements and procedures for applicants to apply for
and the Commission to grant a CP under part 53.
Sec. 53.1306 Contents of Applications for Construction Permits;
General Information
This proposed rule would add Sec. 53.1306, which would require
applicants for a CP under part 53 to submit the general information
required by Sec. 53.1109, as well as financial information.
Sec. 53.1309 Contents of Applications for Construction Permits;
Technical Information
This proposed rule would add Sec. 53.1309, which would require
applicants for a CP under part 53 to submit a PSAR and a description of
the program to protect Safeguards
[[Page 87001]]
Information from unauthorized disclosure.
Sec. 53.1312 Contents of Applications for Construction Permits; Other
Application Content
This proposed rule would add Sec. 53.1312, which would require
applicants for a CP under part 53 to submit an environmental report and
to provide additional details in the PSAR if the application references
an ESP, standard design approval, or standard DC.
Sec. 53.1315 Review of Applications
This proposed rule would add Sec. 53.1315, which would require
applications for CPs under part 53 to be reviewed for compliance with
applicable standards and establish procedural requirements for
reviewing applications and holding hearings in accordance with part 2.
Sec. 53.1318 Finality of Referenced NRC Approvals, Permits, and
Certifications
This proposed rule would add Sec. 53.1318, which would address the
finality of ESPs, standard design approvals, and standard DCs
referenced in the CP application.
Sec. 53.1324 Referral to the Advisory Committee on Reactor Safeguards
This proposed rule would add Sec. 53.1324, which would require the
ACRS to report on any portions of the application for a CP under part
53 concerning safety.
Sec. 53.1327 Authorization To Conduct Limited Work Authorization
Activities
This proposed rule would add Sec. 53.1327, which would govern
authorization to conduct LWA activities.
Sec. 53.1330 Exemptions, Departures, and Variances
This proposed rule would add Sec. 53.1330, which would govern
requests for and issuance of exemptions from the Commission's
regulations and exemptions, departures, and variances from NRC
approvals, permits, and certifications.
Sec. 53.1333 Issuance of Construction Permits
This proposed rule would add Sec. 53.1333, which would establish
the conditions under which the Commission may issue CPs and
accompanying terms and conditions under part 53.
Sec. 53.1336 Finality of Construction Permits
This proposed rule would add Sec. 53.1336, which would address the
finality of CPs.
Sec. 53.1342 Duration of Construction Permits
This proposed rule would add Sec. 53.1342, which would establish
requirements for the expiration of a CP.
Sec. 53.1345 Transfer of Construction Permits
This proposed rule would add Sec. 53.1345, which would govern the
transfer of CPs under part 53.
Sec. 53.1348 Termination of Construction Permits
This proposed rule would add Sec. 53.1348, which would require the
holder of a permit under part 53 to provide written certification to
the Commission within 30 days of determining to permanently cease
construction.
Sec. 53.1360 Operating Licenses
This proposed rule would add Sec. 53.1360, which would provide an
overview of the requirements and procedures for applicants to apply for
and the Commission to issue an OL under part 53.
Sec. 53.1366 Contents of Applications for Operating Licenses; General
Information
This proposed rule would add Sec. 53.1366, which would require an
application for an OL under part 53 to include the information required
by Sec. 53.1109 as well as financial information.
Sec. 53.1369 Contents of Applications for Operating Licenses;
Technical Information
This proposed rule would add Sec. 53.1369, which would require an
application for an OL under part 53 to include certain technical
information in an FSAR at a level of detail sufficient for the
Commission to reach a final conclusion on all safety matters.
Sec. 53.1372 Contents of Applications for Operating Licenses; Other
Application Content
This proposed rule would add Sec. 53.1372, which would require an
application for an OL under part 53 to include an environmental report
and a description of availability controls.
Sec. 53.1375 Review of Applications
This proposed rule would add Sec. 53.1375, which would establish
the standards and procedures for reviewing applications and holding
hearings on OLs under part 53.
Sec. 53.1381 Referral to the Advisory Committee on Reactor Safeguards
This proposed rule would add Sec. 53.1381, which would require the
ACRS to report on any portions of the application for a CP under part
53 concerning safety.
Sec. 53.1384 Exemptions, Departures, and Variances
This proposed rule would add Sec. 53.1384, which would govern
requests for and the issuance of exemptions from the Commission's
regulations and exemptions, departures, and variances from NRC
approvals, permits, and certifications.
Sec. 53.1387 Issuance of Operating Licenses
This proposed rule would add Sec. 53.1387, which would establish
the conditions under which the Commission may issue OLs and
accompanying conditions and limitations, including TS, under part 53.
Sec. 53.1390 Backfitting of Operating Licenses
This proposed rule would add Sec. 53.1390, which would prevent the
Commission from modifying, adding, or deleting any terms or conditions
of the OL, except in accordance with Sec. 53.1590.
Sec. 53.1396 Duration of Operating Licenses
This proposed rule would add Sec. 53.1396, which would provide
that an OL under part 53 may be valid for up to 40 years.
Sec. 53.1399 Transfer of an Operating License
This proposed rule would add Sec. 53.1399, which would provide
that an OL under part 53 may be transferred under Sec. 53.1570.
Sec. 53.1402 Application for Renewal
This proposed rule would add Sec. 53.1402, which would provide
that an application for a renewed OL under part 53 must be filed in
accordance with Sec. 53.1595.
Sec. 53.1405 Continuation of an Operating License
This proposed rule would add Sec. 53.1405, which would govern the
continuing obligations of the holder of an OL under part 53 following
the permanent cessation of operations.
[[Page 87002]]
Sec. 53.1410 Combined Licenses
This proposed rule would add Sec. 53.1410, which would provide an
overview of the requirements and procedures for applicants to apply for
and the Commission to issue a COL under part 53.
Sec. 53.1413 Contents of Applications for Combined Licenses; General
Information
This proposed rule would add Sec. 53.1413, which would require an
application for a COL under part 53 to include the information required
by Sec. 53.1109 as well as financial information.
Sec. 53.1416 Contents of Applications for Combined Licenses; Technical
Information
This proposed rule would add Sec. 53.1416, which would require
applicants for a COL under part 53 to submit an FSAR with a level of
technical information sufficient to reach a final conclusion on all
safety matters.
Sec. 53.1419 Contents of Applications for Combined Licenses; Other
Application Content
This proposed rule would add Sec. 53.1419, which would require
applicants for a COL under part 53 to submit an environmental report, a
description of availability controls, the ITAAC that the licensee must
perform. It would also include ITAAC requirements for applications that
reference an ESP, standard DC, ML, or combination thereof.
Sec. 53.1422 Review of Applications
This proposed rule would add Sec. 53.1422, which would require
applications for COLs under part 53 to be reviewed for compliance with
applicable standards and establish procedural requirements for
reviewing applications and holding hearings in accordance with part 2.
Sec. 53.1425 Finality of Referenced NRC Approvals
This proposed rule would add Sec. 53.1425 which would address the
finality of ESPs, standard DC rules, standard design approvals, or MLs
referenced in the application for a COL under part 53.
Sec. 53.1431 Referral to the Advisory Committee on Reactor Safeguards
This proposed rule would add Sec. 53.1431, which would require the
ACRS to report on any portions of the application for a COL under part
53 concerning safety.
Sec. 53.1434 Authorization To Conduct Limited Work Authorization
Activities
This proposed rule would add Sec. 53.1434, which would address
authorization to conduct LWA activities.
Sec. 53.1437 Exemptions, Departures, and Variances
This proposed rule would add Sec. 53.1437, which would govern the
conditions in which the Commission may grant an exemption for one or
more of its regulations, or an exemption, variance, or departure from a
permit, design approval, or license.
Sec. 53.1440 Issuance of Combined Licenses
This proposed rule would add Sec. 53.1440, which would establish
the conditions under which the Commission may issue COLs and
accompanying conditions and limitations, including TS, under part 53.
Sec. 53.1443 Finality of Combined Licenses
This proposed rule would add Sec. 53.1443, which would govern
permissible modifications or amendments that the Commission may make to
a COL, as well as permissible changes that a licensee may make to
facilities and procedures as described in the FSAR.
Sec. 53.1449 Inspection During Construction
This proposed rule would add Sec. 53.1449, which would establish
requirements related to inspections, tests, or analyses for the holder
of a COL under part 53.
Sec. 53.1452 Operation Under a Combined License
This proposed rule would add Sec. 53.1452, which would establish
requirements describing the notifications, hearings, and findings to be
made prior to commencing facility operations.
Sec. 53.1455 Duration of a Combined License
This proposed rule would add Sec. 53.1455, which would govern the
duration of a COL under part 53.
Sec. 53.1456 Transfer of a Combined License
This proposed rule would add Sec. 53.1456, which would permit the
transfer of a COL under part 53 in accordance with Sec. 53.1570.
Sec. 53.1458 Application for Renewal
This proposed rule would add Sec. 53.1458, which would provide
that an application for renewal of a COL must be filed in accordance
with Sec. 53.1595.
Sec. 53.1461 Continuation of Combined License
This proposed rule would add Sec. 53.1461, which would govern the
continuing obligations of the holder of a COL under part 53 following
the permanent cessation of operations.
Sec. 53.1470 Standardization of Commercial Nuclear Plant Designs:
Licenses To Construct and Operate Nuclear Power Reactors of Identical
Design at Multiple Sites
This proposed rule would add Sec. 53.1470, which would govern the
requirements and procedures for filing and issuing applications for a
CP, OL, or COL under part 53 in which the applicant seeks approval of
the same design for multiple sites.
Subpart I--Maintaining and Revising Licensing-Basis Information
This proposed rule would add subpart I, which would address the
maintenance of licensing-basis information for part 53.
Sec. 53.1500 Licensing-Basis Information
This proposed rule would add Sec. 53.1500, describing the purpose
of subpart I, which would be to provide the requirements for the
maintenance of licensing-basis information for commercial nuclear
plants licensed under part 53.
Sec. 53.1502 Specific Terms and Conditions of Licenses
This proposed rule would add Sec. 53.1502, which would outline the
specific terms and conditions for obtaining a license under part 53.
Sec. 53.1505 Changes to Licensing-Basis Information Requiring Prior
NRC Approval
This proposed rule would add Sec. 53.1505, which would provide an
overview of the process for licensees to request, and the Commission to
issue, amendments to licensing-basis information under part 53.
Sec. 53.1510 Application for Amendment of License
This proposed rule would add Sec. 53.1510, which would require
licensees under part 53 to file an application to request an amendment
to the license. Applicants must assess how their requested changes
would impact the safety criteria and analysis
[[Page 87003]]
requirements in subpart B and C, as applicable, whether the amendment
involves no significant hazards consideration using the standards in
Sec. 53.1520 and consider potential impacts on environmental factors.
Sec. 53.1515 Public Notices; State Consultation
This proposed rule would add Sec. 53.1515, which would outline the
Commission's procedures for issuing a notification in the Federal
Register and consulting with the State in which the commercial nuclear
facility is located in connection with its consideration of
applications for an amendment to an OL or COL under part 53.
Sec. 53.1520 Issuance of Amendment
This proposed rule would add Sec. 53.1520, which would outline
criteria for the Commission to consider in issuing license amendments
under part 53.
Sec. 53.1525 Revising Certification Information Within a Design
Certification Rule
This proposed rule would add Sec. 53.1525, which would address the
requirements for applicants to request, and the Commission to grant, an
exemption to a DC rule under part 53.
Sec. 53.1530 Revising Design Information Within a Manufacturing
License
This proposed rule would add Sec. 53.1530, which would require the
holder of an ML to request an amendment under Sec. 53.1510 and, as
applicable, Sec. 53.1520 to make changes to the design of a
manufactured reactor. It would also outline the requirements for
holders of a COL under part 53 to request amendments for changes to the
design information of a manufactured reactor.
Sec. 53.1535 Amendments During Construction
This proposed rule would add Sec. 53.1535, which would outline the
process for licensees under part 53 to request amendments to CPs or
LWAs during construction.
Sec. 53.1540 Updating Licensing-Basis Information and Determining the
Need for NRC Approval
This proposed rule would add Sec. 53.1540, which would provide an
overview of the regulations in subpart I for holders of an OL or COL
under part 53 to modify licensing-basis information and definitions
relevant to Sec. Sec. 53.1545 through 53.1565.
Sec. 53.1545 Updating Final Safety Analysis Reports
This proposed rule would add Sec. 53.1545, which would require
licensees under part 53 to regularly update FSARs in accordance with
the requirements of this section to reflect changes to licensing-basis
information.
Sec. 53.1550 Evaluating Changes to Facility as Described in Final
Safety Analysis Reports
This proposed rule would add Sec. 53.1550, which would require
licensees under part 53 to follow the guidelines outlined in this
section in determining whether changes to licensing-basis information
described in the FSAR (as updated) require them to obtain a license
amendment.
Sec. 53.1560 Updating Program Documents Included in Licensing-Basis
Information
This proposed rule would add Sec. 53.1560, which would require the
holders of an OL or COL under part 53 to regularly update the program
documents that they submitted in their application for a license.
Sec. 53.1565 Evaluating Changes to Programs Included in Licensing-
Basis Information
This proposed rule would add Sec. 53.1565, which would enable
licensees under part 53 to make changes to the facility, procedures, or
organization, or address changes to site environs as described in
program documents without NRC approval if these changes satisfy the
criteria outlined in this section.
Sec. 53.1570 Transfer of Licenses
This proposed rule would add Sec. 53.1570, which would outline the
requirements for an application for transfer of a license issued under
part 53.
Sec. 53.1575 Termination of Licenses
This proposed rule would add Sec. 53.1575, which would outline the
process for terminating an OL or COL issued under part 53.
Sec. 53.1580 Information Requests
This proposed rule would add Sec. 53.1580, which would address the
process and circumstances under which the NRC may send information
requests to the various types of licensees within part 53.
Sec. 53.1585 Revocation, Suspension, Modification of Licenses and
Approvals for Cause
This proposed rule would add Sec. 53.1585, which would address
grounds for the revocation, suspension, or modification of a license or
standard design approval issued under part 53.
Sec. 53.1590 Backfitting
This proposed rule would add Sec. 53.1590, which would define
backfitting and establish requirements to be met by the NRC when it
takes backfitting actions under part 53.
Sec. 53.1595 Renewal
This proposed rule would add Sec. 53.1595, which would provide for
the renewal of a license under part 53 upon expiration.
Subpart J--Reporting and Other Administrative Requirements
This proposed rule would add subpart J, to establish various
reporting and other administrative requirements for licensees under
part 53.
Sec. 53.1600 General Information
This proposed rule would add Sec. 53.1600, which provides an
overview of the sections that would require applicants and licensees
under part 53 to provide NRC inspectors with unfettered access to sites
and facilities, maintain records and make reports, demonstrate
compliance with financial qualification and reporting requirements, and
maintain required financial protection for accidents.
Sec. 53.1610 Unfettered Access for Inspections
This proposed rule would add Sec. 53.1610, which would require
applicants and licensees under part 53 to provide unfettered access to
NRC inspectors, including access to records, premises, activities, and
licensed materials, in addition to providing office space to
accommodate temporary or resident inspectors.
Sec. 53.1620 Maintenance of Records, Making of Reports
This proposed rule would add Sec. 53.1620, which would require
part 53 licensees to maintain all records and make reports as required
by the conditions of the license or by the regulations in part 53.
Sec. 53.1630 Immediate Notification Requirements for Operating
Commercial Nuclear Plants
This proposed rule would add Sec. 53.1630, which would impose
immediate notification requirements on
[[Page 87004]]
part 53 licensees following the declaration of an Emergency Class or
the discovery of certain non-emergency events.
Sec. 53.1640 Licensee Event Report System
This proposed rule would add Sec. 53.1640, which would require any
commercial plant licensee holding an OL under part 53 to submit a
Licensee Event Report in accordance with the specifications outlined in
this section.
Sec. 53.1645 Reports of Radiation Exposure to Members of the Public
The proposed rule would add Sec. 53.1645, which would require
annual reports to the Commission, including radiological reports as
required by part 20, an Annual Radioactive Effluent Release Report, and
an Annual Environmental Operating Report.
Sec. 53.1650 Facility Information and Verification
The proposed rule would add Sec. 53.1650, which would include a
reporting requirement for applicants and holders of a CP or license
under part 53 to support safeguards agreements between the United
States and the IAEA.
Sec. 53.1660 Financial Requirements
This proposed rule would add Sec. 53.1660, which would introduce
requirements and procedures related to financial qualifications and
reporting requirements for applicants, licensees, and CP holders under
part 53.
Sec. 53.1670 Financial Qualifications
This proposed rule would add Sec. 53.1670, which would require an
applicant for a CP, OL, or COL under part 53 to must demonstrate
possession or ability to obtain funds necessary for the activities for
which the permit or license is sought.
Sec. 53.1680 Annual Financial Reports
This proposed rule would add Sec. 53.1680, which would require
licensees and holders of a CP under part 53 to submit annual financial
reports to the Commission, with exceptions for those that submit
financial forms to the Securities and Exchange Commission or the
Federal Energy Regulatory Commission.
Sec. 53.1690 Licensee's Change of Status; Financial Qualifications
This proposed rule would add Sec. 53.1690, which would require
electric utility licensees that hold an OL or COL for a commercial
nuclear plant under part 53 to provide the NRC with the financial
qualifications information outlined in this section within seventy-five
days of ceasing to be an electric utility.
Sec. 53.1700 Creditor Regulations
This proposed rule would add Sec. 53.1700, which would establish
regulations with respect to the creditors of any facility under part
53.
Sec. 53.1710 Financial Protection
This proposed rule would add Sec. 53.1710, which would establish
requirements for licenses under part 53 to obtain and maintain
insurance to cover the costs of an accident.
Sec. 53.1720 Insurance Required To Stabilize and Decontaminate Plant
Following an Accident
This proposed rule would add Sec. 53.1720, which would require
commercial nuclear plant licensees under part 53 to obtain insurance
sufficient to cover the costs of stabilizing and decontaminating the
plant in the event of an accident.
Sec. 53.1730 Financial Protection Requirements
This proposed rule would add Sec. 53.1730, which would require
commercial nuclear plant licensees under part 53 to satisfy the
provisions of part 140.
Subpart M--Enforcement
This proposed rule would add subpart M, which would address certain
violations and penalties associated with violations of part 53
regulations.
Sec. 53.9000 Violations
This proposed rule would add Sec. 53.9000, providing notice of the
Commission's authority to obtain injunctions or other court orders for
the violations enumerated in this section.
Sec. 53.9010 Criminal Penalties
This proposed rule would add Sec. 53.9010, providing notice to all
persons and entities subject to part 53 that they are subject to
criminal sanctions for willful violations, attempted violations, or
conspiracy to violate certain regulations under part 53.
Sec. 70.20a General License to Possess Special Nuclear Material for
Transport
This proposed rule would revise Sec. 70.20a(b) to include a
reference to part 53.
Sec. 70.22 Contents of Applications
This proposed rule would revise Sec. 70.22, paragraphs (b),
(h)(1), (j)(1), and (k) to include the appropriate references to part
53.
Sec. 70.24 Criticality Accident Requirements
This proposed rule would revise Sec. 70.24(d) to include the
appropriate references to part 53.
Sec. 70.32 Conditions of Licenses
This proposed rule would revise Sec. 70.32(c)(1) and (d) to
incorporate the appropriate references to part 53.
Sec. 70.50 Reporting Requirements
This proposed rule would revise Sec. 70.50(d) to clarify the
applicability of the reporting requirements of this section to part 53
licensees.
Sec. 72.3 Definitions
This proposed rule would revise the definition of ``Independent
spent fuel storage installation or ISFSI'' in Sec. 72.3 to include a
reference to facilities licensed under part 53.
Sec. 72.30 Financial Assurance and Recordkeeping for Decommissioning
This proposed rule would revise Sec. 72.30(e)(5) to include the
appropriate references to part 53.
Sec. 72.32 Emergency Plan
This proposed rule would revise Sec. 72.32(c)(2) to include a
reference to the exclusion area as defined in part 53.
Sec. 72.40 Issuance of License
This proposed rule would revise Sec. 72.40(c) regarding the
issuance of a license under part 72 to include a reference to previous
licensing actions, including the issuance of a CP under part 53.
Sec. 72.75 Reporting Requirements for Specific Events and Conditions
This proposed rule would revise Sec. 72.75(i)(1)(ii) regarding
reporting requirements for specific events and conditions with
references to reactors licensed under part 53.
Sec. 72.184 Safeguards Contingency Plan
This proposed rule would revise Sec. 72.184(a) regarding the
requirements of a licensee's safeguarding contingency plan with a
reference to nuclear facilities licensed under part 53.
Sec. 72.210 General License Issued
This proposed rule would revise Sec. 72.210 to issue a general
license for the storage of spent fuel in an independent spent storage
installation at power to persons authorized to possess or operate
nuclear power reactors under part 53.
[[Page 87005]]
Sec. 72.212 Conditions of General License Issued Under Sec. 72.210
This proposed rule would revise Sec. 72.212(b)(8) regarding the
conditions of a general license issued under Sec. 72.210 to include a
reference to license amendments for a facility made pursuant to part
53.
Sec. 72.218 Termination of Licenses
This proposed rule would revise Sec. 72.218(a) to include a
reference to the notification required under part 53 regarding the plan
for managing spent fuel prior to decommissioning. It would also extend
the provisions of Sec. 72.218(b) to a reactor operating or COL under
part 53.
Sec. 73.1 Purpose and Scope
This proposed rule would revise Sec. 73.1(b)(1)(i) to extend the
scope of part 73 to production and utilization facilities licensed
under part 53, in addition to parts 50 and 52.
Sec. 73.2 Definitions
This proposed rule would revise Sec. 73.2 introductory text and
paragraph (a) such that terms defined in part 53 have the same meaning
in part 73.
Sec. 73.8 Information Collection Requirements: OMB Approval
This proposed rule would revise Sec. 73.8(b) with the new
information collection requirements contained in proposed Sec. Sec.
73.77, 73.100, 73.110, and 73.120.
Sec. 73.50 Requirements for Physical Protection of Licensed Activities
This proposed rule would revise Sec. 73.50 to exempt nuclear
reactor facilities licensed under part 53, in addition to parts 50 and
52, from the requirements of this section.
Sec. 73.55 Requirements for Physical Protection of Licensed Activities
in Nuclear Power Reactors Against Radiological Sabotage
This proposed rule would revise Sec. 73.55, paragraphs (a)(4) and
(6), (i)(4)(iii), (l)(1), (l)(7)(ii), (p)(1)(i), (r)(2), and
(r)(4)(iii), to incorporate the appropriate references to part 53
regarding requirements for physical protection of licensed activities
in nuclear power reactors against radiological sabotage.
Sec. 73.56 Personnel Access Authorization Requirements for Nuclear
Power Plants
This proposed rule would revise Sec. 73.56(a)(3) to apply this
section's personnel AA requirements to applicants for an OL or holders
of a COL under part 53 who do not demonstrate compliance with certain
requirements under part 53.
Sec. 73.57 Requirements for Criminal History Records Checks of
Individuals Granted Unescorted Access to a Nuclear Power Facility, a
Non-power Reactor, or Access to Safeguards Information
This proposed rule would revise Sec. 73.57(a)(3) to incorporate
the appropriate references to OLs granted under part 53 and Commission
findings under Sec. 53.1452(g) regarding the requirement for license
applicants to submit fingerprints for all personnel with unescorted
access.
Sec. 73.58 Safety/Security Interface Requirements for Nuclear Power
Reactors
This proposed rule would revise Sec. 73.58(a) to extend the
requirements of this section to part 53 licensees.
Sec. 73.67 Licensee Fixed Site and In-Transit Requirements for the
Physical Protection of Special Nuclear Material of Moderate and Low
Strategic Significance
This proposed rule would revise Sec. 73.67(d) and (f) to include a
reference to licensees authorized to operate a nuclear power plant
under part 53.
Sec. 73.77 Cybersecurity Event Notifications
This proposed rule would revise Sec. 73.77, paragraphs (a), (b),
(c)(6) and (7) regarding the notification process for cybersecurity
events to include notifications for the declaration of an emergency
class made under part 53.
Subpart J--Security Requirements at Commercial Nuclear Plants
This proposed rule would add new Subpart J of part 73 containing
Sec. Sec. 73.100, 73.110, and 73.120, to establish security
requirements for commercial nuclear plants licensed under part 53.
Sec. 73.100 Technology-Inclusive Requirements for Physical Protection
of Licensed Activities at Commercial Nuclear Plants Against
Radiological Sabotage
This proposed rule would add Sec. 73.100, which would establish a
performance-based regulatory framework for physical protection as an
alternative to the prescriptive requirements of Sec. 73.55, which also
governs physical protection programs for part 50 and 52 licensees.
Sec. 73.110 Technology-Inclusive Requirements for Protection of
Digital Computer and Communication Systems and Networks
This proposed rule would add Sec. 73.110, which would establish a
consequence-based approach to cybersecurity and would require that part
53 licensees demonstrate reasonable assurance that digital computer and
communication systems and networks are adequately protected against
cyberattacks in a manner that is commensurate with the potential
consequences of those attacks.
Sec. 73.120 Access Authorization Program for Commercial Nuclear Plants
This proposed rule would add Sec. 73.120, which would establish
performance objectives as an alternative to compliance with the AA
provisions of Sec. Sec. 73.55, 73.56, and 73.57. This proposed rule
would afford part 53 licensees additional flexibility in establishing
an AA program that demonstrates compliance with the performance
objectives and requirements of this section.
Sec. 73.1200 Notification of Physical Security Events
This proposed rule would revise Sec. 73.1200, paragraphs (a),
(c)(1), (e)(1), (e)(3), (e)(4), (g)(1), (o)(5)(i), (o)(6)(i), (r), and
(s) to extend the requirements of this section to part 53 licensees.
Sec. 73.1205 Written Follow-Up Reports of Physical Security Events
This proposed rule would revise Sec. 73.1205(b)(2) to extend the
requirements of this section to part 53 licensees.
Sec. 73.1210 Recordkeeping of Physical Security Events
This proposed rule would revise Sec. 73.1210(a)(1) and (b)(3)(i)
to extend the requirements of this section to part 53 licensees.
Sec. 73.1215 Suspicious Activity Reports
This proposed rule would revise Sec. 73.1215(d)(1) to include a
reference to Sec. 73.100.
Appendix B to part 73--General Criteria for Security Personnel
This proposed rule would revise appendix B to part 73 to state that
terms defined in part 53 have the same meaning when used in this
appendix.
Sec. 74.31 Nuclear Material Control and Accounting for Special Nuclear
Material of Low Strategic Significance
This proposed rule would revise Sec. 74.31(a) to include a
reference to
[[Page 87006]]
production or utilization facilities licensed under part 53, in
addition to parts 50 and 70.
Sec. 74.41 Nuclear Material Control and Accounting for Special Nuclear
Material of Moderate Strategic Significance
This proposed rule would revise Sec. 74.41(a) to include a
reference to nuclear reactors licensed under part 53.
Sec. 74.51 Nuclear Material Control and Accounting for Strategic
Special Nuclear Material
This proposed rule would revise Sec. 74.51(a) to include a
reference to nuclear reactors licensed under part 53.
Sec. 75.4 Definitions
This proposed rule would revise Sec. 75.4 such that terms defined
in Sec. 53.020 have the same meaning when used in this part. The
definition of ``Facility'' would also be revised to include any plant
or location where more than 1 effective kilogram of nuclear material is
licensed pursuant to part 53.
Sec. 95.5 Definitions
This proposed rule would revise the definition of ``License'' in
Sec. 95.5 to include those issued under part 53.
Sec. 95.39 External Transmission of Documents and Material
This proposed rule would revise Sec. 95.39(a) to apply
restrictions to the external transmission of documents and material
containing classified information in connection with NRC licenses,
certificates, standard design approvals, or standard DCs issued under
part 53.
Sec. 140.2 Scope
This proposed rule would revise Sec. 140.2(a)(1) and (2) to
include part 53 applicants and licensees within the scope of part 140
regulations.
Sec. 140.10 Scope
This proposed rule would revise Sec. 140.10 to apply the
provisions of subpart B to applicants or holders of a license to
operate a nuclear reactor under part 53, as well as applicants and
holders of a COL under part 53.
Sec. 140.11 Amounts of Financial Protection for Certain Reactors
This proposed rule would revise Sec. 140.11(b) to require the
licensee's primary financial protection to cover all reactors in any
case where a person is authorized under part 53 to operate two or more
nuclear reactors at the same location.
Sec. 140.12 Amount of Financial Protection Required for Other Reactors
This proposed rule would revise Sec. 140.12(c) to require the
licensee's primary financial protection to cover all reactors in any
case where a person is authorized under part 53 to operate two or more
nuclear reactors at the same location.
Sec. 140.13 Amount of Financial Protection Required of Certain Holders
of Construction Permits and Combined Licenses Under 10 CFR Part 52
This proposed rule would revise Sec. 140.13 with the appropriate
references to part 53 regarding the requirement for holders of a CP or
COL under part 53 to obtain financial protection.
Sec. 140.20 Indemnity Agreements and Liens
This proposed rule would revise Sec. 140.20(a)(1)(i) and (ii) with
appropriate references to part 53.
Sec. 150.15 Persons Not Exempt
The proposed rule would revise Sec. 150.15, paragraphs (a)(7)(iii)
and (a)(8) to add a reference to facilities licensed under parts 53 and
52.
Sec. 170.3 Definitions
The proposed rule would revise Sec. 170.3 to incorporate
references to part 53 into the definitions of ``Manufacturing
license,'' ``Part 55 Reviews,'' ``Power reactor,'' and ``Special
projects.''
Sec. 170.12 Payment of Fees
The proposed rule would revise Sec. 170.12(d)(1)(v) regarding
special project fees in connection with FSARs to include part 53.
Sec. 170.21 Schedule of Fees for Production and Utilization
Facilities, Review of Standard Referenced Design Approvals, Special
Projects, Inspections, And import and Export Licenses
The proposed rule would revise Sec. 170.21, footnote 1 to include
fees charged for approvals issued under the exemption provision in
Sec. 53.080.
Sec. 170.41 Failure by Applicant or Licensee to Pay Prescribed Fees
The proposed rule would revise Sec. 170.41 to include a general
reference to part 53 in connection with remedial actions that the
Commission might take when an applicant or licensee fails to pay a
prescribed fee required by this part.
Sec. 171.3 Scope
The proposed rule would revise Sec. 171.3 to apply the provisions
of this part to any person holding an OL for a power reactor licensed
under part 53 or a COL issued under part 53.
Sec. 171.5 Definitions
This proposed rule would revise the definitions of ``Operating
license'' and ``Power reactor'' in Sec. 171.5 to incorporate the
appropriate references to part 53.
Sec. 171.15 Annual fees: Non-Power Production or Utilization Licenses,
Reactor Licenses, and Independent Spent Fuel Storage Licenses
This proposed rule would revise Sec. 171.15, paragraphs (a),
(b)(2)(iii), (c)(1), and (d)(1) regarding annual fees that are
applicable to part 53 licensees.
Sec. 171.17 Proration
This proposed rule would revise Sec. 171.17, paragraphs (a),
(a)(1)(ii) and (a)(2) with references to part 53 licenses.
VIII. Regulatory Flexibility Certification
The Regulatory Flexibility Act of 1980, as amended at 5 U.S.C. 601
et seq, requires that agencies consider the impact of their rulemakings
on small entities and, consistent with applicable statutes, consider
alternatives to minimize these impacts on the businesses,
organizations, and government jurisdictions to which they apply.
In accordance with the Small Business Administration's (SBA's)
regulation at 13 CFR 121.903(c), the NRC has developed its own size
standards for performing an RFA analysis and has verified with the SBA
Office of Advocacy that its size standards are appropriate for NRC
analyses. The NRC size standards at Sec. 2.810, ``NRC size
standards,'' are used to determine whether an applicant or licensee
qualifies as a small entity in the NRC's regulatory programs. Section
2.810 defines the following types of small entities:
Small business is a for-profit concern and is a--(1) Concern that
provides a service or a concern not engaged in manufacturing with
average gross receipts of $8.0 million or less over its last 5
completed fiscal years; or (2) Manufacturing concern with an average
number of 500 or fewer employees based upon employment during each pay
period for the preceding 12 calendar months.
Small organization is a not-for-profit organization which is
independently
[[Page 87007]]
owned and operated and has annual gross receipts of $8.0 million or
less.
Small governmental jurisdiction is a government of a city, county,
town, township, village, school district, or special district with a
population of less than 50,000.
Small educational institution is one that is--(1) Supported by a
qualifying small governmental jurisdiction; or (2) Not State or
publicly supported and has 500 or fewer employees.
Number of Small Entities Affected
The NRC is currently not aware of any known small entities as
defined in Sec. 2.810 that are planning to apply for a commercial
nuclear plant ESP, CP, OL, ML, or COL under part 53 that would be
impacted by this proposed rule. Based on this finding, the NRC has
preliminarily determined that the proposed rule would not have a
significant economic impact on a substantial number of small entities.
Economic Impact on Small Entities
Depending on how the ownership and/or operating responsibilities
for such an enterprise were structured, applicants for a commercial
nuclear plant rated 8 Megawatts electric (MWe) or less could
conceivably qualify as small entities as defined by Sec. 2.810. Owners
that operate power reactors rated greater than 8 MWe could generate
sufficient electricity revenue that exceeds the gross annual receipts
limit of $8 million, assuming a 90 percent capacity factor and the June
2021 DOE's Energy Information Administration U.S. average price of
electricity to the ultimate customer for all sectors of 11.3 cents per
kilowatt-hour.
Although the NRC is not aware of any small entities that would be
affected by the proposed rule, there is a possibility that future
applications for a commercial nuclear plant permit or license could be
submitted by small entities who plan to own and operate a commercial
nuclear plant rated 8 MWe or less. Commercial nuclear plants that are
rated 8 MWe or less would most likely be used to support electrical
demand for military bases or small remote towns and would provide
process heat, so they would not directly compete with a larger
commercial nuclear plant that would typically produce electricity for
the grid. As a result of these differing purposes, the NRC would expect
that small and large entities would not be in direct competition with
each other.
Therefore, the NRC preliminarily concludes that this proposed rule
would not have a significant economic impact on a substantial number of
small entities.
Request for Comments
The NRC is seeking comment on both its initial RFA analysis and on
its preliminary conclusion that this proposed rule would not have a
significant economic impact on a substantial number of small entities
because of the likelihood that most expected applicants would not
qualify as a small entity. Additionally, the NRC is seeking comment on
its preliminary conclusion that if a small entity were to submit a
commercial nuclear plant application, the small entity would not incur
a significant economic impact as it would most likely not be in
competition with a large entity.
Any small entity that could be subject to this regulation that
determines, because of its size, it is likely to bear a
disproportionate adverse economic impact should notify the Commission
of this opinion in a comment that indicates--
1. The applicant's size and how the proposed regulation would
impose a significant economic burden on the applicant as compared to
the economic burden on a larger applicant;
2. How the proposed regulations could be modified to take into
account the applicant's differing needs or capabilities;
3. The benefits that would accrue or the detriments that would be
avoided if the proposed regulations were modified as suggested by the
applicant;
4. How the proposed regulation, as modified, would more closely
equalize the impact of NRC regulations or create more equal access to
the benefits of Federal programs as opposed to providing special
advantages to any individual or group; and
5. How the proposed regulation, as modified, would still adequately
demonstrate compliance with the NRC's obligations under the Act.
IX. Regulatory Analysis
The NRC has prepared a draft regulatory analysis for this proposed
rule. The analysis examines the costs and benefits of the alternatives
considered by the NRC. The conclusion from the analysis is that this
proposed rule and associated guidance would result in net averted costs
to the industry and the NRC of $28.1 million using a 7-percent discount
rate and $34.5 million using a 3-percent discount rate due to
reductions in exemption requests. The analysis also assumes one
applicant under part 53. As the number of applicants increases, so do
the estimated averted costs. The NRC requests public comment on the
draft regulatory analysis, which is available as indicated in the
``Availability of Documents'' section of this document. Comments on the
draft regulatory analysis may be submitted to the NRC as indicated
under the ADDRESSES caption of this document.
X. Backfitting and Issue Finality
This section describes the backfitting and issue finality
implications of this proposed rule and the draft guidance documents
described in section XVIII, ``Availability of Guidance,'' in this
document, as applied to pertinent NRC approvals and certain applicants
that reference NRC approvals in their applications. The NRC's current
backfitting provisions associated with nuclear power plants appear in
Sec. 50.109, ``Backfitting,'' and apply to CPs and OLs under part 50.
Issue finality provisions (analogous to the backfitting provisions in
Sec. 50.109) for approvals under part 52 are located in various
provisions of part 52. The NRC Management Directive 8.4, ``Management
of Backfitting, Forward Fitting, Issue Finality, and Information
Requests,'' describes the Commission's policies on backfitting and
issue finality.
This proposed rule would provide a regulatory scheme for entities
to apply for approvals under part 53. The part 50 backfitting
provisions and part 52 issue finality provisions apply to actions taken
by the NRC under part 50 or part 52, respectively, or actions taken by
the NRC under other parts of 10 CFR chapter I that, for holders of
certain approvals under part 50 or part 52, inextricably affect their
activities regulated under part 50 or part 52. Issuance and
implementation of proposed part 53 would not constitute actions taken
under part 50 or part 52. Also, proposed part 53 would not allow an
applicant to reference approvals issued under part 50 or part 52.
Therefore, the issuance and implementation of proposed part 53 would
not affect part 50 or part 52 entities' activities regulated under part
50 or part 52. Therefore, the addition of part 53 through this proposed
rule would not be within the scope of the part 50 backfitting and part
52 issue finality provisions.
The NRC also proposes conforming changes to parts 1, 2, 10, 11, 19,
20, 21, 25, 26, 30, 40, 50, 51, 70, 72, 73, 74, 75, 95, 140, 150, 170,
and 171 to reflect the addition of part 53. These changes would not
meet the definition of ``backfitting'' in Sec. 50.109 or Sec. 70.76,
``Backfitting,'' because the proposed changes would not modify or add
to the systems, structures, components, or
[[Page 87008]]
design of a facility or to the procedures or organization required to
operate a facility under part 50 or 70. These changes would not meet
the definition of ``backfitting'' in Sec. 72.62, ``Backfitting,''
because the proposed changes would not add, eliminate, or modify the
SSCs of an independent spent fuel storage installation (ISFSI) or the
procedures or organization required to operate an ISFSI. These proposed
changes would not inextricably affect activities regulated under parts
50, 52, 70, or 72. Therefore, the proposed changes to parts 1, 2, 10,
11, 19, 20, 21, 25, 26, 30, 40, 50, 51, 70, 72, 73, 74, 75, 95, 140,
150, 170, and 171 would not constitute backfitting under parts 50, 70,
or 72 or affect the issue finality of an approval under part 52.
The NRC is issuing 10 draft guidance documents that, if issued as
final guidance documents, would provide guidance on the methods
acceptable to the NRC for complying with aspects of this proposed rule.
These documents would not apply to holders of approvals issued under
part 50 or part 52. Further, as discussed in the guidance documents,
applicants and licensees would not be required to comply with the
positions set forth in the guidance. Therefore, issuance of the
guidance documents as final guidance would not constitute backfitting
under part 50 or affect the issue finality of any approval issued under
part 52.
XI. Cumulative Effects of Regulation
The NRC seeks to minimize any potential negative consequences
resulting from the cumulative effects of regulation (CER). The CER
describes the challenges that licensees, or other impacted entities
such as State partners, may face while implementing new regulatory
positions, programs, or requirements (e.g., rules, generic letters,
backfits, inspections). The CER is an organizational effectiveness
challenge that may result from a licensee or impacted entity
implementing a number of complex regulatory actions, programs, or
requirements within limited available resources. The NRC's CER process
involved engaging with external stakeholders throughout this proposed
rule and related regulatory activities. Public involvement has included
numerous public meetings to examine the part 53 risk-informed,
technology-inclusive requirements for commercial nuclear plants and the
publication of numerous versions of preliminary proposed rule language.
The NRC is considering holding additional public meetings during the
remainder of the rulemaking process.
In parallel with this proposed rule, the NRC is issuing 10 draft
implementing guidance documents for comment to support informed
external stakeholder feedback. Section XVII, ``Availability of
Guidance,'' of this document describes how the public can access the
draft implementing guidance.
In addition to the questions in the ``Specific Requests for
Comments'' section of this document, the NRC is requesting CER feedback
on the following questions:
1. In light of any current or projected CER challenges, does the
proposed rule's effective date provide sufficient time to implement the
new proposed requirements, including changes to programs, procedures,
and the facility?
2. If CER challenges currently exist or are expected, what should
be done to address them? For example, if more time is required for
implementation of the new requirements, what period of time is
sufficient?
3. Do other (NRC or other agency) regulatory actions (e.g., orders,
generic communications, license amendment requests, inspection findings
of a generic nature) influence the implementation of the proposed
rule's requirements?
4. Are there unintended consequences? Does the proposed rule create
conditions that would be contrary to the proposed rule's purpose and
objectives? If so, what are the unintended consequences, and how should
they be addressed?
5. Please comment on the NRC's cost and benefit estimates in the
regulatory analysis that supports this proposed rule. The draft
regulatory analysis is available as indicated under the ``Availability
of Documents'' section of this document.
XII. Plain Writing
The Plain Writing Act of 2010 (Pub. L. 111-274) requires Federal
agencies to write documents in a clear, concise, and well-organized
manner. The NRC has written this document to be consistent with the
Plain Writing Act as well as the Presidential Memorandum, ``Plain
Language in Government Writing,'' published June 10, 1998 (63 FR
31885). The NRC requests comment on this document with respect to the
clarity and effectiveness of the language used.
XIII. Environmental Assessment and Proposed Finding of No Significant
Environmental Impact
The Commission has preliminarily determined under the National
Environmental Policy Act of 1969, as amended, and the Commission's
regulations in subpart A of part 51, that this rule, if adopted, would
not be a major Federal action significantly affecting the quality of
the human environment, and an EIS is not required. The implementation
of the proposed rule requirements does not have a significant impact on
the environment. The proposed rulemaking would either have requirements
that are administrative in application, matters of procedure, or
provide an equivalent level of safety as existing requirements;
therefore, there would be similar environmental impacts from the
implementation of the part 53 regulations as there are for existing
requirements.
The preliminary determination of this EA is that there will be no
significant effect on the quality of the human environment from this
action. Public stakeholders should note, however, that comments on any
aspect of this EA may be submitted to the NRC as indicated under the
ADDRESSES section of this document. The EA is available as indicated
under the ``Availability of Documents'' section of this document.
The NRC has sent a copy of the EA, and this proposed rule to every
State Liaison Officer and has requested comments.
XIV. Paperwork Reduction Act
This proposed rule contains new collections of information
contained in parts 26, 50, 53, and 73 and NRC Forms 361S, 366, 366A,
366B, 893, and 894 that are subject to the Paperwork Reduction Act of
1995 (44 U.S.C. 3501 et seq). The collections of information have been
submitted to the OMB for review and approval. The proposed changes to
parts 2, 10, 11, 19, 20, 21, 25, 30, 40, 51, 70, 72, 74, 75, 95, 140,
150, 170, and 171 do not contain any new or amended collections of
information subject to the Paperwork Reduction Act of 1995. Existing
collections of information were approved by the OMB, approval numbers
3150-0062 (part 11), 3150-0044 (part 19), 3150-0014 (part 20), 3150-
0035 (part 21), 3150-0046 (part 25), 3150-0017 (part 30), 3150-0020
(part 40), 3150-0021 (part 51), 3150-0024 (NRC Form 396), 3150-0090
(NRC Form 398), 3150-0009 (part 70), 3150-0132 (part 72), 3150-0123
(part 74), 3150-0055 (part 75), 3150-0047 (part 95), 3150-0039 (part
140), and 3150-0032 (part 150).
Type of submission, new or revision: Revision and new.
The title of the information collection: Risk-Informed, Technology-
Inclusive Regulatory Framework for Advanced Reactors.
[[Page 87009]]
The form number if applicable: NRC Forms 361S, 366, 366A, 366B,
893, and 894.
How often the collection is required or requested: Once, on
occasion, every 30 days, biannually, annually, biennially, every four
years, every five years, every ten years.
Who will be required or asked to respond: Part 53 commercial
nuclear plant licensees and license applicants for commercial nuclear
plants to be licensed under part 53.
An estimate of the number of annual responses: 15 (2 responses for
Part 26, 11 responses for Part 53, 2 responses for Part 50 and 0
responses for Part 73 and NRC Forms 361S, 366, 366A, 366B, 893, and
894)
The estimated number of annual respondents: 2 (2 respondents for
Part 26, 2 respondents for Part 53, 2 respondents for Part 50 and 0
respondents for Part 73 and NRC Forms 361S, 366, 366A, 366B, 893, and
894)
An estimate of the total number of hours needed annually to comply
with the information collection requirement or request: 230,244 hours.
(656 hours for Part 26, 220,801 hours for Part 53, 8,767 hours for Part
50 and 0 hours for Part 73 and NRC Forms 361S, 366, 366A, 366B, 893,
and 894)
Abstract: The NRC is proposing to establish an optional technology-
inclusive regulatory framework for use by applicants for new commercial
nuclear plant designs. The regulatory requirements developed in this
rulemaking would use methods of evaluation, including risk-informed and
performance-based methods, that are flexible and practicable for
application to a variety of new reactor technologies. The NRC's goals
in amending these regulations are to continue to provide reasonable
assurance of adequate protection of public health and safety and the
common defense and security at reactor sites at which new nuclear
reactor designs are deployed to at least the same degree of protection
as required for current-generation LWRs; protect health and minimize
danger to life or property to at least the same degree of protection as
required for current-generation LWRs; provide greater operational
flexibilities where supported by enhanced margins of safety that may be
provided in new nuclear designs; and promote regulatory stability,
predictability, and clarity.
The proposed rule covers diverse topics, which result in
recordkeeping and reporting requirements related to contents of
applications, plant design and analysis, siting, construction and
manufacturing, licensing-basis information, facility operations,
programs, staffing, FFD, physical security, cyber-security, AA,
decommissioning, and quality assurance.
In addition to the new information collections in the proposed
regulations, part 53 would result in new collections via NRC Forms
361S, 366, 366A, 366B, 893, and 894. NRC Forms 366, 366A, and 366B
would be modified to include part 53 reportable events covering an
equivalent scope as the requirements in 10 CFR 50.73, but without LWR-
specific terminology to ensure technology inclusiveness. The proposed
rule also would require part 53 licensees to use NRC Forms 893 and 894
to report on positive drug and alcohol test results (NRC Form 893) and
annual fitness-for-duty program performance (NRC Form 894). Finally, a
new version of NRC Form 361 (NRC Form 361S) would be created for use by
part 53 licensees, covering an equivalent scope as the requirements in
10 CFR 50.72, but without LWR-specific terminology to ensure technology
inclusiveness.
The NRC is seeking public comment on the potential impact of the
information collections contained in this proposed rule and on the
following issues:
1. Is the proposed information collection necessary for the proper
performance of the functions of the NRC, including whether the
information will have practical utility? Please explain your response.
2. Is the estimate of the burden of the proposed information
collection accurate? Please explain your response.
3. Is there a way to enhance the quality, utility, and clarity of
the information to be collected? Please explain your response.
4. How can the burden of the proposed information collection on
respondents be minimized, including the use of automated collection
techniques or other forms of information technology? Please explain
your response.
The OMB clearance documents and proposed rule is available as
indicated under the ``Availability of Documents'' section in this
document or may be viewed free of charge by contacting the NRC's PDR
reference staff at 1-800-397-4209, at 301-415-4737, or by email to
[email protected]. You may obtain information and comment
submissions related to the OMB clearance package by searching on https://www.regulations.gov under Docket ID NRC-2019-0062.
You may submit comments on any aspect of these proposed information
collections, including suggestions for reducing the burden and on the
above issues, by the following methods:
Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2019-0062.
Mail comments to: FOIA, Library, and Information
Collections Branch, Office of the Chief Information Officer, Mail Stop:
T6-A10M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001
or by email to [email protected] or to the OMB reviewer at:
OMB Office of Information and Regulatory Affairs (3150-XXXX, 3150-0002,
-0104, -0146, -0238), Attn: Desk Officer for the Nuclear Regulatory
Commission, 725 17th Street NW, Washington, DC 20503.
Submit comments by December 2, 2024. Comments received after this
date will be considered if it is practical to do so, but the NRC staff
is able to ensure consideration only for comments received on or before
this date.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a collection of information unless the document requesting
or requiring the collection displays a currently valid OMB control
number.
XV. Criminal Penalties
For the purposes of Section 223 of the Act, the NRC is issuing this
proposed rule that would add a new part 53 and amend parts 26 and 73
under one or more of Sections 161b, 161i, or 161o of the Act, except as
noted in proposed Sec. 53.9010(b) and Sec. 26.825(b). Willful
violations of the part 53 and part 26 regulations not listed in
proposed Sec. 53.9010(b) and Sec. 26.825(b) would be subject to
criminal enforcement. Criminal penalties as they apply to regulations
in part 53 would be discussed in Sec. 53.9010.
XVI. Voluntary Consensus Standards
The NTTAA requires that Federal agencies use technical standards
that are developed or adopted by voluntary consensus standards bodies
unless the use of such a standard is inconsistent with applicable law
or otherwise impractical. In this proposed rule, the NRC would revise
regulations by adding a risk-informed, technology-inclusive regulatory
framework for commercial advanced nuclear reactors. This action does
not constitute the establishment of a standard that contains generally
applicable requirements.
XVII. Availability of Guidance
As discussed in section II, Background, of this document, the NRC's
development of proposed part 53
[[Page 87010]]
built upon recent and ongoing activities such as those described in
SECY-19-0117. Because a number of those activities are ongoing to
support new reactor applications under the existing regulatory
framework of 10 CFR parts 50 and 52, the NRC staff identified in its
response to SRM-SECY-20-0032 that the timing of guidance document
development to support the part 53 rulemaking was a key risk and
uncertainty to publishing the final part 53 rule. To mitigate this
risk, the NRC engaged external stakeholders to ensure a common
prioritization of the development of these guidance documents and to
work diligently on those that would be needed to support this
rulemaking, forthcoming applications, or broader efforts such as the
Advanced Reactor Demonstration Program being sponsored by the DOE. The
NRC also recognizes that guidance development to support part 53 and
advanced reactors will continue as the industry and NRC learn lessons
from licensing reviews and operating experience. Therefore, the NRC
categorized guidance supporting the part 53 rulemaking into three
categories: (1) guidance issued or under development to support
applications under the existing regulatory framework; (2) implementing
guidance for part 53-specific proposed rule language; and (3) future
guidance activities that would need to be completed after the part 53
proposed rule is published for public comment.
(1) Hundreds of guidance documents exist for the current fleet of
operating reactors. While some of the guidance is specific to LWR
technologies, other guidance is technology inclusive in nature and
should be considered, as appropriate, in the development of all
licensing applications and NRC reviews. In addition, the NRC has
undertaken efforts to incorporate or reference the most relevant
guidance in its efforts to develop additional guidance for future
advanced reactors. The NRC has issued the following guidance to support
licensing reviews of advanced reactors under the existing regulatory
framework that will continue to inform applicant development and NRC
reviews under parts 50 and 52. Conforming changes to these guidance
documents would be needed to ensure they are applicable under part 53.
The NRC will issue revisions or part 53-related companions to these
guidance documents for public comment after the publication of this
proposed rule and then finalize and issue the guidance documents with
or after the final part 53 rule.
RG 1.233, ``Guidance for a Technology-Inclusive, Risk-
Informed, and Performance-Based Methodology to Inform the Licensing
Basis and Content of Applications for Licenses, Certifications, and
Approvals for Non-Light-Water Reactors''
RG 1.247 for trial use, ``Acceptability of Probabilistic Risk
Assessment Results for Non-Light-Water Reactor Risk-Informed
Activities''
NUREG-2246, ``Fuel Qualification for Advanced Reactors''
RG 1.87, Revision 2, ``Acceptability of ASME Code, Section
III, Division 5, ``High Temperature Reactors''
RG 1.246, ``Acceptability of ASME Code, Section XI, Division
2, `Requirements for Reliability And Integrity Management (RIM)
Programs for Nuclear Power Plants,' for Non-Light Water Reactors''
Also, the NRC continues to develop additional guidance to support
licensing reviews of advanced reactors under the existing regulatory
framework. Some of these guidance documents have been issued and others
will be issued before the finalization of part 53 to support near-term
applicants and NRC reviews. For example, the NRC has been and continues
to be engaged with the DOE and industry to develop content of
application guidance and other regulatory guidance for advanced
reactors to support applications and subsequent operations under the
existing regulatory framework. These guidance documents, such as the
industry-led Technology-Inclusive Content of Application Project
guidance found in NEI 21-07, Revision 1, and the NRC-led Advanced
Reactor Content of Application Project (ARCAP) interim staff guidance
(ISG) documents and NRC regulatory guidance endorsing NEI 21-07,
Revision 1, will support developers in preparing advanced reactor
applications. These guidance documents provide an overview of the
information that should be included in an advanced reactor application,
a review roadmap for the NRC with the principal purpose of ensuring
consistency, quality, and uniformity of NRC reviews, and a well-defined
base from which the NRC can evaluate proposed changes in the scope and
requirements of reviews. While specific sections of the information are
primarily aligned with the LMP methodology, as endorsed in RG 1.233, as
one acceptable process for applicants to use when developing portions
of an application, the concepts and general information may be used to
inform the review of an application submitted using other traditional
licensing approach methodologies (as applicable). Other sections of the
information are generally applicable and independent of the methodology
used to develop an advanced reactor application. The ARCAP ISGs provide
references to numerous regulatory guidance documents that should be
considered by both applicants and the NRC in developing and reviewing,
respectively, advanced reactor applications. The NRC has issued the
following documents separately from this proposed rule. The NRC may
issue other, related guidance documents with or after the final part 53
rule.
RG 1.253, ``Guidance for a Technology Inclusive Content of
Application Methodology to Inform the Licensing Basis and Content of
Applications for Licenses, Certifications, and Approvals for Non-Light-
Water Reactors''
DANU-ISG-2022-01, ``Advanced Reactor Content of Application
Project, `Review of Risk-Informed, Technology-Inclusive Advanced
Reactor Applications--Roadmap' ''
DANU-ISG-2022-02, ``Advanced Reactor Content of Application
Project Chapter 2, `Site Information' ''
DANU-ISG-2022-03, ``Advanced Reactor Content of Application
Project Chapter 9, `Control of Routine Plant Radioactive Effluents,
Plant Contamination and Solid Waste' ''
DANU-ISG-2022-04, ``Advanced Reactor Content of Application
Project Chapter 10, `Control of Occupational Dose' ''
DANU-ISG-2022-05, ``Advanced Reactor Content of Application
Project Chapter 11, `Organization and Human-System Considerations' ''
DANU-ISG-2022-06, ``Advanced Reactor Content of Application
Project Chapter 12, `Post-Construction Inspection, Testing, and
Analysis Program' ''
DANU-ISG-2022-07, ``Advanced Reactor Content of Application
Project, `Risk-Informed Inservice Inspection/Inservice Testing' ''
DANU-ISG-2022-08, ``Advanced Reactor Content of Application
Project, `Risk-Informed Technical Specifications' ''
DANU-ISG-2022-09, ``Advanced Reactor Content of Application
Project, `Risk-Informed, Performance-Based Fire Protection Program (for
Operations)' ''
RG 1.242, ``Performance-Based Emergency Preparedness for Small
Modular Reactors, Non-Light-Water Reactors, and Non-Power Production or
Utilization Facilities''
RG 4.7, ``General Site Suitability Criteria for Nuclear Power
Stations''
[[Page 87011]]
(2) The NRC is issuing for comment nine draft guidance documents
for the implementation of the proposed requirements in this rulemaking.
The guidance is available in ADAMS under the Accession Numbers as
indicated under the ``Availability of Documents'' section in this
document. Comments on this draft regulatory guidance may be submitted
by the methods outlined in the ADDRESSES section of this document.
Interested persons may obtain information and comment submissions
related to the draft guidance by searching on https://www.regulations.gov under Docket ID NRC-2019-0062.
DG-1413, ``Technology-Inclusive Identification of Licensing
Events for Commercial Nuclear Plants''
This DG describes an acceptable approach for identifying licensing
events that can be used to inform the design basis, licensing basis,
and content of applications for commercial nuclear plants, including
large LWRs and non-LWRs. It applies to nuclear power reactor designers,
applicants, and licensees of commercial nuclear plants applying for
permits, licenses, certifications, and approvals under parts 50, 52,
and 53. In this DG, the term ``licensing events'' is used in a generic
sense to refer to collections of designated event categories such as,
but not limited to AOOs, DBAs, DBEs, and postulated accidents.
Specifically, this DG provides an acceptable approach for: (1)
conducting a comprehensive and systematic search for initiating events;
(2) using a systematic process to delineate a comprehensive set of
event sequences; (3) grouping initiating events and event sequences
into designated licensing event categories; and (4) providing assurance
that the set of licensing events is complete.
DG-5073, ``Fitness For Duty Programs for Commercial Nuclear
Plants And Manufacturing Facilities Licensed Under 10 CFR part 53''
This DG describes guidance for applicants under part 53 and
licensees and other entities described in Sec. 26.3(f) who would elect
to or be required to implement FFD programs for facilities licensed
under part 53. The FFD program requirements would be detailed in
subpart M of part 26 and involve, in part, policies, procedures, drug
and alcohol testing, laboratory requirements, behavioral observation,
MRO responsibilities, fitness determinations, reporting, and
recordkeeping. The FFD program for facilities licensed under part 53
subject to part 26 would also include requirements for a PMRP and FFD
program change control that licensees or other entities must implement
to maintain an effective FFD program.
DG-5074, ``Access Authorization Program for Commercial Nuclear
Plants''
This DG describes a method that the staff considers acceptable to
comply with requirements in proposed Sec. 73.120, ``Access
authorization program for commercial nuclear plants,'' related to an AA
program. This document provides guidance and would be one NRC-approved
method (not the only method) for meeting regulatory requirements for
part 53. The proposed language in Sec. 73.120 would provide
flexibility through availability of the use of an alternate approach,
commensurate with risk and consequence to public health and safety, for
part 53 applicants who demonstrate in an analysis that the offsite
consequences satisfy the criterion defined in proposed Sec.
53.860(a)(2)(i).
DG-5075, ``Establishing Cybersecurity Programs for Commercial
Nuclear Plants Licensed Under 10 CFR part 53''
This DG describes an approach the NRC staff deems acceptable for
complying with the Commission's proposed regulations for establishing,
implementing, and maintaining a cybersecurity program at commercial
nuclear plants that would be licensed under part 53. This guidance
provides an approach for meeting the requirements of proposed Sec.
73.110, ``Technology-inclusive requirements for protection of digital
computer and communication systems and networks.''
DG-5076, ``Guidance for Technology Inclusive Requirements for
Physical Protection of Licensed Activities at Commercial Nuclear
Plants''
This DG describes methods and approaches that the NRC staff
considers acceptable for meeting the proposed physical security
requirements of part 53 and Sec. 73.100. The guidance is intended to
provide methods and considerations for complying with Sec. 53.440(f)
safety and security design process considerations, determining
eligibility for meeting the performance criterion in Sec. 53.860 to
relieve the applicant from the applicable requirements to defend
against radiological sabotage outlined in Sec. 73.55 or Sec. 73.100,
and (if the required analysis for eligibility is not satisfied)
applying the physical security requirements of Sec. 73.100 as an
alternative pathway from Sec. 73.55 for protection against
radiological sabotage.
DG-5078, ``Fatigue Management for Nuclear Power Plant
Personnel at Commercial Nuclear Plants Licensed Under 10 CFR part 53''
This DG describes proposed methods that the NRC staff considers
acceptable for addressing certain aspects of FFD programs that would be
established at commercial nuclear facilities licensed under part 53.
This guidance, in conjunction with the existing RG 5.73, ``Fatigue
Management for Nuclear Plant Personnel,'' would provide comprehensive
guidance regarding acceptable methods for the development and
implementation of licensee fatigue-management programs.
The NRC is issuing for public comment the following draft ISG
documents for the implementation of NRC staff review of applications
under the proposed requirements in this rulemaking:
DRO-ISG-2023-01, ``Operator Licensing Programs''
This draft ISG provides guidance for the review of tailored
operator licensing programs that are submitted for review consistent
with the technical requirements of proposed Sec. 53.730(g). This
guidance primarily addresses the review of operator licensing
examination processes to facilitate the ability of reviewers to assess
whether a proposed approach to the testing of licensed operators and
trainees reflects sound assessment testing practices that are suitable
for the screening of competent licensed operators. Additionally, this
ISG provides further review guidance in other areas such as licensed
operator continuing training and proficiency programs.
DRO-ISG-2023-02, ``Interim Staff Guidance Augmenting NUREG-
1791, `Guidance for Assessing Exemption Requests from the Nuclear Power
Plant Licensed Operator Staffing Requirements Specified in 10 CFR
50.54(m),' for Licensing Commercial Nuclear Plants under 10 CFR part
53''
This draft ISG provides guidance for the review of customized
facility operator staffing plans that are submitted for review
consistent with the technical requirements of proposed Sec. 53.730(f).
This ISG is structured as a companion document to the existing NUREG-
1791 and adapts the existing HFE-based methodologies of that document
for use in the evaluation of staffing plans that would be submitted
within the context of part 53 facilities. Additionally, this ISG
provides further guidance to address other staffing-related
considerations, such as provisions for engineering expertise.
DRO-ISG-2023-03, ``Development of Scalable Human Factors
Engineering Review Plans''
[[Page 87012]]
This draft ISG applies to the HFE review of applications for OLs,
COLs, DCs, and standard design approvals for commercial nuclear plants
submitted under proposed part 53. The purpose of this ISG is to
facilitate NRC understanding of an acceptable method for developing a
scalable (i.e., application-specific) plan for the review of these
applications for compliance with applicable HFE requirements. The ISG
describes a process and provides implementation guidance for the NRC to
tailor HFE review plans to each application to achieve an effective and
efficient review.
(3) The NRC has identified future guidance activities that would
need to be completed after the part 53 proposed rule is published for
public comment to support advanced reactor applications and NRC
reviews. For example, the NRC recognizes that new guidance would be
needed for the implementation of provisions in proposed Sec. 53.620(d)
and the associated licensing provisions in proposed subpart H that
would allow and establish requirements for the loading of fuel into a
manufactured reactor for subsequent transport to and use at a
commercial nuclear plant that will operate the facility pursuant to a
COL. The NRC has not yet initiated the development of guidance
documents in this category but will engage stakeholders during the
development of these documents to ensure common prioritization. In
addition, the NRC works with standards development organizations,
advanced reactor developers, DOE, and other stakeholders to identify
and facilitate new consensus codes and standards needed for advanced
reactor development. The NRC will continue its membership and
participation on standards development committees and working groups to
support standards for advanced reactor technologies, where appropriate.
XVIII. Public Meeting
The NRC will conduct a public meeting on this proposed rule for the
purpose of describing the proposed rule and implementation guidance to
the public and answering questions from the public on the proposed rule
and implementation guidance.
The NRC will publish a notice of the public meeting's location,
time, and agenda on the NRC's public meeting website at least 10
calendar days before the meeting. Stakeholders should monitor the NRC's
public meeting website for information about the public meeting at:
https://www.nrc.gov/public-involve/public-meetings/index.cfm.
XIX. Availability of Documents
The documents identified in the following table are available to
interested persons through one or more of the following methods, as
indicated.
------------------------------------------------------------------------
ADAMS accession No./Web
Document link/ Federal Register
Citation
------------------------------------------------------------------------
Proposed Rule Documents
------------------------------------------------------------------------
Federal Register Notification, ``Proposed ML24095A161.
Rule: Risk-Informed, Technology-Inclusive
Regulatory Framework for Advanced
Reactors,'' October, 2024.
``Draft Environmental Assessment for the ML24095A163.
Proposed Rule--Risk Informed, Technology-
Inclusive Regulatory Framework for Advanced
Reactors,'' October, 2024.
``Draft Regulatory Analysis for the Proposed ML24095A166.
Rule: Risk Informed, Technology-Inclusive
Regulatory Framework for Advanced
Reactors,'' October, 2024.
------------------------------------------------------------------------
Information Collection Documents
------------------------------------------------------------------------
Draft Supporting Statement for Information ML21162A109.
Collection Analysis--10 CFR Part 53.
Draft Supporting Statement for Information ML23030A400.
Collection Analysis--10 CFR Part 26.
Draft Supporting Statement for Information ML24220A036.
Collection Analysis--10 CFR Part 50.
Draft Supporting Statement for Information ML23030A576.
Collection Analysis--10 CFR Part 73.
Draft Supporting Statement for Information ML24220A034.
Collection Analysis--NRC Form 361S.
Draft Supporting Statement for Information ML24220A035.
Collection Analysis--NRC Form 366.
Draft Supporting Statement for Information ML24220A033.
Collection Analysis--NRC Form 893 and 894.
Proposed Rule--Part 26 Burden Tables for Risk ML24240A008.
Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors.
Proposed Rule--Part 50 Burden Tables for Risk ML24220A061.
Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors.
Proposed Rule--Part 53 Burden Tables for Risk ML24220A060.
Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors.
Proposed Rule--Part 73 Burden Tables for Risk ML24240A009.
Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors.
Draft NRC Form 361S, ``Part 53 Plant Event ML23032A443.
Notification Worksheet''.
Draft NRC Form 366, ``Licensee Event Report ML23032A445.
(LER)''.
Draft NRC Form 366A, ``Licensee Event Report ML23032A447.
(LER) Continuation Sheet''.
Draft NRC Form 366B, ``Licensee Event Report ML23032A454.
(LER) (Failure Continuation)''.
Draft NRC Form 893, ``10 CFR Part 26, Subpart ML23032A435.
M, Single FFD Policy Violation Form''.
Draft NRC Form 894, ``10 CFR Part 26, Subpart ML23032A439.
M, Annual Reporting Form for FFD Performance
Information''.
------------------------------------------------------------------------
Draft Regulatory Guidance Documents
------------------------------------------------------------------------
DG-1413, ``Technology-Inclusive ML22257A173.
Identification Of Licensing Events For
Commercial Nuclear Plants,'' October, 2024.
DG-5073, ``Fitness-For-Duty Programs For ML22200A037.
Commercial Nuclear Plants And Manufacturing
Facilities Licensed Under 10 CFR Part 53,''
October, 2024.
DG-5074, ``Access Authorization Program for ML22199A246.
Commercial Nuclear Plants,'' October, 2024.
DG-5075, ``Establishing Cybersecurity ML22199A257.
Programs For Commercial Nuclear Plants
Licensed Under 10 CFR Part 53,'' October,
2024.
DG-5076, ``Guidance for Technology Inclusive ML22203A131.
Requirements for Physical Protection of
Licensed Activities at Commercial Nuclear
Plants,'' October, 2024.
[[Page 87013]]
DG-5078, ``Fatigue Management For Nuclear ML22264A109.
Power Plant Personnel At Commercial Nuclear
Plants Licensed Under 10 CFR Part 53,''
October, 2024.
------------------------------------------------------------------------
Draft ISG Documents
------------------------------------------------------------------------
Draft ISG DRO-ISG-2023-01, ``Operator ML22266A066.
Licensing Programs,'' October, 2024.
Draft ISG DRO-ISG-2023-02, ``Interim Staff ML22266A068.
Guidance Augmenting NUREG-1791, `Guidance
for Assessing Exemption Requests from the
Nuclear Power Plant Licensed Operator
Staffing Requirements Specified in 10 CFR
50.54(m),' for Licensing Commercial Nuclear
Plants under 10 CFR Part 53,'' October, 2024.
Draft ISG DRO-ISG-2023-03, ``Development of ML22266A072.
Scalable Human Factors Engineering Review
Plans,'' October, 2024.
------------------------------------------------------------------------
Other References
------------------------------------------------------------------------
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Conformity Assessment Activities,'' dated circulars_a119_a119fr.
February 19, 1998.
PNNL, Technical Letter Report, ``The Use of ML18081A607.
Electronic Communications to Perform
Determinations of Fitness,'' dated August
2017.
Pre-decisional DG, ML22276A149.
``Technology[dash]Inclusive,
Risk[dash]Informed, and
Performance[dash]Based Methodology for
Seismic Design of Commercial Nuclear
Plants,'' dated October 3, 2022.
Research Information Letter 2021-04, ML21113A066.
``Feasibility Study on a Potential
Consequence-Based Seismic Design Approach
for Nuclear Facilities,'' dated April 2021.
RG 1.110, Revision 1, ``Cost-Benefit Analysis ML13241A052.
for Radwaste Systems for Light-Water-Cooled
Nuclear Power Reactors,'' dated October 2013.
RG 1.134, Revision 4, ``Medical Assessment Of ML14189A385.
Licensed Operators Or Applicants For
Operator Licenses At Nuclear Power Plants,''
dated September 2014.
RG 1.174, ``An Approach for Using ML17317A256.
Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific Changes
to the Licensing Basis,'' Revision 3, dated
January 2018.
RG 1.208, ``A Performance-Based Approach to ML070310619.
Define the Site-Specific Earthquake Ground
Motion,'' dated March 2007.
RG 1.232, ``Guidance for Developing Principal ML17325A611.
Design Criteria for Non-Light-Water
Reactors,'' Revision 0, dated April 2018.
RG 1.233, Revision 0, ``Guidance for a ML20091L698.
Technology-Inclusive, Risk-Informed, and
Performance-Based Methodology to Inform the
Licensing Basis and Content of Applications
for Licenses, Certifications, and Approvals
for Non-Light-Water Reactors,'' dated June
2020.
RG 1.247, ``Acceptability of Probabilistic ML21235A008.
Risk Assessment Results for Non-Light-Water
Reactor Risk-Informed Activities,'' issued
March 2022 for trial use.
RG 5.73, ``Fatigue Management for Nuclear ML083450028.
Power Plant Personnel,'' dated March 20,
2009.
RG 5.77, ``Insider Mitigation Program,'' ML16342B024.
Revision 1, dated September 08, 2022.
RG 5.81, ``Target Set Identification and ML13151A355.
Development for Nuclear Power Reactors,''
Revision 1, dated December 2019 (non-public).
SECY-18-0096, ``Functional Containment ML18115A157.
Performance Criteria For Non-Light-Water-
Reactors,'' dated September 28, 2018.
SECY-19-0117, ``Technology-Inclusive, Risk- ML18311A264 (package).
Informed, and Performance-Based Methodology
to Inform the Licensing Basis and Content of
Applications for Licenses, Certifications,
and Approvals for Non-Light-Water
Reactors,'' dated December 2019.
SECY-20-0032, ``Rulemaking Plan on `Risk- ML19340A056.
Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors (RIN-3150-
AK31; NRC-2019-0062,''' dated April 13, 2020.
SECY-20-0070, ``(Redacted) Technical ML20126G265 (package).
Evaluation of the Security Bounding Time
Concept for Operating Nuclear Power
Plants,'' dated November 8, 2021.
SECY-22-0072, ``Proposed Rule: Alternative ML21334A003 (package).
Physical Security Requirements for Advanced
Reactors (RIN 3150-AK19),'' dated August 2,
2022.
SECY-83-293, ``Amendments to 10 CFR 50 ML21278A823 (non-public);
Related to Anticipated Transients Without ML21278A994 (non-
Scram (ATWS) Events,'' dated July 19, 1983. public).
SECY-93-092, ``Issues Pertaining to the ML040210725.
Advanced Reactor (PRISM, MHTGR, and PIUS)
and CANDU 3 Designs and their Relationship
to Current Regulatory Requirements,'' dated
April 8, 1993.
[[Page 87015]]
SRM-SECY-10-0121, ``Modifying the ML110610166.
Risk[dash]Informed Regulatory Guidance for
New Reactors,'' dated March 2, 2011.
SRM-SECY-17-0100, ``Security Baseline ML18283A072.
Inspection Program Assessment Results and
Recommendations for Program Efficiencies,''
dated October 8, 2018.
SRM-SECY-20-0032, ``Rulemaking Plan on ML20276A293.
`Risk[dash]Informed,
Technology[dash]Inclusive Regulatory
Framework for Advanced Reactors (RIN-3150-
AK31; NRC-2019-0062),''' dated October 2,
2020.
SRM-SECY-20-0045, ``Population Related Siting ML22194A885.
Considerations for Advanced Reactors,''
dated July 30, 2022.
SRM-SECY-98-144, ``Staff Requirements--SECY- ML003753593.
98-144--White Paper on Risk-Informed and
Performance-Based Regulations,'' dated
February 24, 1999.
SECY-23-0021, ``Proposed Rule: Risk-Informed, ML21162A095.
Technology-Inclusive Regulatory Framework
for Advanced Reactors (RIN 3150-AK31),''
March 1, 2023.
SECY-23-0021, Enclosure 1, ``Draft Federal ML21162A102.
Register Notification''.
SECY-23-0021, Enclosure 2, ``Draft ML21162A104.
Environmental Assessment for the Proposed
Rule--Risk Informed, Technology-Inclusive
Regulatory Framework for Advanced Reactors''.
SECY-23-0021, Enclosure 3, ``Draft Regulatory ML21165A112.
Analysis for the Proposed Rule: Risk
Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors''.
SECY-23-0021, Enclosure 4, ``Alternative ML22244A001.
Approaches Considered for Selected Topics
During the Development of 10 CFR Part 53''.
SECY-23-0021, Enclosure 5, ``Estimated ML22304A099 (non-public).
Resources for The Risk-Informed, Technology-
Inclusive Regulatory Framework For Advanced
Reactors Rulemaking''.
Staff Requirements--SECY-23-0021, ``Proposed ML24064A047 (package).
Rule: Risk-Informed, Technology-Inclusive
Regulatory Framework for Advanced Reactors
(RIN 3150-AK31),'' March 4, 2024.
------------------------------------------------------------------------
Throughout the development of this rule, the NRC may post documents
related to this rule, including public comments, on the Federal
rulemaking website at https://www.regulations.gov under Docket ID NRC-
2019-0062. The Federal rulemaking website allows you to receive alerts
when changes or additions occur in a docket folder. To subscribe: (1)
Navigate to the docket folder (NRC-2019-0062-0012); (2) click the
``Sign up for Email Alerts'' link; and (3) enter your email address and
select how frequently you would like to receive emails (daily, weekly,
or monthly).
List of Subjects
10 CFR Part 1
Flags, Organization and functions (Government Agencies), Seals and
insignia.
10 CFR Part 2
Administrative practice and procedure, Antitrust, Byproduct
material, Classified information, Confidential business information,
Freedom of information, Environmental protection, Hazardous waste,
Nuclear energy, Nuclear materials, Nuclear power plants and reactors,
Penalties, Reporting and recordkeeping requirements, Sex
discrimination, Source material, Special nuclear material, Waste
treatment and disposal.
10 CFR Part 10
Administrative practice and procedure, Classified information,
Government employees, Security measures.
10 CFR Part 11
Hazardous materials transportation, Investigations, Nuclear energy,
Nuclear materials, Penalties, Reporting and recordkeeping requirements,
Security measures, Special nuclear material.
10 CFR Part 19
Criminal penalties, Environmental protection, Nuclear Energy,
Nuclear materials, Nuclear power plants and reactors, Occupational
safety and health, Penalties, Radiation protection, Reporting and
recordkeeping requirements, Sex discrimination.
10 CFR Part 20
Byproduct material, Criminal penalties, Hazardous waste, Licensed
material, Nuclear energy, Nuclear materials, Nuclear power plants and
reactors, Occupational safety and health, Packaging and containers,
Penalties, Radiation protection, Reporting and recordkeeping
requirements, Source material, Special nuclear material, Waste
treatment and disposal.
10 CFR Part 21
Nuclear power plants and reactors, Penalties, Radiation protection,
Reporting and recordkeeping requirements.
10 CFR Part 25
Classified information, Criminal penalties, Investigations,
Penalties, Reporting and recordkeeping requirements, Security measures.
10 CFR Part 26
Administrative practice and procedure, Alcohol abuse, Alcohol
testing, Appeals, Drug abuse, Drug testing, Employee assistance
programs, Fitness for duty, Management actions, Nuclear power plants
and reactors, Privacy, Protection of information, Radiation protection,
Reporting and recordkeeping requirements.
10 CFR Part 30
Byproduct material, Criminal penalties, Government contracts,
Intergovernmental relations, Isotopes, Nuclear energy, Nuclear
materials, Penalties, Radiation protection, Reporting and recordkeeping
requirements, Whistleblowing.
10 CFR Part 40
Criminal penalties, Exports, Government contracts, Hazardous
materials transportation, Hazardous waste, Nuclear energy, Nuclear
materials, Penalties, Reporting and recordkeeping requirements, Source
material, Uranium, Whistleblowing.
10 CFR Part 50
Administrative practice and procedure, Antitrust, Backfitting,
Classified information, Criminal penalties, Education, Emergency
planning, Fire prevention, Fire protection, Intergovernmental
relations, Nuclear power plants and reactors, Penalties, Radiation
protection, Reactor siting criteria, Reporting and
[[Page 87016]]
recordkeeping requirements, Whistleblowing.
10 CFR Part 51
Administrative practice and procedure, Environmental impact
statements, Hazardous waste, Nuclear energy, Nuclear materials, Nuclear
power plants and reactors, Reporting and recordkeeping requirements.
10 CFR Part 53
Administrative practice and procedure, Antitrust, Backfitting,
Construction permit, Combined license, Classified information, Criminal
penalties, Early site permit, Emergency planning, Fees, Fire
prevention, Fire protection, Inspection, Intergovernmental relations,
Limited work authorization, Manufacturing license, Nuclear power plants
and reactors, Operating license, Penalties, Prototype, Radiation
protection, Reactor siting criteria, Reporting and recordkeeping
requirements, Standard design, Standard design certification, Training
programs.
10 CFR Part 70
Classified information, Criminal penalties, Emergency medical
services, Hazardous materials transportation, Material control and
accounting, Nuclear energy, Nuclear materials, Packaging and
containers, Penalties, Radiation protection, Reporting and
recordkeeping requirements, Scientific equipment, Security measures,
Special nuclear material, Whistleblowing.
10 CFR Part 72
Administrative practice and procedure, Hazardous waste, Indians,
Intergovernmental relations, Nuclear energy, Penalties, Radiation
protection, Reporting and recordkeeping requirements, Security
measures, Spent fuel, Whistleblowing.
10 CFR Part 73
Criminal penalties, Exports, Hazardous materials transportation,
Imports, Nuclear energy, Nuclear materials, Nuclear power plants and
reactors, Penalties, Reporting and recordkeeping requirements, Security
measures.
10 CFR Part 74
Accounting, Criminal penalties, Hazardous materials transportation,
Material control and accounting, Nuclear energy, Nuclear materials,
Packaging and containers, Penalties, Radiation protection, Reporting
and recordkeeping requirements, Scientific equipment, Special nuclear
material.
10 CFR Part 75
Criminal penalties, Intergovernmental relations, Nuclear energy,
Nuclear materials, Nuclear power plants and reactors, Penalties,
Reporting and recordkeeping requirements, Security measures, Treaties.
10 CFR Part 95
Classified information, Criminal penalties, Penalties, Reporting
and recordkeeping requirements, Security measures.
10 CFR Part 140
Criminal penalties, Extraordinary nuclear occurrence, Insurance,
Intergovernmental relations, Nuclear materials, Nuclear power plants
and reactors, Penalties, Reporting and recordkeeping requirements.
10 CFR Part 150
Criminal penalties, Hazardous materials transportation,
Intergovernmental relations, Nuclear energy, Nuclear materials,
Penalties, Reporting and recordkeeping requirements, Security measures,
Source material, Special nuclear material.
10 CFR Part 170
Byproduct material, Import and export licenses, Intergovernmental
relations, Non-payment penalties, Nuclear energy, Nuclear materials,
Nuclear power plants and reactors, Source material, Special nuclear
material.
10 CFR Part 171
Annual charges, Approvals, Byproduct material, Holders of
certificates, Intergovernmental relations, Nonpayment penalties,
Nuclear materials, Nuclear power plants and reactors, Registrations,
Source material, Special nuclear material.
For the reasons set out in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended; the Energy Reorganization
Act of 1974, as amended; and 5 U.S.C. 552 and 553, the NRC is proposing
the following amendments to 10 CFR parts 1, 2, 10, 11, 19, 20, 21, 25,
26, 30, 40, 50, 51, 70, 72, 73, 74, 75, 95, 140, 150, 170, and 171 and
adding 10 CFR part 53:
PART 1--STATEMENT OF ORGANIZATION AND GENERAL INFORMATION
0
1. The authority citation for part 1 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 23, 25, 29, 161,
191 (42 U.S.C. 2033, 2035, 2039, 2201, 2241); Energy Reorganization
Act of 1974, secs. 201, 203, 204, 205, 209 (42 U.S.C. 5841, 5843,
5844, 5845, 5849); Administrative Procedure Act (5 U.S.C. 552, 553);
Reorganization Plan No. 1 of 1980, 5 U.S.C. Appendix (Reorganization
Plans).
Sec. 1.43 [Amended]
0
2. In Sec. 1.43, in paragraph (a)(2) remove the cross reference ``10
CFR parts 50, 52, and 54'' and add in its place the cross reference
``10 CFR parts 50, 52, 53, and 54''.
PART 2--AGENCY RULES OF PRACTICE AND PROCEDURE
0
3. The authority citation for part 2 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 29, 53, 62, 63, 81,
102, 103, 104, 105, 161, 181, 182, 183, 184, 186, 189, 191, 234 (42
U.S.C. 2039, 2073, 2092, 2093, 2111, 2132, 2133, 2134, 2135, 2201,
2231, 2232, 2233, 2234, 2236, 2239, 2241, 2282); Energy
Reorganization Act of 1974, secs. 201, 206 (42 U.S.C. 5841, 5846);
Nuclear Waste Policy Act of 1982, secs. 114(f), 134, 135, 141 (42
U.S.C. 10134(f), 10154, 10155, 10161); Administrative Procedure Act
(5 U.S.C. 552, 553, 554, 557, 558); National Environmental Policy
Act of 1969 (42 U.S.C. 4332); 44 U.S.C. 3504 note. Section 2.205(j)
also issued under 28 U.S.C. 2461 note.
Sec. 2.1 [Amended]
0
4. In Sec. 2.1, in paragraph (e) remove the phrase ``part 52'' and add
in its place the phrase ``part 52 or part 53''.
0
5. In Sec. 2.4, revise the definitions for ``Contested proceeding''
and ``Facility'' to read as follows:
Sec. 2.4 Definitions.
* * * * *
Contested proceeding means--
(1) A proceeding in which there is a controversy between the NRC
staff and the applicant for a license or permit concerning the issuance
of the license or permit or any of the terms or conditions thereof;
(2) A proceeding in which the NRC is imposing a civil penalty or
other enforcement action, and the subject of the civil penalty or
enforcement action is an applicant for or holder of a license or
permit, or is or was an applicant for or holder of a license or permit,
or is or was an applicant for a standard design certification under
part 52 or part 53 of this chapter; and
(3) A proceeding in which a petition for leave to intervene in
opposition to an application for a license or permit has been granted
or is pending before the Commission.
* * * * *
Facility means production facility or a utilization facility as
defined in Sec. Sec. 50.2 and 53.020 of this chapter.
* * * * *
[[Page 87017]]
Sec. 2.100 [Amended]
0
6. In Sec. 2.100, remove the phrase ``subpart E of part 52'' and add
in its place the phrase ``subpart E of part 52 or subpart H of part
53''.
0
7. In Sec. 2.101, revise paragraphs (a)(3)(i), (a)(5), (a)(9)
introductory text and paragraph (a)(9)(i) to read as follows:
Sec. 2.101 Filing of application.
(a) * * *
(3) * * *
(i) Submit to the Director, Office of Nuclear Reactor Regulation,
or Director, Office of Nuclear Material Safety and Safeguards, as
appropriate, such additional copies as the regulations in part 50,
subpart A of part 51, and part 53 of this chapter require;
* * * * *
(5) An applicant for a construction permit under parts 50 or 53 of
this chapter or a combined license under parts 52 or 53 of this chapter
for a production or utilization facility which is subject to Sec.
51.20(b) of this chapter, and is of the type specified in Sec.
50.21(b)(2) or (b)(3); or Sec. 50.22; or part 53, as applicable, of
this chapter, or is a testing facility, may submit the information
required of applicants by parts 50, 52, or 53 of this chapter in two
parts. One part shall be accompanied by the information required by
Sec. 50.30(f) of this chapter, Sec. 52.80(b) of this chapter, or
Sec. 53.1100(f) of this chapter, as applicable. The other part shall
include any information required by Sec. 50.34(a) and, if applicable,
Sec. 50.34a of this chapter; or Sec. Sec. 52.79 and 52.80(a) of this
chapter; or Sec. Sec. 53.1109, 53.1306, 53.1309, and 53.1312 of this
chapter; or Sec. Sec. 53.1109, 53.1413, 53.1416, and 53.1419 of this
chapter, as applicable. One part may precede or follow other parts by
no longer than 6 months. If it is determined that either of the parts
as described above is incomplete and not acceptable for processing, the
Director, Office of Nuclear Reactor Regulation, or Director, Office of
Nuclear Material Safety and Safeguards, as appropriate, will inform the
applicant of this determination and the respects in which the document
is deficient. Such a determination of completeness will generally be
made within a period of 30 days. Whichever part is filed first shall
also include the fee required by Sec. 50.30(e) or Sec. 53.1100(e) and
Sec. 170.21 of this chapter and the information required by Sec. Sec.
50.33, 50.34(a)(1), and 52.79(a)(1) of this chapter; or Sec. Sec.
53.1109, 53.1309, and 53.1416 of this chapter, as applicable, and Sec.
50.37 or Sec. 53.1115, as applicable, of this chapter. The Director,
Office of Nuclear Reactor Regulation, or Director, Office of Nuclear
Material Safety and Safeguards, as appropriate, will accept for
docketing an application for a construction permit under part 50 or
part 53 of this chapter or a combined license under parts 52 or 53 of
this chapter for a production or utilization facility which is subject
to Sec. 51.20(b) of this chapter, and is of the type specified in
Sec. 50.21(b)(2) or (b)(3), or Sec. 50.22, or part 53, as applicable,
of this chapter or is a testing facility where one part of the
application as described above is complete and conforms to the
requirements of part 50 of this chapter. The additional parts will be
docketed upon a determination by the Director, Office of Nuclear
Reactor Regulation, or Director, Office of Nuclear Material Safety and
Safeguards, as appropriate, that it is complete.
* * * * *
(9) An applicant for a construction permit for a utilization
facility which is subject to Sec. 51.20(b) of this chapter and is of
the type specified in Sec. 50.21(b)(2) or (b)(3), or Sec. 50.22, or
part 53 of this chapter, an applicant for or holder of an early site
permit under part 52 or part 53 of this chapter, or an applicant for a
combined license under parts 52 or 53 of this chapter, who seeks to
conduct the activities authorized under Sec. 50.10(d) or Sec. 53.1130
of this chapter may submit a complete application under paragraphs
(a)(1) through (a)(4) of this section which includes the information
required by Sec. 50.10(d) or Sec. 53.1130 of this chapter.
Alternatively, the applicant (other than an applicant for or holder of
an early site permit) may submit its application in two parts:
(i) Part one must include the information required by Sec.
50.33(a) through (f) or Sec. 53.1109(a) through (e) and Sec. 53.1306
of this chapter, and the information required by Sec. 50.10(d)(2) and
(d)(3) or Sec. 53.1130(a)(2) and (a)(3) of this chapter, as
applicable.
* * * * *
0
8. In Sec. 2.104, revise paragraph (a) to read as follows:
Sec. 2.104 Notice of hearing.
(a) In the case of an application on which a hearing is required by
the Act or this chapter, or in which the Commission finds that a
hearing is required in the public interest, the Secretary will issue a
notice of hearing to be published in the Federal Register. The notice
must be published at least 15 days, and in the case of an application
concerning a limited work authorization, construction permit, early
site permit, or combined license for a facility of the type described
in Sec. 50.21(b) or 50.22, or subpart H of part 53 of this chapter, as
applicable, or a testing facility, at least 30 days, before the date
set for hearing in the notice.\1\ In addition, in the case of an
application for a limited work authorization, construction permit,
early site permit, or combined license for a facility of the type
described in Sec. 50.22 or subpart H of part 53 of this chapter, as
applicable, or a testing facility, the notice must be issued as soon as
practicable after the NRC has docketed the application. If the
Commission decides, under Sec. 2.101(a)(2), to determine the
acceptability of the application based on its technical adequacy as
well as completeness, the notice must be issued as soon as practicable
after the application has been tendered.
* * * * *
\1\ If the notice of hearing concerning an application for a
limited work authorization, construction permit, early site permit,
or combined license for a facility of the type described in Sec.
50.21(b) or Sec. 50.22, or subpart H of part 53 of this chapter, as
applicable, or a testing facility, does not specify the time and
place of initial hearing, a subsequent notice will be published in
the Federal Register which will provide at least 30-day notice of
the time and place of that hearing. After this notice is given, the
presiding officer may reschedule the commencement of the initial
hearing for a later date or reconvene a recessed hearing without
again providing at least 30-day notice.
0
9. In Sec. 2.105, revise paragraph (a) introductory text and
paragraphs (a)(4), (a)(10), (a)(12), (a)(13), (b)(3) introductory text,
(b)(3)(i), (ii), and (iv) to read as follows:
Sec. 2.105 Notice of proposed action.
(a) If a hearing is not required by the Act or this chapter, and if
the Commission has not found that a hearing is in the public interest,
it will, before acting thereon, publish in the Federal Register, as
applicable, or on the NRC's website, https://www.nrc.gov, or both, at
the Commission's discretion, either a notice of intended operation
under Sec. 52.103(a) or Sec. 53.1452(a) of this chapter, as
applicable, and a proposed finding that inspections, tests, analyses,
and acceptance criteria for a combined license under subpart C of part
52 or under subpart H of part 53 of this chapter, have been or will be
met, or a notice of proposed action with respect to an application for:
* * * * *
(4) An amendment to an operating license, combined license, or
manufacturing license for a facility licensed under Sec. 50.21(b) or
Sec. 50.22 or under subpart H of part 53 of this chapter, as
applicable, or for a testing facility, as follows:
(i) If the Commission determines under Sec. 50.58 or Sec. 53.1515
of this
[[Page 87018]]
chapter that the amendment involves no significant hazards
consideration, though it will provide notice of opportunity for a
hearing pursuant to this section, it may make the amendment immediately
effective and grant a hearing thereafter; or
(ii) If the Commission determines under Sec. Sec. 50.58 and 50.91
or Sec. 53.1515 of this chapter, as applicable, that an emergency
situation exists or that exigent circumstances exist and that the
amendment involves no significant hazards consideration, it will
provide notice of opportunity for a hearing pursuant to Sec. 2.106 (if
a hearing is requested, it will be held after issuance of the
amendment);
* * * * *
(10) In the case of an application for an operating license for a
facility of a type described in Sec. 50.21(b) or Sec. 50.22, or part
53 of this chapter or a testing facility, a notice of opportunity for
hearing shall be issued as soon as practicable after the application
has been docketed; or
* * * * *
(12) An amendment to an early site permit issued under subpart A of
part 52, or under subpart H of part 53 of this chapter, as follows:
(i) If the early site permit does not provide authority to conduct
the activities allowed under Sec. 50.10(e)(1) or Sec. 53.1130(b)(1)
of this chapter, the amendment will involve no significant hazards
consideration, and though the NRC will provide notice of opportunity
for a hearing under this section, it may make the amendment immediately
effective and grant a hearing thereafter; and
(ii) If the early site permit provides authority to conduct the
activities allowed under Sec. 50.10(e)(1) or Sec. 53.1130(b)(1) of
this chapter and the Commission determines under Sec. Sec. 50.58 and
50.91 or Sec. 53.1515 of this chapter that an emergency situation
exists or that exigent circumstances exist and that the amendment
involves no significant hazards consideration, it will provide notice
of opportunity for a hearing under Sec. 2.106 of this chapter (if a
hearing is requested, which will be held after issuance of the
amendment).
(13) A manufacturing license under subpart F of part 52 or subpart
H of part 53 of this chapter.
(b) * * *
(3) For a notice of intended operation under Sec. 52.103(a) or
Sec. 53.1452(a) of this chapter, the following information:
(i) The identification of the NRC action as making the finding
required under Sec. 52.103(g) or Sec. 53.1452(g) of this chapter;
(ii) The manner in which the licensee notifications under Sec.
52.99(c) or Sec. 53.1449(c), of this chapter which are required to be
made available by Sec. 52.99(e)(2) or Sec. 53.1449(e)(2), of this
chapter may be obtained and examined;
* * * * *
(iv) Any conditions, limitations, or restrictions to be placed on
the license in connection with the finding under Sec. 52.103(g) or
Sec. 53.1452(g) of this chapter, and the expiration date or
circumstances (if any) under which the conditions, limitations or
restrictions will no longer apply.
* * * * *
0
10. In Sec. 2.106, revise paragraphs (a)(2), (a)(3), and (b)(2)
introductory text to read as follows:
Sec. 2.106 Notice of issuance.
(a) * * *
(2) An amendment of a license for a facility of the type described
in Sec. 50.21(b) or Sec. 50.22, or part 53 of this chapter, as
applicable, or a testing facility, whether or not a notice of proposed
action has been previously published; and
(3) The finding under Sec. 52.103(g) or Sec. 53.1452(g) of this
chapter.
(b) * * *
(2) In the case of a finding under Sec. 52.103(g) or Sec.
53.1452(g) of this chapter:
* * * * *
0
11. In Sec. 2.109, revise paragraphs (b), (c), and (d) to read as
follows:
Sec. 2.109 Effect of timely renewal application.
* * * * *
(b) If the licensee of a nuclear power plant licensed under Sec.
50.21(b) or Sec. 50.22 or under subpart H of part 53 of this chapter
files a sufficient application for renewal of either an operating
license or a combined license at least 5 years before the expiration of
the existing license, the existing license will not be deemed to have
expired until the application has been finally determined.
(c) If the holder of an early site permit licensed under subpart A
of part 52 or under subpart H of part 53 of this chapter, as
applicable, files a sufficient application for renewal under Sec.
52.29 or Sec. 53.1173 of this chapter, as applicable, at least 12
months before the expiration of the existing early site permit, the
existing permit will not be deemed to have expired until the
application has been finally determined.
(d) If the licensee of a manufacturing license under subpart F of
part 52, or under subpart H of part 53 of this chapter files a
sufficient application for renewal under Sec. 52.177 or Sec. 53.1295
of this chapter at least 12 months before the expiration of the
existing license, the existing license will not be deemed to have
expired until the application has been finally determined.
* * * * *
0
12. In Sec. 2.110, revise paragraphs (a)(1) and (b) to read as
follows:
Sec. 2.110 Filing and administrative action on submittals for
standard design approval or early review of site suitability issues.
(a)(1) A submittal for a standard design approval under subpart E
of part 52 or under subpart H of part 53 of this chapter shall be
subject to Sec. Sec. 2.101(a) and 2.390 to the same extent as if it
were an application for a permit or license.
* * * * *
(b) Upon initiation of review by the NRC staff of a submittal for
an early review of site suitability issues under appendix Q to part 50
of this chapter, or for a standard design approval under subpart E of
part 52 or under subpart H of part 53 of this chapter, the Director,
Office of Nuclear Reactor Regulation, shall publish in the Federal
Register a notice of receipt of the submittal, inviting comments from
interested persons within 60 days of publication or other time as may
be specified, for consideration by the NRC staff and ACRS in their
review.
* * * * *
0
13. In Sec. 2.202, revise paragraph (e) to read as follows:
Sec. 2.202 Orders.
* * * * *
(e)(1) If the order involves the modification of a part 50 or a
part 53 license and is a backfit, the requirements of Sec. 50.109 or
Sec. 53.1590 of this chapter, as applicable, shall be followed, unless
the licensee has consented to the action required.
(2) If the order involves the modification of combined license
under subpart C of part 52, or subpart H of part 53 of this chapter,
the requirements of Sec. 52.98 or Sec. 53.1443 of this chapter, as
applicable, shall be followed unless the licensee has consented to the
action required.
(3) If the order involves a change to an early site permit under
subpart A of part 52 or under subpart H of part 53 of this chapter, the
requirements of Sec. 52.39 or Sec. 53.1188 of this chapter, as
applicable, must be followed, unless the applicant or licensee has
consented to the action required.
(4) If the order involves a change to a standard design
certification rule referenced by that plant's application, the
requirements, if any, in the referenced design certification rule with
respect to changes must be followed, or, in the absence of these
requirements,
[[Page 87019]]
the requirements of Sec. 52.63 or Sec. 53.1263 of this chapter, as
applicable, must be followed, unless the applicant or licensee has
consented to follow the action required.
(5) If the order involves a change to a standard design approval
referenced by that plant's application, the requirements of Sec.
52.145 or Sec. 53.1221 of this chapter, as applicable, must be
followed unless the applicant or licensee has consented to follow the
action required.
(6) If the order involves a modification of a manufacturing license
under subpart F of part 52 or under subpart H of part 53 of this
chapter, the requirements of Sec. 52.171 or Sec. 53.1288 of this
chapter, as applicable, must be followed, unless the applicant or
licensee has consented to the action required.
0
14. In Sec. 2.309, revise paragraphs (a), (f)(1)(i), (f)(1)(vi) and
(vii), (g), (h)(2), (i)(2), (j) to read as follows:
Sec. 2.309 Hearing requests, petitions to intervene, requirements for
standing, and contentions.
(a) General requirements. Any person whose interest may be affected
by a proceeding and who desires to participate as a party must file a
written request for hearing and a specification of the contentions
which the person seeks to have litigated in the hearing. In a
proceeding under Sec. 52.103 or Sec. 53.1452 of this chapter, as
applicable, the Commission, acting as the presiding officer, will grant
the request if it determines that the requestor has standing under the
provisions of paragraph (d) of this section and has proposed at least
one admissible contention that meets the requirements of paragraph (f)
of this section. For all other proceedings, except as provided in
paragraph (e) of this section, the Commission, presiding officer, or
the Atomic Safety and Licensing Board designated to rule on the request
for hearing and/or petition for leave to intervene, will grant the
request/petition if it determines that the requestor/petitioner has
standing under the provisions of paragraph (d) of this section and has
proposed at least one admissible contention that meets the requirements
of paragraph (f) of this section. In ruling on the request for hearing/
petition to intervene submitted by petitioners seeking to intervene in
the proceeding on the HLW repository, the Commission, the presiding
officer, or the Atomic Safety and Licensing Board shall also consider
any failure of the petitioner to participate as a potential party in
the pre-license application phase under subpart J of this part in
addition to the factors in paragraph (d) of this section. If a request
for hearing or petition to intervene is filed in response to any notice
of hearing or opportunity for hearing, the applicant/licensee shall be
deemed to be a party.
* * * * *
(f) * * *
(1) * * *
(i) Provide a specific statement of the issue of law or fact to be
raised or controverted, provided further, that the issue of law or fact
to be raised in a request for hearing under Sec. 52.103(b) or Sec.
53.1452(b) of this chapter, as applicable, must be directed at
demonstrating that one or more of the acceptance criteria in the
combined license have not been, or will not be met, and that the
specific operational consequences of nonconformance would be contrary
to providing reasonable assurance of adequate protection of the public
health and safety;
* * * * *
(vi) In a proceeding other than one under Sec. 52.103 or Sec.
53.1452 of this chapter provide sufficient information to show that a
genuine dispute exists with the applicant/licensee on a material issue
of law or fact. This information must include references to specific
portions of the application (including the applicant's environmental
report and safety report) that the petitioner disputes and the
supporting reasons for each dispute, or, if the petitioner believes
that the application fails to contain information on a relevant matter
as required by law, the identification of each failure and the
supporting reasons for the petitioner's belief; and
(vii) In a proceeding under Sec. 52.103(b) or Sec. 53.1452(b) of
this chapter, as applicable, the information must be sufficient, and
include supporting information showing, prima facie, that one or more
of the acceptance criteria in the combined license have not been, or
will not be met, and that the specific operational consequences of
nonconformance would be contrary to providing reasonable assurance of
adequate protection of the public health and safety. This information
must include the specific portion of the report required by Sec.
52.99(c) or Sec. 53.1449(c) of this chapter, as applicable, which the
requestor believes is inaccurate, incorrect, and/or incomplete (i.e.,
fails to contain the necessary information required by Sec. 52.99(c)
or Sec. 53.1449(c) of this chapter, as applicable). If the requestor
identifies a specific portion of the report under Sec. 52.99(c) or
Sec. 53.1449(c) of this chapter, as applicable, as incomplete and the
requestor contends that the incomplete portion prevents the requestor
from making the necessary prima facie showing, then the requestor must
explain why this deficiency prevents the requestor from making the
prima facie showing.
* * * * *
(g) Selection of hearing procedures. A request for hearing and/or
petition for leave to intervene may, except in a proceeding under Sec.
52.103 or Sec. 53.1452 of this chapter, as applicable, also address
the selection of hearing procedures, taking into account the provisions
of Sec. 2.310. If a request/petition relies upon Sec. 2.310(d), the
request/petition must demonstrate, by reference to the contention and
the bases provided and the specific procedures in subpart G of this
part, that resolution of the contention necessitates resolution of
material issues of fact which may be best determined through the use of
the identified procedures.
(h) * * *
(2) If the proceeding pertains to a production or utilization
facility (as defined in Sec. 50.2 or Sec. 53.020 of this chapter)
located within the boundaries of the State, local governmental body, or
Federally-recognized Indian Tribe seeking to participate as a party, no
further demonstration of standing is required. If the production or
utilization facility is not located within the boundaries of the State,
local governmental body, or Federally-recognized Indian Tribe seeking
to participate as a party, the State, local governmental body, or
Federally-recognized Indian Tribe also must demonstrate standing.
* * * * *
(i) * * *
(2) Except in a proceeding under Sec. 52.103 or Sec. 53.1452 of
this chapter, as applicable, the participant who filed the hearing
request, intervention petition, or motion for leave to file new or
amended contentions after the deadline may file a reply to any answer.
The reply must be filed within 7 days after service of that answer.
* * * * *
(j) Decision on request/petition. (1) In all proceedings other than
a proceeding under Sec. 52.103 or Sec. 53.1452 of this chapter, as
applicable, the presiding officer shall issue a decision on each
request for hearing or petition to intervene within 45 days of the
conclusion of the initial pre-hearing conference or, if no pre-hearing
conference is conducted, within 45 days after the filing of answers and
replies
[[Page 87020]]
under paragraph (i) of this section. With respect to a request to admit
amended or new contentions, the presiding officer shall issue a
decision on each such request within 45 days of the conclusion of any
pre-hearing conference that may be conducted regarding the proposed
amended or new contentions or, if no pre-hearing conference is
conducted, within 45 days after the filing of answers and replies, if
any. In the event the presiding officer cannot issue a decision within
45 days, the presiding officer shall issue a notice advising the
Commission and the parties, and the notice shall include the expected
date of when the decision will issue.
(2) The Commission, acting as the presiding officer, shall
expeditiously grant or deny the request for hearing in a proceeding
under Sec. 52.103 or Sec. 53.1452 of this chapter, as applicable. The
Commission's decision may not be the subject of any appeal under Sec.
2.311.
0
15. Amend Sec. 2.310 by:
0
a. In paragraphs (a) and (h) introductory text, removing the cross-
reference ``parts 30, 32 through 36, 39, 40, 50, 52, 54, 55, 61, 70 and
72 of this chapter'' and adding, in its place, the cross reference
``parts 30, 32 through 36, 39, 40, 50, 52, 53, 54, 55, 61, 70 and 72 of
this chapter''; and
0
b. Revising paragraphs (i) and (j).
The revisions read as follows.
Sec. 2.310 Selection of hearing procedures.
* * * * *
(i) In design certification rulemaking proceedings under part 52 or
part 53 of this chapter, any informal hearing held under Sec. 52.51 or
Sec. 53.1242 of this chapter, as applicable, must be conducted under
the procedures of subpart O of this part.
(j) Proceedings on a Commission finding under Sec. 52.103(c) and
(g) or Sec. 53.1452(c) and (g) of this chapter, as applicable, shall
be conducted in accordance with the procedures designated by the
Commission in each proceeding.
* * * * *
0
16. In Sec. 2.329, revise paragraph (a) to read as follows:
Sec. 2.329 Prehearing conference.
(a) Necessity for prehearing conference; timing. The Commission or
the presiding officer may, and in the case of a proceeding on an
application for a construction permit or an operating license for a
facility of a type described in Sec. Sec. 50.21(b) or 50.22, or part
53 of this chapter, or a testing facility, must direct the parties or
their counsel to appear at a specified time and place for a conference
or conferences before trial. A prehearing conference in a proceeding
involving a construction permit or operating license for a facility of
a type described in Sec. Sec. 50.21(b) or 50.22 or part 53 of this
chapter must be held within sixty (60) days after discovery has been
completed or any other time specified by the Commission or the
presiding officer.
* * * * *
0
17. In Sec. 2.339, revise paragraph (d) to read as follows:
Sec. 2.339 Expedited decision-making procedure.
* * * * *
(d) The provisions of this section do not apply to an initial
decision directing the issuance of a limited work authorization under
10 CFR 50.10 or 10 CFR 53.1130; an early site permit under subpart A of
part 52 or under subpart H of part 53 of this chapter; a construction
permit or construction authorization under part 50 or part 53 of this
chapter; a combined license under subpart C of part 52 or under subpart
H of part 53 of this chapter; or a manufacturing license under subpart
F of part 52 or under subpart H of part 53.
0
18. In Sec. 2.340, revise paragraphs (b), (c), (d), (f), (i), and (j)
to read as follows:
Sec. 2.340 Initial decision in certain contested proceedings;
immediate effectiveness of initial decisions; issuance of
authorizations, permits and licenses.
* * * * *
(b) Initial decision--combined license under 10 CFR parts 52 or 53.
(1) Matters in controversy; presiding officer consideration of matters
not put in controversy by parties. In any initial decision in a
contested proceeding on an application for a combined license under
parts 52 or 53 of this chapter (including an amendment to or renewal of
combined license), the presiding officer shall make findings of fact
and conclusions of law on the matters put into controversy by the
parties and any matter designated by the Commission to be decided by
the presiding officer. The presiding officer shall also make findings
of fact and conclusions of law on any matter not put into controversy
by the parties, but only to the extent that the presiding officer
determines that a serious safety, environmental, or common defense and
security matter exists, and the Commission approves of an examination
of and decision on the matter upon its referral by the presiding
officer under, inter alia, the provisions of Sec. Sec. 2.323 and
2.341.
(2) Presiding officer initial decision and issuance of permit or
license.
(i) In a contested proceeding for the initial issuance or renewal
of a combined license under parts 52 or 53 of this chapter, or the
amendment of a combined license where the NRC has not made a
determination of no significant hazards consideration, the Commission
or the Director, Office of Nuclear Reactor Regulation, as appropriate
after making the requisite findings, shall issue, deny, or
appropriately condition the permit or license in accordance with the
presiding officer's initial decision once that decision becomes
effective.
(ii) In a contested proceeding for the amendment of a combined
license under parts 52 or 53 of this chapter where the NRC has made a
determination of no significant hazards consideration, the Commission
or the Director, Office of Nuclear Reactor Regulation, as appropriate
(appropriate official), after making the requisite findings and
complying with any applicable provisions of Sec. 2.1202(a) or Sec.
2.1403(a), may issue the amendment before the presiding officer's
initial decision becomes effective. Once the presiding officer's
initial decision becomes effective, the appropriate official shall take
action with respect to that amendment in accordance with the initial
decision. If the presiding officer's initial decision becomes effective
before the appropriate official issues the amendment, then the
appropriate official, after making the requisite findings, shall issue,
deny, or appropriately condition the amendment in accordance with the
presiding officer's initial decision.
(c) Initial decision on findings under 10 CFR 52.103 or 10 CFR
53.1452 with respect to acceptance criteria in nuclear power reactor
combined licenses. In any initial decision under Sec. 52.103(g) or
Sec. 53.1452(g) of this chapter with respect to whether acceptance
criteria have been or will be met, the presiding officer shall make
findings of fact and conclusions of law on the matters put into
controversy by the parties, and any matter designated by the Commission
to be decided by the presiding officer. Matters not put into
controversy by the parties but identified by the presiding officer as
matters requiring further examination, shall be referred to the
Commission for its determination; the Commission may, in its
discretion, treat any of these referred matters as a request for action
under Sec. 2.206 and process the matter in accordance with Sec.
52.103(f) or Sec. 53.1452(f) of this chapter.
(d) Initial decision--manufacturing license under 10 CFR parts 52
or 53. (1) Matters in controversy; presiding officer consideration of
matters not put in controversy by parties. In any initial decision in a
contested proceeding on an application for a manufacturing
[[Page 87021]]
license under subpart C of part 52 or subpart H of part 53 of this
chapter (including an amendment to or renewal of a manufacturing
license), the presiding officer shall make findings of fact and
conclusions of law on the matters put into controversy by the parties
and any matter designated by the Commission to be decided by the
presiding officer. The presiding officer also shall make findings of
fact and conclusions of law on any matter not put into controversy by
the parties, but only to the extent that the presiding officer
determines that a serious safety, environmental, or common defense and
security matter exists, and the Commission approves of an examination
of and decision on the matter upon its referral by the presiding
officer under, inter alia, the provisions of Sec. Sec. 2.323 and
2.341.
(2) Presiding officer initial decision and issuance of permit or
license.
(i) In a contested proceeding for the initial issuance or renewal
of a manufacturing license under subpart C of part 52 or subpart H of
part 53 of this chapter, or the amendment of a manufacturing license,
the Commission or the Director, Office of Nuclear Reactor Regulation,
as appropriate, after making the requisite findings, shall issue, deny,
or appropriately condition the permit or license in accordance with the
presiding officer's initial decision once that decision becomes
effective.
(ii) In a contested proceeding for the initial issuance or renewal
of a manufacturing license under subpart C of part 52 or subpart H of
part 53 of this chapter, or the amendment of a manufacturing license,
the Commission or the Director, Office of Nuclear Reactor Regulation,
as appropriate (appropriate official), may issue the license, permit,
or license amendment in accordance with Sec. 2.1202(a) or Sec.
2.1403(a) before the presiding officer's initial decision becomes
effective. If, however, the presiding officer's initial decision
becomes effective before the license, permit, or license amendment is
issued under Sec. 2.1202 or Sec. 2.1403, then the Commission or the
Director, Office of Nuclear Reactor Regulation, as appropriate, shall
issue, deny, or appropriately condition the license, permit, or license
amendment in accordance with the presiding officer's initial decision.
* * * * *
(f) Immediate effectiveness of certain presiding officer decisions.
A presiding officer's initial decision directing the issuance or
amendment of a limited work authorization under Sec. 50.10 or Sec.
53.1130 of this chapter; an early site permit under subpart A of part
52 or under subpart H of part 53 of this chapter; a construction permit
or construction authorization under part 50 or part 53 of this chapter;
an operating license under part 50 or part 53 of this chapter; a
combined license under subpart C of part 52 or subpart H or part 53 of
this chapter; a manufacturing license under subpart F of part 52 or
subpart H of part 53 of this chapter; a renewed license under part 53
or part 54 of this chapter; or a license under part 72 of this chapter
to store spent fuel in an independent spent fuel storage facility
(ISFSI) or a monitored retrievable storage installation (MRS); an
initial decision directing issuance of a license under part 61 of this
chapter; or an initial decision under Sec. 52.103(g) or Sec.
53.1452(g) of this chapter that acceptance criteria in a combined
license have been met, is immediately effective upon issuance unless
the presiding officer finds that good cause has been shown by a party
why the initial decision should not become immediately effective.
* * * * *
(i) Issuance of authorizations, permits, and licenses--production
and utilization facilities. The Commission or the Director, Office of
Nuclear Reactor Regulation, as appropriate, shall issue a limited work
authorization under Sec. 50.10 or Sec. 53.1130 of this chapter; an
early site permit under subpart A of part 52 or subpart H of part 53 of
this chapter; a construction permit or construction authorization under
part 50 or part 53 of this chapter; an operating license under part 50
or part 53 of this chapter; a combined license under subpart C of part
52 or part 53 of this chapter; or a manufacturing license under subpart
F of part 52 or part 53 of this chapter within 10 days from the date of
issuance of the initial decision:
(1) If the Commission or the Director has made all findings
necessary for issuance of the authorization, permit or license, not
within the scope of the initial decision of the presiding officer; and
(2) Notwithstanding the pendency of a petition for reconsideration
under Sec. 2.345, a petition for review under Sec. 2.341, or a motion
for stay under Sec. 2.342, or the filing of a petition under Sec.
2.206.
(j) Issuance of finding on acceptance criteria under 10 CFR 52.103
or 10 CFR 53.1452. The Commission or the Director, Office of Nuclear
Reactor Regulation, as appropriate, shall make the finding under Sec.
52.103(g) or Sec. 53.1452(g) of this chapter, that acceptance criteria
in a combined license are met within 10 days from the date of the
presiding officer's initial decision:
(1) If the Commission or the Director is otherwise able to make the
finding under Sec. 52.103(g) or Sec. 53.1452(g) of this chapter, that
the prescribed acceptance criteria are met for those acceptance
criteria not within the scope of the initial decision of the presiding
officer;
(2) If the presiding officer's initial decision--with respect to
contentions that the prescribed acceptance criteria have not been met--
finds that those acceptance criteria have been met, and the Commission
or the Director thereafter is able to make the finding that those
acceptance criteria are met;
(3) If the presiding officer's initial decision--with respect to
contentions that the prescribed acceptance criteria will not be met--
finds that those acceptance criteria will be met, and the Commission or
the Director thereafter is able to make the finding that those
acceptance criteria are met; and
(4) Notwithstanding the pendency of a petition for reconsideration
under Sec. 2.345, a petition for review under Sec. 2.341, or a motion
for stay under Sec. 2.342, or the filing of a petition under Sec.
2.206.
* * * * *
Sec. 2.341 [Amended]
0
19. In Sec. 2.341(a)(1), remove the phrase ``Sec. 52.103(c)'' and add
in its place the phrase ``Sec. 52.103(c) or Sec. 53.1452(c)''.
Sec. 2.400 [Amended]
0
20. In Sec. 2.400, remove the phrase ``parts 50 or 52'' and add in its
place the phrase ``part 50 or part 52, or Sec. 53.1470''.
0
21. In Sec. 2.401, revise the section heading and paragraph (a) to
read as follows:
Sec. 2.401 Notice of hearing on construction permit or combined
license applications pursuant to appendix N of 10 CFR parts 50, 52, or
53.
(a) In the case of applications under appendix N of part 50 or
Sec. 53.1470 of this chapter for construction permits for nuclear
power reactors of the type described in Sec. 50.22 or part 53 of this
chapter, or applications under appendix N of part 52 or Sec. 53.1470
of this chapter for combined licenses, the Secretary will issue notices
of hearing pursuant to Sec. 2.104.
* * * * *
0
22. In Sec. 2.402, revise paragraph (a) to read as follows:
Sec. 2.402 Separate hearings on separate issues; consolidation of
proceedings.
(a) In the case of applications under appendix N of part 50 or
Sec. 53.1470 of
[[Page 87022]]
this chapter for construction permits for nuclear power reactors of a
type described in 10 CFR 50.22 or part 53, or applications pursuant to
appendix N of part 52 or Sec. 53.1470 of this chapter for combined
licenses, the Commission or the presiding officer may order separate
hearings on particular phases of the proceeding, such as matters
related to the acceptability of the design of the reactor in the
context of the site parameters postulated for the design or
environmental matters.
* * * * *
Sec. 2.403 [Amended]
0
23. In Sec. 2.403, remove the phrase ``appendix N of part 50'' and add
in its place the phrase ``appendix N to part 50 or Sec. 53.1470''.
Sec. 2.404 [Amended]
0
24. In Sec. 2.404, remove the phrase ``appendix N of part 50'' and add
in its place the phrase ``appendix N to part 50 or Sec. 53.1470''.
Sec. 2.405 [Amended]
0
25. In Sec. 2.405, remove the phrase ``part 52'' and add in its place
the phrase ``part 52 or part 53''.
Sec. 2.406 [Amended]
0
26. In Sec. 2.406, remove the phrase ``appendix N of parts 50 or 52''
and add in its place the phrase ``appendix N to part 50 or part 52 or
Sec. 53.1470''.
Sec. 2.500 [Amended]
0
27. In Sec. 2.500, remove the phrase ``subpart F of part 52'' and add
in its place the phrase ``subpart F of part 52 or subpart H of part
53''.
0
28. In Sec. 2.501, revise the section heading and paragraph (a)
introductory text to read as follows:
Sec. 2.501 Notice of hearing on application under 10 CFR parts 52 or
53 for a license to manufacture nuclear power reactors.
(a) In the case of an application under subpart F of part 52 or
subpart H of part 53 of this chapter for a license to manufacture
nuclear power reactors of the type described in Sec. 50.22 or part 53
of this chapter to be operated at sites not identified in the license
application, the Secretary will issue a notice of hearing to be
published in the Federal Register at least 30 days before the date set
for hearing in the notice.\1\ The notice shall be issued as soon as
practicable after the application has been docketed. The notice will
state:
* * * * *
\1\ The thirty-day (30) requirement of this paragraph is not
applicable to a notice of the time and place of hearing published by
the presiding officer after notice of hearing described in this
section has been published.
0
29. In Sec. 2.643, revise paragraph (b) to read as follows:
Sec. 2.643 Acceptance and docketing of application for limited work
authorization.
* * * * *
(b) The Director will accept for docketing part one of an
application for a construction permit for a utilization facility which
is subject to Sec. 51.20(b) of this chapter and is of the type
specified in Sec. 50.21(b)(2) or (3) or Sec. 50.22 or part 53 of this
chapter or an application for a combined license where part one of the
application as described in Sec. 2.101(a)(9) is complete. Part one
will not be considered complete unless it contains the information
required by Sec. 50.10(d)(3) or Sec. 53.1130(a)(3) of this chapter.
Upon assignment of a docket number, the procedures in Sec. 2.101(a)(3)
and (4) relating to formal docketing and the submission and
distribution of additional copies of the application must be followed.
* * * * *
Sec. 2.645 [Amended]
0
30. In Sec. 2.645, in paragraph (a), remove the phrase ``Sec.
50.33(a) through (f) of this chapter'' and add in its place the phrase
``Sec. Sec. 50.33(a) through (f), 53.1109, and 53.1306(a) or 53.1413
of this chapter, as applicable,''.
Sec. 2.649 [Amended]
0
31. In Sec. 2.649, remove the phrase ``10 CFR 50.10(d)'' and add in
its place the phrase ``10 CFR 50.10(d) or 10 CFR 53.1130(a)''.
Sec. 2.800 [Amended]
0
32. In Sec. 2.800, amend paragraphs (c) and (d) by removing the phrase
``subpart B of part 52'' and adding in its place the phrase ``subpart B
of part 52 or subpart H of part 53''.
Sec. 2.801 [Amended]
0
33. In Sec. 2.801, remove the phrase ``subpart B of part 52'' and add
in its place the phrase ``subpart B of part 52 or subpart H of part
53''.
Sec. 2.813 [Amended]
0
34. In Sec. 2.813(a), remove the phrase ``parts 50, 52, and 100'' and
add in its place the phrase ``parts 50, 52, 53, and 100''.
Sec. 2.1103 [Amended]
0
35. In Sec. 2.1103, remove the phrase ``part 50 of this chapter'' and
add in its place the phrase ``parts 50 or 53 of this chapter''.
0
36. In Sec. 2.1202, revise paragraphs (a)(1) through (3) and (a)(6) to
read as follows:
Sec. 2.1202 Authority and role of NRC staff.
(a) * * *
(1) An application to construct and/or operate a production or
utilization facility (including an application for a limited work
authorization under Sec. Sec. 50.12 or 53.1130 of this chapter, or an
application for a combined license under subpart C of 10 CFR part 52,
or under subpart H of 10 CFR part 53;
(2) An application for an early site permit under subpart A of 10
CFR part 52 or under subpart H of 10 CFR part 53;
(3) An application for a manufacturing license under subpart F of
10 CFR part 52 or under subpart H of 10 CFR part 53;
* * * * *
(6) Production or utilization facility licensing actions that
involve significant hazards considerations as defined in Sec. Sec.
50.92 or 53.1520 of this chapter.
* * * * *
Sec. 2.1301 [Amended]
0
37. In Sec. 2.1301(b), remove ``part 50 and part 52'' and add in its
place ``parts 50, 52, and 53''.
Sec. 2.1403 [Amended]
0
38. In Sec. 2.1403, remove the phrase ``10 CFR 50.92'' and add in its
place the phrase ``10 CFR 50.92 or 10 CFR 53.1520''.
Sec. 2.1500 [Amended]
0
39. In Sec. 2.1500, remove the phrase ``subpart B of part 52'' and add
in its place the phrase ``subpart B of part 52 or under subpart H of
part 53''.
Sec. 2.1502 [Amended]
0
40. In Sec. 2.1502, in paragraph (a), remove the phrase ``Sec.
52.51(b)'' and add in its place the phrase ``Sec. Sec. 52.51(b) or
53.1242(b)(2)''; and in paragraph (b)(1), wherever it appears, remove
the phrase ``Sec. 52.51(a)'' and add in its place the phrase
``Sec. Sec. 52.51(a) or 53.1242(b)''.
PART 10--CRITERIA AND PROCEDURES FOR DETERMINING ELIGIBILITY FOR
ACCESS TO RESTRICTED DATA OR NATIONAL SECURITY INFORMATION OR AN
EMPLOYMENT CLEARANCE
0
41. The authority citation for part 10 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 145, 161 (42 U.S.C.
2165, 2201); Energy Reorganization Act of 1974, sec. 201 (42 U.S.C.
5841); E.O. 10450, 18 FR 2489, 3 CFR, 1949-1953 Comp., p. 936, as
amended; E.O. 10865, 25 FR 1583, 3 CFR, 1959-1963 Comp., p. 398, as
amended; E.O. 12968, 60 FR 40245, 3 CFR, 1995 Comp., p. 391.
Sec. 10.1 [Amended]
0
42. In Sec. 10.1, in paragraph (a)(3) remove the phrase ``under part
52'' and
[[Page 87023]]
add in its place the phrase ``under parts 52 or 53''.
Sec. 10.2 [Amended]
0
43. In Sec. 10.2, in paragraph (b), wherever it appears, remove the
phrase ``under part 52'' and add in its place the phrase ``under parts
52 or 53''.
PART 11--CRITERIA AND PROCEDURES FOR DETERMINING ELIGIBILITY FOR
ACCESS TO OR CONTROL OVER SPECIAL NUCLEAR MATERIAL
0
44. The authority citation for part 11 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 161, 223 (42 U.S.C.
2201, 2273); Energy Reorganization Act of 1974, sec. 201 (42 U.S.C.
5841); 44 U.S.C. 3504 note. Section 11.15(e) also issued under 31
U.S.C. 9701; 42 U.S.C. 2214.
Sec. 11.7 [Amended]
0
45. In Sec. 11.7, in the introductory text, remove the phrase ``parts
10, 25, 50, 70, 72, 73, and 95 of this chapter'' and add in its place
the phrase ``parts 10, 25, 50, 53, 70, 72, 73, and 95 of this
chapter''.
PART 19--NOTICES, INSTRUCTIONS AND REPORTS TO WORKERS: INSPECTION
AND INVESTIGATIONS
0
46. The authority citation for part 19 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 53, 63, 81, 103,
104, 161, 223, 234, 1701 (42 U.S.C. 2073, 2093, 2111, 2133, 2134,
2201, 2273, 2282, 2297f); Energy Reorganization Act of 1974, secs.
201, 211, 401 (42 U.S.C. 5841, 5851, 5891); 44 U.S.C. 3504 note.
0
47. In Sec. 19.2, revise paragraph (a) to read as follows:
Sec. 19.2 Scope.
(a) * * *
(1) All persons who receive, possess, use, or transfer material
licensed by the NRC under the regulations in parts 30 through 36, 39,
40, 60, 61, 63, 70, or 72 of this chapter, including persons licensed
to operate a production or utilization facility under part 50, part 52,
or part 53 of this chapter, persons licensed to possess power reactor
spent fuel in an independent spent fuel storage installation (ISFSI)
under part 72 of this chapter, and in accordance with 10 CFR 76.60 to
persons required to obtain a certificate of compliance or an approved
compliance plan under part 76 of this chapter;
(2) All applicants for and holders of licenses (including
construction permits and early site permits) under parts 50, 52, 53,
and 54 of this chapter;
(3) All applicants for and holders of a standard design approval
under subpart E of part 52 or under subpart H of part 53 of this
chapter; and
(4) All applicants for a standard design certification under
subpart B of part 52 or under subpart H of part 53 of this chapter, and
those (former) applicants whose designs have been certified under that
subpart.
* * * * *
0
48. In Sec. 19.3, revise the definitions for ``License'' and
``Regulated entities'' to read as follows:
Sec. 19.3 Definitions.
* * * * *
License means a license issued under the regulations in parts 30
through 36, 39, 40, 60, 61, 63, 70, or 72 of this chapter, including
licenses to manufacture, construct and/or operate a production or
utilization facility under parts 50, 52, 53, or 54 of this chapter.
* * * * *
Regulated entities means any individual, person, organization, or
corporation that is subject to the regulatory jurisdiction of the NRC,
including (but not limited to) an applicant for or holder of a standard
design approval under subpart E of part 52 or under subpart H of part
53 of this chapter or a standard design certification under subpart B
of part 52 or under subpart H of part 53 of this chapter.
* * * * *
Sec. 19.11 [Amended]
0
49. In Sec. 19.11, in paragraph (a) introductory text, paragraph (b)
introductory text, and paragraph (e)(1), remove the phrase ``of part
52'' wherever may appears and add in its place the phrase ``of part 52
or under subpart H of part 53''.
Sec. 19.14 [Amended]
0
50. In Sec. 19.14, in paragraph (a), wherever it may appear, remove
the phrase ``of part 52'' and add in its place the phrase ``of part 52
or under subpart H of part 53''.
Sec. 19.20 [Amended]
0
51. In Sec. 19.20, add the number ``53,'' in sequential order.
PART 20--STANDARDS FOR PROTECTION AGAINST RADIATION
0
52. The authority citation for part 20 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 11, 53, 63, 65, 81,
103, 104, 161, 170H, 182, 186, 223, 234, 274, 1701 (42 U.S.C. 2014,
2073, 2093, 2095, 2111, 2133, 2134, 2201, 2210h, 2232, 2236, 2273,
2282, 2021, 2297f); Energy Reorganization Act of 1974, secs. 201,
202 (42 U.S.C. 5841, 5842); Low-Level Radioactive Waste Policy
Amendments Act of 1985, sec. 2 (42 U.S.C. 2021b); 44 U.S.C. 3504
note.
Sec. 20.1002 [Amended]
0
53. In Sec. 20.1002, remove the phrase ``parts 30 through 36, 39, 40,
50, 52, 60, 61, 63, 70, or 72 of this chapter'' and add in its place
the phrase ``parts 30 through 36, 39, 40, 50, 52, 53, 60, 61, 63, 70,
or 72 of this chapter''.
0
54. In Sec. 20.1003, revise the definition for ``License'' to read as
follows:
Sec. 20.1003 Definitions.
* * * * *
License means a license issued under the regulations in parts 30
through 36, 39, 40, 50, 53, 60, 61, 63, 70, or 72 of this chapter.
* * * * *
Sec. 20.1101 [Amended]
0
55. In Sec. 20.1101, in paragraph (d), remove the phrase ``subject to
Sec. 50.34a'' and add in its place the phrase ``subject to Sec. Sec.
50.34a or 53.260 of this chapter''.
Sec. 20.1401 [Amended]
0
56. Amend Sec. 20.1401 by:
0
a. In paragraph (a), removing the phrase ``parts 30, 40, 50, 52, 60,
61, 63, 70, and 72 of this chapter'', and adding in its place the
phrase ``parts 30, 40, 50, 52, 53, 60, 61, 63, 70, and 72 of this
chapter''; and
0
b. In paragraphs (a) and (c) removing the phrase ``in accordance with
Sec. 50.83'' and adding in its place the phrase ``in accordance with
Sec. Sec. 50.83 or 53.1080''.
0
57. In Sec. 20.1403, revise paragraph (d) introductory text to read as
follows:
Sec. 20.1403 Criteria for license termination under restricted
conditions.
* * * * *
(d) The licensee has submitted a decommissioning plan or License
Termination Plan (LTP) to the Commission indicating the licensee's
intent to decommission in accordance with Sec. Sec. 30.36(d),
40.42(d), 50.82 (a) and (b), subpart G of part 53, 70.38(d), or 72.54
of this chapter, and specifying that the licensee intends to
decommission by restricting use of the site. The licensee shall
document in the LTP or decommissioning plan how the advice of
individuals and institutions in the community who may be affected by
the decommissioning has been sought and incorporated, as appropriate,
following analysis of that advice.
* * * * *
0
58. In Sec. 20.1404, revise paragraph (a)(4) introductory text to read
as follows:
Sec. 20.1404 Alternate criteria for license termination.
(a) * * *
[[Page 87024]]
(4) Has submitted a decommissioning plan or License Termination
Plan (LTP) to the Commission indicating the licensee's intent to
decommission in accordance with Sec. 30.36(d), 40.42(d), 50.82 (a) and
(b), subpart G of part 53, 70.38(d), or 72.54 of this chapter, and
specifying that the licensee proposes to decommission by use of
alternate criteria. The licensee shall document in the decommissioning
plan or LTP how the advice of individuals and institutions in the
community who may be affected by the decommissioning has been sought
and addressed, as appropriate, following analysis of that advice. In
seeking such advice, the licensee shall provide for:
* * * * *
Sec. 20.1406 [Amended]
0
59. In Sec. 20.1406, in paragraphs (a) and (b), wherever it appears,
remove the phrase ``under part 52'' and add in its place the phrase
``under parts 52 or 53''.
0
60. In Sec. 20.1501, revise paragraph (b) to read as follows:
Sec. 20.1501 General.
* * * * *
(b) Notwithstanding Sec. 20.2103(a) of this part, records from
surveys describing the location and amount of subsurface residual
radioactivity identified at the site must be kept with records
important for decommissioning, and such records must be retained in
accordance with Sec. 30.35(g), Sec. 40.36(f), Sec. 50.75(g), subpart
G of part 53, Sec. 70.25(g), or Sec. 72.30(d) of this chapter, as
applicable.
* * * * *
Sec. 20.1905 [Amended]
0
61. In Sec. 20.1905, in paragraph (g) introductory text, remove the
phrase ``Parts 50 or 52'' and add in its place the phrase ``parts 50,
52, or 53''.
0
62. In Sec. 20.2004, revise paragraph (b)(1) to read as follows:
Sec. 20.2004 Treatment or disposal by incineration.
* * * * *
(b)(1) Waste oils (petroleum derived or synthetic oils used
principally as lubricants, coolants, hydraulic or insulating fluids, or
metalworking oils) that have been radioactively contaminated in the
course of the operation or maintenance of a nuclear power reactor
licensed under parts 50 or 53 of this chapter may be incinerated on the
site where generated provided that the total radioactive effluents from
the facility, including the effluents from such incineration, conform
to the requirements of appendix I to part 50 or Sec. 53.425(d) of this
chapter and the effluent release limits contained in applicable license
conditions other than effluent limits specifically related to
incineration of waste oil. The licensee shall report any changes or
additions to the information supplied under Sec. Sec. 50.34, 50.34a,
or under subpart H of part 53 of this chapter associated with this
incineration pursuant to Sec. Sec. 50.71 or 53.1620 of this chapter,
as appropriate. The licensee shall also follow the procedures of
Sec. Sec. 50.59 or 53.1565 of this chapter with respect to such
changes to the facility or procedures.
* * * * *
0
63. In Sec. 20.2201, revise paragraphs (a)(2)(i), (b)(2)(i), and (c)
to read as follows:
Sec. 20.2201 Reports of theft or loss of licensed material.
(a) * * *
(2) * * *
(i) Licensees having an installed Emergency Notification System
shall make the reports to the NRC Operations Center under Sec. Sec.
50.72 or 53.1630 of this chapter, and
* * * * *
(b) * * *
(2) * * *
(i) For holders of an operating license for a nuclear power plant,
the events included in paragraph (b) of this section must be reported
under the procedures described in Sec. Sec. 50.73(b), (c), (d), (e),
and (g) or 53.1640(b), (c), (d), and (e) of this chapter and must
include the information required in paragraph (b)(1) of this section,
and
* * * * *
(c) A duplicate report is not required under paragraph (b) of this
section if the licensee is also required to submit a report pursuant to
Sec. Sec. 30.55(c), 37.57, 37.81, 40.64(c), 50.72, 50.73, 53.1630,
53.1640, 70.52, 73.27(b), 73.67(e)(3)(vii), 73.67(g)(3)(iii), 73.1205,
or 150.19(c) of this chapter.
* * * * *
Sec. 20.2202 [Amended]
0
64. In Sec. 20.2202, in paragraph (d)(1), remove the phrase ``10 CFR
50.72'' and add in its place the phrase ``Sec. Sec. 50.72 or 53.1630
of this chapter;''.
0
65. In Sec. 20.2203, revise paragraph (c) to read as follows:
Sec. 20.2203 Reports of exposures, radiation levels, and
concentrations of radioactive material exceeding the constraints or
limits.
* * * * *
(c) For holders of an operating license or a combined license for a
nuclear power plant, the occurrences included in paragraph (a) of this
section must be reported under the procedures described in Sec. Sec.
50.73(b), (c), (d), (e), and (g) or 53.1640(b), (c), (d), and (e) of
this chapter, and must include the information required by paragraph
(b) of this section. Occurrences reported under Sec. Sec. 50.73 or
53.1640 of this chapter need not be reported by a duplicate report
under paragraph (a) of this section.
* * * * *
Sec. 20.2206 [Amended]
0
66. In Sec. 20.2206, in paragraph (a)(1), remove the phrase ``or Sec.
50.22'' and add in its place the phrase ``, Sec. 50.22, or part 53''.
PART 21--REPORTING OF DEFECTS AND NONCOMPLIANCE
0
67. The authority citation for part 21 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 53, 63, 81, 103,
104, 161, 223, 234, 1701 (42 U.S.C. 2073, 2093, 2111, 2133, 2134,
2201, 2273, 2282, 2297f); Energy Reorganization Act of 1974, secs.
201, 206 (42 U.S.C. 5841, 5846); Nuclear Waste Policy Act of 1982,
secs. 135, 141 (42 U.S.C. 10155, 10161); 44 U.S.C. 3504 note.
0
68. In Sec. 21.2, revise paragraphs (a)(2) through (4), (b), and (c)
to read as follows:
Sec. 21.2 Scope.
(a) * * *
(1) * * *
(2) Each individual, corporation, partnership, or other entity
doing business within the United States, and each director and
responsible officer of such an organization, that constructs a
production or utilization facility licensed for manufacture,
construction, or operation under parts 50, 52, or 53 of this chapter,
an ISFSI for the storage of spent fuel licensed under part 72 of this
chapter, an MRS for the storage of spent fuel or high-level radioactive
waste under part 72 of this chapter, or a geologic repository for the
disposal of high-level radioactive waste under parts 60 or 63 of this
chapter; or supplies basic components for a facility or activity
licensed, other than for export, under parts 30, 40, 50, 52, 53, 60,
61, 63, 70, 71, or 72 of this chapter;
(3) Each individual, corporation, partnership, or other entity
doing business within the United States, and each director and
responsible officer of such an organization, applying for a design
certification rule under parts 52 or 53 of this chapter; or supplying
basic components with respect to that design certification, and each
individual, corporation, partnership, or other entity doing business
within the United States, and each director and responsible officer of
such an organization, whose
[[Page 87025]]
application for design certification has been granted under parts 52 or
53 of this chapter, or who has supplied or is supplying basic
components with respect to that design certification;
(4) Each individual, corporation, partnership, or other entity
doing business within the United States, and each director and
responsible officer of such an organization, applying for or holding a
standard design approval under parts 52 or 53 of this chapter; or
supplying basic components with respect to a standard design approval
under parts 52 or 53 of this chapter;
(b) For persons licensed to construct a facility under either a
construction permit issued under Sec. Sec. 50.23 or 53.1333 of this
chapter or a combined license under parts 52 or 53 of this chapter (for
the period of construction until the date that the Commission makes the
finding under Sec. Sec. 52.103(g) or 53.1452(g) of this chapter), or
to manufacture a facility under parts 52 or 53 of this chapter,
evaluation of potential defects and failures to comply and reporting of
defects and failures to comply under Sec. Sec. 50.55(e) or 53.605 of
this chapter satisfies each person's evaluation, notification, and
reporting obligation to report defects and failures to comply under
this part and the responsibility of individual directors and
responsible officers of these licensees to report defects under Section
206 of the Energy Reorganization Act of 1974.
(c) For persons licensed to operate a nuclear power plant under
part 50, part 52, or part 53 of this chapter, evaluation of potential
defects and appropriate reporting of defects under Sec. Sec. 50.72,
50.73, 53.1630, 53.1640, or 73.1200 and 73.1205 of this chapter,
satisfies each person's evaluation, notification, and reporting
obligation to report defects under this part, and the responsibility of
individual directors and responsible officers of these licensees to
report defects under Section 206 of the Energy Reorganization Act of
1974.
* * * * *
0
69. In Sec. 21.3, revise the definitions for ``Basic component'',
``Commercial grade item'', ``Critical characteristics'', ``Dedicating
entity'', ``Dedication'', ``Defect'', and ``Substantial safety hazard''
to read as follows:
Sec. 21.3 Definitions.
* * * * *
Basic component. (1)(i) When applied to nuclear power plants
licensed under part 53 of this chapter, basic component means a safety-
related structure, system, or component (SSC), or part thereof, and
when applied to nuclear power plants licensed under parts 50 or 52, of
this chapter, basic component means an SSC, or part thereof that
affects its safety function necessary to assure:
(A) The integrity of the reactor coolant pressure boundary;
(B) The capability to shut down the reactor and maintain it in a
safe-shutdown condition; or
(C) The capability to prevent or mitigate the consequences of
accidents which could result in potential offsite exposures comparable
to those referred to in Sec. Sec. 50.34(a)(1), 50.67(b)(2), or 100.11
of this chapter, as applicable.
(ii) Basic components are items designed and manufactured under a
quality assurance program complying with appendix B to part 50 of this
chapter, or commercial grade items which have successfully completed
the dedication process.
(2) When applied to standard design certifications and approvals
under part 53 of this chapter, basic component means the design or
procurement information approved or to be approved within the scope of
the design certification or approval for a safety-related SSC, or part
thereof. When applied to standard design certifications under subpart B
of part 52 of this chapter and standard design approvals under part 52
of this chapter, basic component means the design or procurement
information approved or to be approved within the scope of the design
certification or approval for an SSC, or part thereof, that affects its
safety function necessary to assure:
(i) The integrity of the reactor coolant pressure boundary;
(ii) The capability to shut down the reactor and maintain it in a
safe-shutdown condition; or
(iii) The capability to prevent or mitigate the consequences of
accidents which could result in potential offsite exposures comparable
to those referred to in Sec. Sec. 50.34(a)(1), 50.67(b)(2), or 100.11
of this chapter, as applicable.
(3) When applied to other facilities and other activities licensed
under 10 CFR parts 30, 40, 50 (other than nuclear power plants), 60,
61, 63, 70, 71, or 72 of this chapter, basic component means a
structure, system, or component, or part thereof, that affects their
safety function, that is directly procured by the licensee of a
facility or activity subject to the regulations in this part and in
which a defect or failure to comply with any applicable regulation in
this chapter, order, or license issued by the Commission could create a
substantial safety hazard.
(4) In all cases, basic component includes safety-related design,
analysis, inspection, testing, fabrication, replacement of parts, or
consulting services that are associated with the component hardware,
design certification, design approval, or information in support of an
early site permit application under part 52 or part 53 of this chapter,
whether these services are performed by the component supplier or
others.
Commercial grade item. (1) When applied to nuclear power plants
licensed under parts 50 or 53 of this chapter, commercial grade item
means an SSC, or part thereof that affects its safety function, that
was not designed and manufactured as a basic component. Commercial
grade items do not include items where the design and manufacturing
process require in-process inspections and verifications to ensure that
defects or failures to comply are identified and corrected (i.e., one
or more critical characteristics of the item cannot be verified).
(2) When applied to facilities and activities licensed pursuant to
parts 30, 40, 50 (other than nuclear power plants), 60, 61, 63, 70, 71,
or 72 of this chapter, commercial grade item means an item that is:
(i) Not subject to design or specification requirements that are
unique to those facilities or activities;
(ii) Used in applications other than those facilities or
activities; and
(iii) To be ordered from the manufacturer/supplier on the basis of
specifications set forth in the manufacturer's published product
description (for example, a catalog).
* * * * *
Critical characteristics. When applied to nuclear power plants
licensed under parts 50, 52, or 53 of this chapter, critical
characteristics are those important design, material, and performance
characteristics of a commercial grade item that, once verified, will
provide reasonable assurance that the item will perform its intended
safety function.
Dedicating entity. When applied to nuclear power plants licensed
under parts 50, 52, or 53 of this chapter, dedicating entity means the
organization that performs the dedication process. Dedication may be
performed by the manufacturer of the item, a third-party dedicating
entity, or the licensee itself. The dedicating entity, under Sec.
21.21(c) of this part, is responsible for identifying and evaluating
deviations, reporting defects and failures to comply for the dedicated
item, and maintaining auditable records of the dedication process.
Dedication. (1) When applied to nuclear power plants licensed
pursuant to 10 CFR parts 30, 40, 50, 53, or 60, dedication is an
acceptance process
[[Page 87026]]
undertaken to provide reasonable assurance that a commercial grade item
to be used as a basic component will perform its intended safety
function and, in this respect, is deemed equivalent to an item designed
and manufactured under a 10 CFR part 50, appendix B, quality assurance
program. This assurance is achieved by identifying the critical
characteristics of the item and verifying their acceptability by
inspections, tests, or analyses performed by the purchaser or third-
party dedicating entity after delivery, supplemented as necessary by
one or more of the following: commercial grade surveys; product
inspections or witness at holdpoints at the manufacturer's facility,
and analysis of historical records for acceptable performance. In all
cases, the dedication process must be conducted under the applicable
provisions of 10 CFR part 50, appendix B. The process is considered
complete when the item is designated for use as a basic component.
(2) When applied to facilities and activities licensed pursuant to
10 CFR parts 30, 40, 50 (other than nuclear power plants), 60, 61, 63,
70, 71, or 72, dedication occurs after receipt when that item is
designated for use as a basic component.
Defect means:
(1) A deviation in a basic component delivered to a purchaser for
use in a facility or an activity subject to the regulations in this
part if, on the basis of an evaluation, the deviation could create a
substantial safety hazard;
(2) The installation, use, or operation of a basic component
containing a defect as defined in this section;
(3) A deviation in a portion of a facility subject to the early
site permit, standard design certification, standard design approval,
construction permit, combined license or manufacturing licensing
requirements of parts 50, 52, or 53 of this chapter, provided the
deviation could, on the basis of an evaluation, create a substantial
safety hazard and the portion of the facility containing the deviation
has been offered to the purchaser for acceptance;
(4) A condition or circumstance involving a basic component that
could contribute to the exceeding of a safety limit, as defined in the
technical specifications of a license for operation issued under part
50, part 52, or part 53 of this chapter; or
(5) An error, omission or other circumstance in a design
certification, or standard design approval that, on the basis of an
evaluation, could create a substantial safety hazard.
* * * * *
Substantial safety hazard means a loss of safety function to the
extent that there is a major reduction in the degree of protection
provided to public health and safety for any facility or activity
licensed or otherwise approved or regulated by the NRC, other than for
export, under part 30, 40, 50, 52, 53, 60, 61, 63, 70, 71, or 72 of
this chapter.
* * * * *
Sec. 21.21 [Amended]
0
70. Amend Sec. 21.21 by:
0
a. In paragraph (a)(3), removing the phrase ``under part 52'' and add
in its place the phrase ``under parts 52 or 53''; and
0
b. In paragraphs (d)(1)(i) and (ii) removing the phrase ``parts 30, 40,
50, 52, 60, 61, 63, 70, 71, or 72 of this chapter'' and adding in its
place the phrase ``parts 30, 40, 50, 52, 53, 60, 61, 63, 70, 71, or 72
of this chapter''.
Sec. 21.51 [Amended]
0
71. In Sec. 21.51, in paragraphs (a)(4) and (5) remove the phrase ``of
part 52'' and add in its place the phrase ``of part 52 or under subpart
H of part 53''.
Sec. 21.61 [Amended]
0
72. In Sec. 21.61, in paragraph (b) remove the phrase ``under part
52'' and add in its place the phrase ``under parts 52 or 53''.
PART 25--ACCESS AUTHORIZATION
0
73. The authority citation for part 25 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 145, 161, 223, 234
(42 U.S.C. 2165, 2201, 2273, 2282); Energy Reorganization Act of
1974, sec. 201 (42 U.S.C. 5841); 44 U.S.C. 3504 note; E.O. 10865, 25
FR 1583, as amended, 3 CFR, 1959-1963 Comp., p. 398; E.O. 12829, 58
FR 3479, 3 CFR, 1993 Comp., p. 570; E.O. 13526, 75 FR 707, 3 CFR,
2009 Comp., p. 298; E.O. 12968, 60 FR 40245, 3 CFR, 1995 Comp., p.
391. Section 25.17(f) and Appendix A also issued under 31 U.S.C.
9701; 42 U.S.C. 2214.
0
74. In Sec. 25.5, revise the definition for ``License'' to read as
follows:
Sec. 25.5 Definitions.
* * * * *
License means a license issued pursuant to 10 CFR parts 50, 52, 53,
60, 63, 70, or 72.
* * * * *
Sec. 25.17 [Amended]
0
75. In Sec. 25.17, in paragraph (a), remove the phrase ``under 10 CFR
parts 50, 52, 54, 60, 63, 70, 72, or 76'' and add in its place the
phrase ``under 10 CFR parts 50, 52, 53, 54, 60, 63, 70, 72, or 76''.
Sec. 25.35 [Amended]
0
76. In Sec. 25.35, in paragraph (a), wherever it appears, remove the
phrase ``under part 52'' and add in its place the phrase ``under parts
52 or 53''.
PART 26--FITNESS FOR DUTY PROGRAMS
0
77. The authority citation for part 26 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 53, 103, 104, 107,
161, 223, 234, 1701 (42 U.S.C. 2073, 2133, 2134, 2137, 2201, 2273,
2282, 2297f); Energy Reorganization Act of 1974, secs. 201, 202 (42
U.S.C. 5841, 5842); 44 U.S.C. 3504 note.
0
78. In Sec. 26.3, revise paragraph (d) and add paragraph (f) to read
as follows:
Sec. 26.3 Scope.
* * * * *
(d) Contractor/vendors (C/Vs) who implement FFD programs or program
elements, to the extent that the licensees and other entities specified
in paragraphs (a) through (c) and (f) of this section rely on those C/V
FFD programs or program elements to meet the requirements of this part,
shall comply with the requirements of this part.
* * * * *
(f) No later than the start of construction activities, licensees
and other entities that have applied for or have been issued a license
under part 53 of this chapter, other than a manufacturing license (ML),
must implement the requirements in subpart M of this part or all the
requirements of this part except subparts K and M. Holders of an ML
under part 53 of this chapter must implement the requirements in
subpart M or all the requirements of this part except subparts K and M,
before commencing activities that assemble a manufactured reactor.
0
79. In Sec. 26.4, revise paragraphs (a) introductory text, (a)(1),
(a)(4), (b), (c), (e) introductory text, (e)(4), (f), (g) introductory
text, and (h) to read as follows:
Sec. 26.4 FFD program applicability to categories of individuals.
(a) All persons who are granted unescorted access to nuclear power
reactor protected areas by the licensees in Sec. 26.3(a) and, as
applicable, (c) and perform the following duties shall be subject to an
FFD program that meets all of the requirements of this part, except
subpart K of this part, and those persons who are granted unescorted
access to either nuclear power reactor protected areas or remote
facilities where safety-significant systems or components may be
operated within the design basis of
[[Page 87027]]
a licensed commercial nuclear plant, by the licensees and other
entities in Sec. 26.3(f) and perform the following duties must be
subject to an FFD program that satisfies the requirements in subpart M
of this part, unless the licensee or other entity subjects these
individuals to an FFD program that satisfies all of the requirements of
this part except for those requirements in subparts K and M:
(1) For persons who are granted unescorted access by the licensees
in Sec. 26.3(a) and, as applicable, (c), operating or onsite directing
of the operation of systems and components that a risk-informed
evaluation process has shown to be significant to public health and
safety; for those persons who are granted unescorted access by the
licensees and other entities in Sec. 26.3(f), operating or directing
of the operation of systems and components that a risk-informed
evaluation process has shown to be significant to public health and
safety;
* * * * *
(4) For persons who are granted unescorted access to nuclear power
reactor protected areas by the licensees in Sec. 26.3(a) and, as
applicable, (c), performing maintenance or onsite directing of the
maintenance of SSCs that a risk-informed evaluation process has shown
to be significant to public health and safety; for those persons who
are granted unescorted access to nuclear power reactor protected areas
by the licensees and other entities in Sec. 26.3(f), performing
maintenance or directing of the maintenance of SSCs that a risk-
informed evaluation process has shown to be significant to public
health and safety; and
* * * * *
(b) All persons who are granted unescorted access to nuclear power
reactor protected areas by the licensees in Sec. 26.3(a) and, as
applicable, (c) and who do not perform the duties described in
paragraph (a) of this section shall be subject to an FFD program that
meets all of the requirements of this part, except Sec. Sec. 26.205
through 26.209 and subpart K of this part. All persons who are granted
unescorted access to a facility licensed under part 53 of this chapter,
and who do not perform or direct the performance of the duties
described in Sec. 26.4(a), must be subject to the requirements in
subpart M of this part, unless the licensee or other entity implements
an FFD program that satisfies all of the requirements of this part,
except Sec. Sec. 26.205 through 26.209 and subparts K and M.
(c) All persons who are required by a licensee in Sec. 26.3(a)
and, as applicable, (c) to physically report to the licensee's
Technical Support Center or Emergency Operations Facility by licensee
emergency plans and procedures shall be subject to an FFD program that
meets all of the requirements of this part, except Sec. Sec. 26.205
through 26.209 and subpart K of this part. Also, for licensees or other
entities in Sec. 26.3(f), all persons without unescorted access to the
facility who make decisions and/or direct actions regarding plant
safety and security, and all persons who participate remotely in
emergency response activities or physically report to the Technical
Support Center or Emergency Operations Facility (or an equivalent
facility), must be subject to an FFD program that satisfies all of the
requirements described in subpart M of this part, unless the licensee
or other entity implements an FFD program that satisfies all of the
requirements of this part, except Sec. Sec. 26.205 through 26.209 and
subparts K and M.
* * * * *
(e) When construction activities, as defined in Sec. 26.5, begin,
any individual whose duties for the licensees and other entities in
Sec. 26.3(c) require him or her to have the following types of access
or perform the following activities at the location where the nuclear
power plant will be constructed and operated shall be subject to an FFD
program that meets all of the requirements of this part, except
subparts I, K, and M of this part, and for any individual whose duties
for the licensees and other entities in Sec. 26.3(f) require him or
her to have the following types of access, perform construction
activities as defined in Sec. 26.5, or perform the following
activities must be subject to an FFD program as described in subpart M
or an FFD program that satisfies all of the requirements of this part,
except subparts I, K, and M:
* * * * *
(4) Witnesses or determines inspections, tests, and analyses
certification required under part 52 or part 53 of this chapter;
* * * * *
(f) Any individual who is constructing or directing the
construction of safety- or security-related SSCs shall be subject to an
FFD program that meets the requirements of subpart K, or, if
applicable, subpart M of this part, unless the licensee or other entity
subjects these individuals to an FFD program that meets all of the
requirements of this part, except for subparts I, K, and M of this
part.
(g) All FFD program personnel who are involved in the day-to-day
operations of the program, as defined by the procedures of the
licensees and other entities in Sec. 26.3(a) through (c), and, as
applicable, (d) and whose duties require them to have the following
types of access or perform the following activities shall be subject to
an FFD program that meets all of the requirements of this part, except
subparts I, K, and M of this part, and, at the licensee's or other
entity's discretion, subpart C of this part. All personnel whose duties
require them to have the following types of access or perform the
following activities at facilities licensed under part 53 of this
chapter must be subject to the requirements in subpart M or an FFD
program that satisfies all of the requirements of this part, except
subparts I, K, and M, and, at the licensee's or other entity's
discretion, subpart C of this part:
* * * * *
(h) Individuals who have applied for authorization to have the
types of access or perform the activities described in paragraphs (a)
through (d) of this section shall be subject to Sec. Sec. 26.31(c)(1),
26.35(b), 26.37, 26.39, and the applicable requirements of subparts C,
E through H, and M of this part.
* * * * *
0
80. Amend Sec. 26.5 by:
0
a. Adding the definitions for ``Biological marker'' and ``Change'';
0
b. Revising the definitions for ``Constructing or construction
activities'' ``Contractor/vendor (C/V)'';
0
c. Adding the definition of ``Illicit substance'';
0
d. Revising the definitions of ``Other entity'' and ``Questionable
validity'';
0
e. Adding the definitions of ``Reduction in FFD program
effectiveness'';
0
f. Revising the definitions of ``Reviewing official'', ``Safety-related
structures, systems, and components (SSCs)'', and ``Security-related
SSCs'';
0
g. Adding the definitions of ``Special nuclear material''; and
0
h. Revising the definition of ``Unit outage''.
The additions and revisions read as follows:
Sec. 26.5 Definitions.
* * * * *
Biological marker means, for a part 53 licensee implementing
subpart M of this part, an endogenous substance that is used to
validate that the biological specimen collected for testing was
produced by the donor.
* * * * *
[[Page 87028]]
Change as used in Sec. 26.603(e) means an action that results in a
modification of, addition to, or removal from the licensee's or other
entity's FFD program.
* * * * *
Constructing or construction activities means, for the purposes of
this part, the tasks involved in building a nuclear power plant that
are performed at the location where the nuclear power plant will be
constructed and operated. These tasks include fabricating, erecting,
integrating, and testing safety- and security-related SSCs, and the
installation of their foundations, including the placement of concrete.
For a licensee or other entity described in Sec. 26.3(f), construction
is defined in Sec. 53.020 of this chapter.
Contractor/vendor (C/V) means any company, or any individual not
employed by a licensee or other entity specified in Sec. 26.3(a)
through (c) and (f), who is providing work or services to a licensee or
other entity covered in Sec. 26.3(a) through (c) and (f), either by
contract, purchase order, oral agreement, or other arrangement.
* * * * *
Illicit substance means a substance that causes impairment and
possible addiction but is not an illegal drug as defined in Sec. 26.5.
* * * * *
Other entity means any corporation, firm, partnership, limited
liability company, association, C/V, or other organization who is
subject to this part under Sec. 26.3(a) through (c) and (f) but is not
licensed by the NRC.
* * * * *
Questionable validity means the results of validity screening or
initial validity tests at a licensee testing facility indicating that a
urine specimen may be adulterated, substituted, dilute, or invalid. For
a part 53 licensee or other entity, questionable validity means the
results of validity screening or initial validity tests indicating that
a biological specimen obtained from an individual pursuant to subpart M
of this part may be adulterated, substituted, dilute, or invalid.
Reduction in FFD program effectiveness means, for a part 53
licensee or other entity implementing subpart M of this part, a change
or series of changes to an element of the FFD program that reduces or
eliminates the licensee's ability to satisfy or maintain site-specific
FFD program performance when compared to historical site-specific
performance, the licensee's fleet-level program performance, or
industry performance.
* * * * *
Reviewing official means an employee of a licensee or other entity
specified in Sec. 26.3(a) through (c) and (f), who is designated by
the licensee or other entity to be responsible for reviewing and
evaluating any potentially disqualifying FFD information about an
individual, including, but not limited to, the results of a
determination of fitness, as defined in Sec. 26.189, in order to
determine whether the individual may be granted or maintain
authorization.
Safety-related structures, systems, and components (SSCs) means,
for part 50 or part 52 licensees and other entities described in Sec.
26.3(a) through (d), those SSCs that are relied on to remain functional
during and following design basis events to ensure the integrity of the
reactor coolant pressure boundary, the capability to shut down the
reactor and maintain it in a safe shutdown condition, or the capability
to prevent or mitigate the consequences of accidents that could result
in potential offsite exposure comparable to the guidelines in Sec.
50.34(a)(1) of this chapter. For part 53 licensees and other entities
described in Sec. 26.3(d) and (f), safety-related has the same meaning
as that in Sec. 53.020 of this chapter.
Security-related SSCs means, for the purposes of this part, those
structures, systems, and components that the licensee will rely on to
implement the licensee's physical security and safeguards contingency
plans that either are required under part 73 of this chapter if the
licensee is a construction permit applicant or holder or an early site
permit holder, as described in Sec. 26.3(c)(3) through (c)(5),
respectively, or are included in the licensee's application if the
licensee is a combined license applicant or holder, as described in
Sec. 26.3(c)(1) and (c)(2), respectively, or a licensee or other
entity described in Sec. 26.3(d) or (f).
* * * * *
Special nuclear material (SNM) has the same meaning as that in
Sec. 70.4 of this chapter.
* * * * *
Unit outage means, for the purposes of this part, for electricity-
generation units, that the reactor unit is disconnected from the
electrical grid. Unit outage means, for the purposes of this part, for
non-electricity-generation units, that the reactor unit is disconnected
from the loads to which its output is supplied under normal operating
conditions.
* * * * *
0
81. In Sec. 26.8, revise paragraph (b) to read as follows:
Sec. 26.8 Information collection requirements: OMB approval.
* * * * *
(b) The approved information collection requirements contained in
this part appear in Sec. Sec. 26.9, 26.27, 26.29, 26.31, 26.33, 26.35,
26.37, 26.39, 26.41, 26.53, 26.55, 26.57, 26.59, 26.61, 26.63, 26.65,
26.67, 26.69, 26.75, 26.77, 26.85, 26.87, 26.89, 26.91, 26.93, 26.95,
26.97, 26.99, 26.101, 26.103, 26.107, 26.109, 26.111, 26.113, 26.115,
26.117, 26.119, 26.125, 26.127, 26.129, 26.135, 26.137, 26.139, 26.153,
26.157, 26.159, 26.163, 26.165, 26.167, 26.168, 26.169, 26.183, 26.185,
26.187, 26.189, 26.202, 26.203, 26.205, 26.207, 26.211, 26.401, 26.403,
26.405, 26.406, 26.407, 26.411, 26.413, 26.415, 26.417, 26.603, 26.604,
26.605, 26.606, 26.607, 26.608, 26.609, 26.611, 26.613, 26.617, 26.619,
26.711, 26.713, 26.715, 26.717, 26.719, and 26.821.
0
82. Revise Sec. 26.21 to read as follows:
Sec. 26.21 Fitness-for-duty program.
The licensees and other entities specified in Sec. 26.3(a) through
(c) and (f) (for those licensees and other entities that do not
implement the requirements in subparts M and K of this part) shall
establish, implement, and maintain FFD programs that, at a minimum,
comprise the program elements contained in this subpart. The
individuals specified in Sec. 26.4(a) through (e) and (g), and, at the
licensee's or other entity's discretion, Sec. 26.4(f), and, if
necessary, Sec. 26.4(j) shall be subject to these FFD programs.
Licensees and other entities may rely on the FFD program or program
elements of a C/V, as defined in Sec. 26.5, if the C/V's FFD program
or program elements satisfy the applicable requirements of this part.
0
83. Revise Sec. 26.51 to read as follows:
Sec. 26.51 Applicability.
The requirements in this subpart apply to the licensees and other
entities identified in Sec. 26.3(a), (b), and, as applicable, (c) for
the categories of individuals in Sec. 26.4(a) through (d), and, at the
licensee's or other entity's discretion, in Sec. 26.4(g) and, if
necessary, Sec. 26.4(j). The requirements in this subpart also apply
to the licensees and other entities specified in Sec. 26.3(c), as
applicable, for the categories of individuals in Sec. 26.4(e). At the
discretion of a licensee or other entity in Sec. 26.3(c), the
requirements of this subpart also may be applied to the categories of
individuals identified in Sec. 26.4(f). In addition, the requirements
in this subpart apply to the entities in Sec. 26.3(d) to the extent
that a licensee or other entity relies on the C/V to satisfy the
requirements of this subpart. Certain requirements in this subpart also
apply
[[Page 87029]]
to the individuals specified in Sec. 26.4(h). The requirements in this
subpart apply to the FFD programs of licensees and other entities
identified in Sec. 26.3(f) that elect not to implement the
requirements in subpart M for the categories of individuals in Sec.
26.4 and those licensees and other entities that elect to implement the
requirements in Sec. 26.605.
Sec. 26.53 [Amended]
0
84. Amend Sec. 26.53 by:
0
a. In paragraph (e), wherever it appears, remove the phrase ``Sec.
26.3(a) through (c)'' and add in its place the phrase ``Sec. 26.3(a)
through (c) and (f)''; and
0
b. In paragraphs (g), (h), and (i), wherever it appears, remove the
phrase ``(c) and (d)'' and add in its place the phrase ``(c), (d), and
(f)''.
Sec. 26.63 [Amended]
0
85. In Sec. 26.63, in paragraph (d) remove the phrase ``Sec. 26.3(a)
through (d)'' and add in its place the phrase ``Sec. 26.3(a) through
(d) and (f)''.
0
86. Revise Sec. 26.73 to read as follows:
Sec. 26.73 Applicability.
The requirements in this subpart apply to the licensees and other
entities identified in Sec. 26.3(a), (b), and, as applicable, (c) for
the categories of individuals specified in Sec. 26.4(a) through (d)
and (g). The requirements in this subpart also apply to the licensees
and other entities specified in Sec. 26.3(c), as applicable, for the
categories of individuals in Sec. 26.4(e). At the discretion of a
licensee or other entity in Sec. 26.3(c), the requirements of this
subpart also may be applied to the categories of individuals identified
in Sec. 26.4(f). In addition, the requirements in this subpart apply
to the entities in Sec. 26.3(d) to the extent that a licensee or other
entity relies on the C/V to satisfy the requirements of this subpart.
The regulations in this subpart also apply to the individuals specified
in Sec. 26.4(h) and (j), as appropriate. The requirements in this
subpart apply to the FFD programs of licensees and other entities
identified in Sec. 26.3(f) that elect not to implement the
requirements in subpart M for the categories of individuals in Sec.
26.4 and those licensees and other entities that elect to implement the
requirements in Sec. 26.605(b).
0
87. Revise Sec. 26.81 to read as follows:
Sec. 26.81 Purpose and applicability.
This subpart contains requirements for collecting specimens for
drug testing and conducting alcohol tests by or on behalf of the
licensees and other entities in Sec. 26.3(a) through (d) for the
categories of individuals specified in Sec. 26.4(a) through (d) and
(g). At the discretion of a licensee or other entity in Sec. 26.3(c),
specimen collections and alcohol tests must be conducted either under
this subpart for the individuals specified in Sec. 26.4(e) and (f) or
the licensee or other entity may rely on specimen collections and
alcohol tests conducted under the requirements of 49 CFR part 40 for
the individuals specified in Sec. 26.4(e) and (f). The requirements of
this subpart do not apply to specimen collections and alcohol tests
that are conducted under the requirements of 49 CFR part 40, as
permitted in this paragraph and under Sec. Sec. 26.4(j) and
26.31(b)(2) and subpart K. The requirements in this subpart apply to
the FFD programs of licensees and other entities identified in Sec.
26.3(f) that elect not to implement the requirements in subpart M for
the categories of individuals in Sec. 26.4 and those licensees and
other entities that elect to implement the requirements in Sec.
26.605.
0
88. Revise Sec. 26.201 to read as follows:
Sec. 26.201 Applicability.
(a) The requirements in this subpart, with the exception of Sec.
26.202, apply to the licensees and other entities identified in Sec.
26.3(a); if applicable, (c), (d), and (f), for licensees and other
entities not implementing the requirements in subparts K and M. For the
licensees and other entities to whom the requirements in this subpart,
with the exception of Sec. 26.202, apply, the requirements in
Sec. Sec. 26.203 and 26.211 apply to the individuals identified in
Sec. 26.4(a) through (c). In addition, the requirements in Sec. Sec.
26.205 through 26.209 apply to the individuals identified in Sec.
26.4(a).
(b) The requirements in this subpart, with the exception of Sec.
26.203, apply to the licensees or other entities identified in Sec.
26.3(f) implementing this subpart under Sec. Sec. 26.604 and 26.605.
For these licensees and other entities, the requirements in Sec. Sec.
26.202 and 26.211 apply to the individuals identified in Sec. 26.4(a)
through (c) and any person licensed to operate under 10 CFR part 53;
and the requirements in Sec. Sec. 26.205 through 26.209 apply to the
individuals identified in Sec. 26.4(a).
0
89. Add Sec. 26.202 to read as follows:
Sec. 26.202 General provisions for facilities licensed under part 53.
(a) Policy. Licensees must establish a policy for the management of
fatigue for all individuals who are subject to the licensee's FFD
program and incorporate it into the written policy required in Sec.
26.606(a).
(b) Procedures. In addition to the procedures required in Sec.
26.606(b), licensees must develop, implement, and maintain procedures
that--
(1) Describe the process to be followed when any individual
identified in Sec. 26.4(a) through (c) makes a self-declaration that
he or she is not fit to safely and competently perform his or her
duties for any part of a working tour as a result of fatigue. The
procedure must--
(i) Describe the individual's and licensee's rights and
responsibilities related to self-declaration;
(ii) Describe requirements for establishing controls and conditions
under which an individual may be permitted or required to perform work
after that individual declares that he or she is not fit due to
fatigue; and
(iii) Describe the process to be followed if the individual
disagrees with the results of a fatigue assessment that is required
under Sec. 26.211(a)(2);
(2) Describe the process for implementing the controls required
under Sec. 26.205 for the individuals who are performing the duties
listed in Sec. 26.4(a);
(3) Describe the process to be followed in conducting fatigue
assessments under Sec. 26.211; and
(4) Describe the disciplinary actions that the licensee may impose
on an individual following a fatigue assessment, and the conditions and
considerations for taking those disciplinary actions.
(c) Training and assessments. Licensees must include the following
KAs in the content of the training and trainee assessments required in
Sec. 26.608:
(1) Knowledge of the contributors to worker fatigue, circadian
variations in alertness and performance, indications and risk factors
for common sleep disorders, shiftwork strategies for obtaining adequate
rest, and the effective use of fatigue countermeasures; and
(2) Ability to identify symptoms of worker fatigue and contributors
to decreased alertness in the workplace.
(d) Recordkeeping. Licensees must retain the following records for
at least 3 years or until the completion of all related legal
proceedings, whichever is later:
(1) Records of work hours for individuals who are subject to the
work hour controls in Sec. 26.205;
(2) For licensees implementing the requirements of Sec.
26.205(d)(3), records of shift schedules and shift cycles, or, for
licensees implementing the requirements of Sec. 26.205(d)(7), records
of shift schedules and records showing the beginning and end times and
dates of all averaging periods, of individuals
[[Page 87030]]
who are subject to the work hour controls in Sec. 26.205;
(3) The documentation of waivers that is required in Sec.
26.207(a)(4), including the bases for granting the waivers;
(4) The documentation of work hour reviews that is required in
Sec. 26.205(e)(3) and (e)(4); and
(5) The documentation of fatigue assessments that is required in
Sec. 26.211(g).
(e) Reporting. Licensees must include the following information in
a standard format in the annual FFD program performance report required
under Sec. 26.617(b)(2):
(1) A summary for each nuclear power plant site of all instances
during the previous calendar year when the licensee waived one or more
of the work hour controls specified in Sec. 26.205(d)(1) through
(d)(5)(i) and (d)(7) for individuals described in Sec. 26.4(a). The
summary must include only those waivers under which work was performed.
If it was necessary to waive more than one work hour control during any
single extended work period, the summary of instances must include each
of the work hour controls that were waived during the period. For each
category of individuals specified in Sec. 26.4(a), the licensee must
report--
(i) The number of instances when each applicable work hour control
specified in Sec. 26.205(d)(1)(i) through (iii), (d)(2)(i) and (ii),
(d)(3)(i) through (v), and (d)(7) was waived for individuals not
working on outage activities;
(ii) The number of instances when each applicable work hour control
specified in Sec. 26.205(d)(1)(i) through (iii), (d)(2)(i) and (ii),
(d)(3)(i) through (v), (d)(4) and (d)(5)(i), and (d)(7) was waived for
individuals working on outage activities; and
(iii) A summary that shows the distribution of waiver use among the
individuals applicable within each category of individuals identified
in Sec. 26.4(a) (e.g., a table that shows the number of individuals
who received only one waiver during the reporting period, the number of
individuals who received a total of two waivers during the reporting
period).
(2) A summary of corrective actions, if any, resulting from the
analyses of these data, including fatigue assessments.
(f) Audits. Licensees must audit the management of worker fatigue
under Sec. 26.615.
0
90. In Sec. 26.205, revise paragraphs (d)(7)(iii) and (d)(8) to read
as follows:
Sec. 26.205 Work Hours.
* * * * *
(d) * * *
(7) * * *
(iii) Each licensee shall state, in its FFD policy and procedures
required by either Sec. Sec. 26.27 and 26.203(a) and (b) or Sec. Sec.
26.202(a) and (b) and 26.606, the work hour counting system in Sec.
26.205(d)(7)(ii) the licensee is using.
(8) Each licensee shall state, in its FFD policy and procedures
required by either Sec. Sec. 26.27 and 26.203(a) and (b) or Sec. Sec.
26.202(a) and (b) and 26.606, the requirements with which the licensee
is complying: the minimum days off requirements in Sec. 26.205(d)(3)
or maximum average work hours requirements in Sec. 26.205(d)(7).
* * * * *
0
91. In Sec. 26.207, revise paragraph (a)(1)(ii) to read as follows:
Sec. 26.207 Waivers and exceptions.
(a) * * *
(1) * * *
(ii) A supervisor assesses the individual face to face and
determines that there is reasonable assurance that the individual will
be able to safely and competently perform his or her duties during the
additional work period for which the waiver will be granted. The
supervisor performing the assessment shall be trained as required by
either Sec. Sec. 26.29 and 26.203(c) or Sec. Sec. 26.202(c) and
26.608 and shall be qualified to direct the work to be performed by the
individual. If there is no supervisor on site who is qualified to
direct the work, the assessment may be performed by a supervisor who is
qualified to provide oversight of the work to be performed by the
individual. At a minimum, the assessment must address the potential for
acute and cumulative fatigue considering the individual's work history
for at least the past 14 days, the potential for circadian degradations
in alertness and performance considering the time of day for which the
waiver will be granted, the potential for fatigue-related degradations
in alertness and performance to affect risk-significant functions, and
whether any controls and conditions must be established under which the
individual will be permitted to perform work. For licensees and other
entities in Sec. 26.3(f), the assessment may be performed remotely
using electronic communications. In such instances, the assessment must
be supported by someone who is present in-person with the individual
whose alertness may be impaired, and that supporting person must be
trained under the requirements of either Sec. 26.29 and Sec.
26.203(c) or Sec. 26.202(c) and Sec. 26.608.
* * * * *
0
92. In Sec. 26.211, revise paragraphs (a)(1) and (3) and paragraph (b)
introductory text to read as follows:
Sec. 26.211 Fatigue assessments.
(a) * * *
(1) For-cause. In addition to any other test or determination of
fitness that may be required under Sec. Sec. 26.31(c), 26.77,
26.607(b), and 26.619, a fatigue assessment must be conducted in
response to an observed condition of impaired individual alertness
creating a reasonable suspicion that an individual is not fit to safely
and competently perform his or her duties, except if the condition is
observed during an individual's break period. If the observed condition
is impaired alertness with no other behaviors or physical conditions
creating a reasonable suspicion of possible substance abuse, then the
licensee need only conduct a fatigue assessment. If the licensee has
reason to believe that the observed condition is not due to fatigue,
the licensee need not conduct a fatigue assessment;
* * * * *
(3) Post-event. A fatigue assessment must be conducted in response
to events requiring post-event drug and alcohol testing as specified in
Sec. 26.31(c) or post-event tests in Sec. 26.607(b)(4). Licensees may
not delay necessary medical treatment in order to conduct a fatigue
assessment; and
* * * * *
(b) Only supervisors and FFD program personnel who are trained
under either Sec. Sec. 26.29 and 26.203(c) or Sec. Sec. 26.202(c) and
26.608 may conduct a fatigue assessment. The fatigue assessment must be
conducted face to face with the individual whose alertness may be
impaired. For licensees and other entities in Sec. 26.3(f), a fatigue
assessment may be performed remotely using electronic communications.
In such instances, the fatigue assessment must be supported by someone
who is present in-person with the individual whose alertness may be
impaired, and that supporting person must be trained in accordance with
the requirements of either Sec. Sec. 26.29 and 26.203(c) or Sec. Sec.
26.202(c) and 26.608.
* * * * *
0
93. Add Subpart M, consisting of Sec. Sec. 26.601 through 26.619, to
read as follows:
Subpart M--Fitness for Duty Programs for Facilities Licensed Under
10 CFR Part 53
Sec.
26.601 Applicability.
26.603 General provisions.
[[Page 87031]]
26.604 FFD program requirements for facilities that satisfy the
Sec. 26.603(c) criterion.
26.605 FFD program requirements for facilities that do not implement
Sec. 26.604.
26.606 Written policy and procedures.
26.607 Drug and alcohol testing.
26.608 FFD program training.
26.609 Behavioral observation.
26.610 Sanctions.
26.611 Protection of information.
26.613 Appeals process.
26.615 Audits.
26.617 Recordkeeping and reporting.
26.619 Suitability and fitness determinations.
Sec. 26.601 Applicability.
A licensee or other entity in Sec. 26.3(f), at its discretion, may
establish, implement, and maintain a fitness-for-duty (FFD) program
that satisfies the requirements of this subpart for those categories of
individuals in Sec. 26.4, as applicable, and any person licensed to
operate under 10 CFR part 53. If a licensee or other entity in Sec.
26.3(f) does not elect to implement an FFD program that satisfies the
requirements of this subpart, then those categories of individuals in
Sec. 26.4, as applicable, and any person licensed to operate under 10
CFR part 53 must be subject to an FFD program that satisfies all part
26 requirements, except for those requirements in subparts K and M.
Sec. 26.603 General provisions.
(a) FFD program description. An applicant's description of the FFD
program in its Final Safety Analysis Report, required by subpart H of
part 53 of this chapter, must include--
(1) If the applicant performed the analysis under paragraph (c) of
this section, a summary of the analysis, including the assumptions,
methodology, conclusion, and references;
(2) A statement whether the FFD program will be implemented
pursuant to Sec. 26.604 or Sec. 26.605, or will satisfy all part 26
requirements, except for the requirements in subparts K and M;
(3) A discussion of the applicability of the FFD program to those
individuals described in Sec. 26.4 and how the program will be
implemented offsite at a U.S. Nuclear Regulatory Commission (NRC)-
licensed facility authorized to assemble or test a manufactured
reactor, if applicable;
(4) A description of the drug and alcohol testing and fitness
determination process to be implemented through the licensee's or other
entity's procedures, including the collection and testing facilities to
be used, biological specimens to be collected, and sanctions to be
imposed upon a confirmed FFD policy violation; and
(5) A summary of the FFD performance monitoring and review program
(PMRP), including the measures and thresholds required by paragraph
(d)(1) of this section.
(b) FFD program implementation and availability. For the licensees
and other entities in Sec. 26.3(f), other than the holder of a
manufacturing license (ML), the FFD program must be implemented no
later than the start of construction activities, as defined in Sec.
26.5, and maintained until the NRC's docketing of the license holder's
certifications described in Sec. 53.1070 of this chapter. For holders
of an ML, the FFD program must be implemented no later than the start
of activities that assemble the manufactured reactor and maintained
until expiration of the ML.
(c) Criterion and analysis for an FFD program. For a licensee or
other entity to implement an FFD program under Sec. 26.604, the
licensee or other entity must perform a site-specific analysis to
demonstrate that the facility and its operation satisfy the criterion
in Sec. 53.860(a)(2) of this chapter. The licensee or other entity
must maintain the analysis, including updates to reflect changes made
to the staffing, FFD programs, or offsite support resources described
in the analysis, to show that the facility and its operation continue
to satisfy the criterion, until permanent cessation of operations under
Sec. 53.1070 of this chapter.
(d) FFD performance monitoring and review. A licensee or other
entity must establish performance measures and associated thresholds as
described in paragraph (d)(1) of this section and monitor the
effectiveness of its FFD program by comparing performance data against
these performance measures and thresholds, in a manner sufficient to
satisfy the Sec. 26.23 performance objectives.
(1) PRMP elements. The PMRP must be documented and maintained and
include the following program elements:
(i) Performance measures. Performance measures must be identified
and designed to monitor FFD program performance.
(A) If the licensee or other entity is subject to the requirements
in Sec. 26.604, then the monitoring program must include performance
measures for the following: the behavioral observation program;
occurrence of FFD policy violations categorized by licensee employee,
contractor/vendor, and labor category; and occurrence of individuals
with potentially disqualifying information or who possessed FFD
prohibited items.
(B) If the licensee or other entity is subject to the requirements
in Sec. 26.604 and has implemented a drug testing program at its
discretion or is subject to the requirements of Sec. 26.605, then the
monitoring program must include performance measures identified in
paragraph (d)(1)(i)(A) of this section. This monitoring program must
also include performance measures for the pre-access and random
positive testing rates, random testing rate for licensee employees and
contractor/vendors, and the number of subversion attempts categorized
by licensee employee, contractor/vendor, and labor category.
(ii) Thresholds. Licensee- or other entity-specific thresholds for
its site-specific performance measures must be established and used to
facilitate corrective actions to maintain FFD program performance.
Initial thresholds must be based on FFD performance data from
comparable facilities subject to part 26, the licensee's or other
entity's fleet-level program performance if applicable, and industry
FFD performance data.
(iii) Monitoring program. Licensees and other entities must monitor
the performance of their FFD programs against licensee- or other
entity-established performance measures and thresholds as FFD
performance data is received to determine whether a threshold has been
exceeded. Licensees and other entities must perform year-to-year
comparisons of site-specific performance; site-specific performance to
the licensee's or other entity's fleet-level program performance, if
applicable; and site-specific to industry performance.
(iv) Quantitative and qualitative reviews. The PMRP must include a
documented review of the elements in paragraph (d)(1)(i) through (iii)
of this section and the following qualitative elements.
(A) Worker protections. The review must include a documented
assessment of the licensee's or other entity's implementation of the
protections described in Sec. Sec. 26.606(b)(1)(iii), 26.611, and
26.613.
(B) Laboratory test results and Medical Review Officer performance.
The review must include a documented assessment of whether the actions
taken by the Medical Review Officer (MRO) met the requirements in Sec.
26.185 based on the laboratory test results reported under Sec.
26.169. This review must include a comparative analysis between the
point of collection testing and assessment (POCTA) screening result(s)
and the corresponding specimen test results obtained from the U.S.
Department of Health and Human
[[Page 87032]]
Services (HHS)-certified laboratory if the POCTA indicated a positive,
adulterated, substituted, or invalid screening result or discrepant
biological marker, to assess the effectiveness of the POCTA and to
inform MRO decisions under Sec. 26.185 or Sec. 26.607(m)(6).
(C) Change control. The review must include a documented assessment
of the changes made under paragraph (e) of this section to verify that
the summation of program changes has not resulted in a reduction in FFD
program effectiveness.
(2) Corrective actions. Corrective actions must be implemented to
address when FFD performance meets a licensee-established performance
threshold or to resolve a finding resulting from a qualitative review
or audit in a manner that restores performance and corrects root
causes, contributing causes, or both.
(3) Program review periodicity. The documented review in paragraph
(d)(1)(iv) of this section must be conducted biennially to assess and
modify licensee or other entity implementation of its FFD program. This
documented review must demonstrate that the performance measures and
thresholds are appropriate and adjusted as necessary based on site-
level and licensee's or other entity's fleet-level, if applicable,
program performance, and industry performance.
(i) Identified program weaknesses and corrective actions must be
summarized in the annual reporting requirement described in Sec.
26.617(b)(2) or Sec. 26.717, as applicable.
(ii) The program review must be completed and approved by the
licensee or other entity to support the reporting of PMRP weaknesses
and corrective actions as required in paragraph (d)(3)(i) of this
section every odd-numbered year, and the implementation of corrective
actions before May 15 of that odd-numbered year.
(e) FFD program change control. (1) The licensee or other entity
may make changes to its FFD program under this subpart if--
(i) The licensee or other entity performs and retains an analysis
demonstrating that the changes do not reduce the effectiveness of the
FFD program; or
(ii) The change was necessitated or justified by a change to part
26, laboratory processes or procedures, or guidance issued by the HHS
or NRC, as implemented by the licensee or other entity though its
procedures.
(2) A licensee or other entity desiring to make a change that
decreases FFD program effectiveness must implement a mitigating
strategy so the FFD program, as revised, will continue to satisfy the
performance objectives in Sec. 26.23 and not result in a reduction in
program effectiveness.
(3) Except for phencyclidine, and notwithstanding paragraph
(e)(1)(ii) of this section, the change control process may not be used
to reduce the minimum panel of drugs to be tested in Sec.
26.607(c)(1).
(4) The licensee must retain a record of each change made under
this section for a period of at least 5 years from the date the change
was implemented and summarize this change in its annual FFD performance
report required by Sec. 26.617(b)(2) or Sec. 26.717, as applicable.
Sec. 26.604 FFD program requirements for facilities that satisfy the
Sec. 26.603(c) criterion.
(a) FFD program. A licensee or other entity with an analysis that
demonstrates that its facility and operation satisfy the criterion in
Sec. 26.603(c) may elect to establish, implement, and maintain an FFD
program under this section. That FFD program must contain the following
elements:
(1) Applies to those individuals described in Sec. 26.4, as
applicable; and
(2) Implements the following requirements and subparts in this
part:
(i) Sec. 26.23, Performance objectives;
(ii) Sec. 26.603, General provisions;
(iii) Sec. 26.606, Written policies and procedures, (a) and, if
applicable (b);
(iv) Sec. 26.608, FFD program training;
(v) Sec. 26.609, Behavioral observation;
(vi) Sec. 26.610, Sanctions;
(vii) Sec. 26.611, Protection of information;
(viii) Sec. 26.613, Appeals process;
(ix) Sec. 26.615, Audits;
(x) Sec. 26.617, Recordkeeping and reporting;
(xi) Sec. 26.619, Suitability and fitness determinations;
(xii) Subpart A--Administrative Provisions;
(xiii) Subpart I--Managing Fatigue; and
(xiv) Subpart O--Inspections, Violations, and Penalties.
(b) [Reserved]
Sec. 26.605 FFD program requirements for facilities that do not
implement Sec. 26.604.
(a) Licensees and other entities who satisfy the criterion in Sec.
26.603(c), at their discretion, and licensees and other entities who do
not satisfy the criterion in Sec. 26.603(c), must establish,
implement, and maintain an FFD program under this section either during
construction activities as defined in Sec. 26.5, or during activities
performed under an ML that allows the assembly, testing, or both of a
manufactured reactor, as applicable. This FFD program must contain the
following elements:
(1) Applies to those individuals described in Sec. 26.4, as
applicable; and,
(2) Implements the following requirements and subparts in this
part--
(i) Sec. 26.23, Performance objectives;
(ii) Sec. 26.603, General provisions;
(iii) Sec. 26.606, Written policy and procedures;
(iv) Sec. 26.607, Drug and alcohol testing;
(v) Sec. 26.608, FFD program training;
(vi) Sec. 26.609, Behavioral observation;
(vii) Sec. 26.610, Sanctions;
(viii) Sec. 26.611, Protection of information;
(ix) Sec. 26.613, Appeals process;
(x) Sec. 26.615, Audits;
(xi) Sec. 26.617, Recordkeeping and reporting;
(xii) Sec. 26.619, Suitability and fitness determinations;
(xiii) Subpart A--Administrative Provisions;
(xiv) Subpart I--Managing Fatigue, in the case of holders of an ML
that allows the assembly, testing, or both of a manufactured reactor;
and
(xv) Subpart O--Inspections, Violations, and Penalties.
(b) Licensees and other entities who satisfy the criterion in Sec.
26.603(c), at their discretion, and licensees and other entities who do
not satisfy the criterion in Sec. 26.603(c), before the loading of
fuel onsite into a reactor vessel; before receiving a manufactured
reactor; or before individuals subject to part 26 operate, test,
perform maintenance of, or direct the maintenance or surveillance of
security-related equipment or equipment that a risk-informed evaluation
process has shown to be significant to public health and safety, must
establish, implement, and maintain an FFD program that--
(1) Applies to those individuals described in Sec. 26.4, as
applicable; and,
(2) Implements the following requirements and subparts--
(i) Sec. 26.23, Performance objectives;
(ii) Sec. 26.603, General provisions;
(iii) Sec. 26.606, Written policy and procedures;
(iv) Sec. 26.607, Drug and alcohol testing;
(v) Sec. 26.608, FFD program training;
(vi) Sec. 26.609, Behavioral observation;
(vii) Sec. 26.611, Protection of information;
(viii) Sec. 26.613, Appeals process;
(ix) Sec. 26.615, Audits;
(x) Subpart A--Administrative Provisions;
(xi) Subpart C--Granting and Maintaining Authorization;
(xii) Subpart D--Management Actions and Sanctions to be Imposed;
(xiii) Subpart H--Determining Fitness-for-Duty Policy Violations
and
[[Page 87033]]
Determining Fitness, unless using the HHS Guidelines for MRO evaluation
of drug test results, and determining fitness;
(xiv) Subpart I--Managing Fatigue;
(xv) Subpart N--Recordkeeping and Reporting Requirements; and
(xvi) Subpart O--Inspections, Violations, and Penalties.
Sec. 26.606 Written policy and procedures.
(a) Licensees and other entities that implement an FFD program
under this subpart must ensure that--
(1) A written FFD policy statement is provided to each individual
who is subject to the program before the individual is subject to
behavioral observation, drug and alcohol testing, or both.
(2) The FFD policy statement describes the performance objectives
in Sec. 26.23.
(3) The FFD policy statement describes the minimum days off
requirements in Sec. 26.205(d)(3) or maximum average work hours
requirements in Sec. 26.205(d)(7).
(4) The FFD policy statement must be written in sufficient detail
to provide affected individuals with information on what is expected of
them and what consequences may result from a lack of adherence to the
policy, including those elements described in Sec. 26.606(b), part 26-
required sanctions, and required medical/clinical treatment and follow-
up testing for FFD policy violations.
(5) The FFD policy statement describes the individual's
responsibilities to report for work in a physiological and
psychological condition that enables the safe and competent performance
of assigned duties and responsibilities and inform a licensee- or other
entity-designated representative when the individual determines that
this cannot be accomplished.
(b) Licensees and other entities must establish, implement, and
maintain written procedures that address the following topics:
(1) If implementing a drug and alcohol testing program under this
subpart,
(i) The methods and techniques to collect and test for drugs and
alcohol and for the shipping and temporary storage of biological
specimens used for drug testing at HHS-certified laboratories,
(ii) The urine specimen volumes, techniques for split specimen
collections, and the acceptability of a urine specimen as described in
Sec. 26.111 or as described in the HHS Guidelines,
(iii) Protecting the privacy of an individual who provides a
specimen, protecting the integrity of the specimen, and ensuring that
the test results are valid and attributable to the correct individual,
and
(iv) If the licensee or other entity elects to use the HHS
Guidelines, the name of the specific HHS Guideline and revision being
implemented by the licensee or other entity and a description of the
specific sections in the guideline that are being implemented in the
procedure, including specimen collections, drug testing, and evaluation
of test results.
(2) The immediate and follow-up actions that will be taken, and the
procedures to be used, in those cases in which individuals who are
subject to the FFD program:
(i) Have been involved in the use, sale, or possession of illegal
substances, illegal drugs, or illicit substances;
(ii) Are impaired by any illegal substances, illegal drugs, or
illicit substances or the consumption of alcohol as determined by
behavioral observation or a test that measures blood alcohol
concentration;
(iii) If drug and alcohol testing is conducted, attempted to
subvert the testing process by adulterating or diluting specimens (in
vivo or in vitro), substituting specimens, or by any other means;
(iv) If drug and alcohol testing is conducted, refused to provide a
specimen for analysis or follow instructions provided by FFD program
personnel;
(v) Had legal action taken relating to drug or alcohol use; or
(vi) Demonstrated character or actions indicating that the
individual cannot be trusted or relied upon to perform those duties and
responsibilities or maintain access to NRC-licensed facilities, special
nuclear material (SNM), or sensitive information.
(3) The process, including the duties and responsibilities of FFD
program personnel, to be followed if an individual's behavior or
condition raises a concern regarding the possible use, sale, or
possession of illegal drugs on- or offsite; the possible use or
possession of alcohol on the NRC-licensed facility; impairment from any
cause that in any way could adversely affect the individual's ability
to safely and competently perform the individual's duties; or the
receipt of credible information indicating that the individual cannot
be trusted or relied on to perform those duties and responsibilities
making the individual subject to this part.
(4) Operation and oversight of an onsite or offsite collection
facility.
(5) The fatigue management requirements in Sec. Sec. 26.202(b) and
either 26.205(d)(3) or (d)(7).
(6) Measures to prevent subversion of drug and alcohol tests
conducted onsite and offsite.
Sec. 26.607 Drug and alcohol testing.
Licensees and other entities implementing Sec. 26.604, at their
discretion, and licensees and other entities implementing Sec. 26.605
must perform drug and alcohol testing that complies with the following
requirements--
(a) Split specimens. Split specimen collections of oral fluid or
urine must be used for the test conditions described in paragraph (b)
of this section. A split specimen collection need not be used if the
licensee or other entity elects to use a POCTA device for a screening
test conducted during random testing under paragraphs (b)(2) and (h) of
this section or a protected area portal monitor indication that drugs
or alcohol were detected under paragraph (j) of this section. Testing
of the split specimen (specimen B) requires the donor's permission
unless ordered by the MRO to resolve an invalid test result obtained
for specimen A.
(b) Test conditions. Individuals identified in Sec. 26.4 must be
subject to drug and alcohol testing under the following conditions:
(1) Pre-access. A pre-access test must be conducted for drugs and
alcohol before performing or directing the conduct of roles and
responsibilities making the individual subject to this subpart or being
granted unescorted access to the protected areas of the NRC-licensed
facility. A pre-access test must have been conducted no more than 14
days before the individual is granted unescorted access.
(2) Random. Random testing for drugs and alcohol must--
(i) Be administered in a manner that provides reasonable assurance
that individuals are unable to predict the time periods during which
specimens will be collected;
(ii) Require individuals who are selected for random testing to
report to the onsite collection site as soon as reasonably practicable
after notification, within the time period specified in the FFD program
procedure;
(iii) Ensure that all individuals in the population that is subject
to random testing on a given day have an equal probability of being
selected and tested;
(iv) Ensure that an individual completing a test is immediately
eligible for another random test; and
(v) Ensure that the sampling process used to select individuals for
random
[[Page 87034]]
testing provides that the number of random tests performed annually is
equal to at least 50 percent for licensee employees and 50 percent for
contractor/vendors at the NRC-licensed site.
(3) For-cause. A for-cause drug test, alcohol test, or both, must
be conducted onsite in response to an individual's observed behavior or
physical condition indicating possible substance abuse or after
receiving credible information that an individual is engaging in
substance abuse, as defined in Sec. 26.5;
(4) Post-event. A post-event test for drugs and alcohol must be
conducted--
(i) As soon as practical after an event involving a human error
that was committed by an individual specified in Sec. 26.4, where the
human error may have caused or contributed to the event. This test must
be conducted onsite unless the individual requires offsite medical
care. The licensee or other entity must test the individual(s) who
committed or directed the error and need not test individuals who were
affected by the event and whose actions likely did not cause or
contribute to the event. The licensee or other entity must describe in
its procedures what constitutes a human error.
(ii) Within 4 hours of an event unless immediate medical
intervention precludes the conduct of the test on the individual(s) who
caused or contributed to the accident(s), if the event results in--
(A) An illness or personal injury to any individual which results
in death, days away from work, restricted work, transfer to another
job, medical treatment beyond first aid, loss of consciousness, or
other significant illness or injury, as diagnosed by a licensee- or
other entity-designated physician or other licensed health care
professional, even if the illness or injury does not result in death,
days away from work, restricted work or job transfer, medical treatment
beyond first aid, or loss of consciousness; or
(B) Damage to any safety- or security-related structures, systems,
and components; and
(5) Follow-up. An individual subject to part 26 who has violated
the FFD policy for substance use or abuse, or the sale, use, or
possession of illegal drugs must be subject to a follow-up series of
tests for drugs, alcohol, or both to verify an individual's continued
abstinence from substance abuse.
(c) Urine and oral fluid specimens. (1) All urine or oral fluid
specimens must be subject to validity testing, including an adulterant
and biological marker, and tested for the substances listed in Sec.
26.31(d)(1), except as allowed by Sec. 26.603(e)(3).
(2) For the use of urine as the biological specimen to be tested,
the following requirements must be implemented--
(i) Sec. 26.115, Collecting a urine specimen under direct
observation;
(ii) Sec. 26.119, Determining ``shy'' bladder; and
(iii) Sec. 26.163, Cutoff levels for drugs and drug metabolites,
(a)(2) regarding special analysis testing.
(3) For alcohol testing onsite, the following requirements must be
implemented--
(i) Sec. 26.91, Acceptable devices for conducting initial and
confirmatory tests for alcohol and methods of use;
(ii) Sec. 26.93, Preparing for alcohol testing;
(iii) Sec. 26.95, Conducting an initial test for alcohol using a
breath specimen;
(iv) Sec. 26.97, Collecting oral fluid specimens for alcohol and
drug testing;
(v) Sec. 26.99, Determining the need for a confirmatory test for
alcohol;
(vi) Sec. 26.101, Conducting a confirmatory test for alcohol; and,
(vii) Sec. 26.103, Determining a confirmed positive test result
for alcohol.
(4) For all test conditions in paragraph (b) of this section,
except for the use of a POCTA screening device in paragraph (h) of this
section, and for MRO-directed tests under Sec. 26.185, drug testing
must be performed at an HHS-certified laboratory for the specific
biological specimen to be tested. Only HHS-certified laboratory test
results from urine and oral fluid specimens may be used for the
issuance of a part 26-required sanction. The licensee or other entity
must establish and maintain a contract with a primary and a back-up
HHS-certified laboratory (with a different Certifying Scientist) for
the specimen(s) to be tested. These contracts must stipulate that the
laboratories are subject to inspection or audit by the licensee or
other entity and that records and documents must be provided and/or
able to be photocopied and removed from the premises to support the
inspection or audit.
(d) Privacy and integrity. The specimen collection and drug and
alcohol testing procedures of FFD programs must protect the donor's
privacy and the integrity of the specimen and implement quality
controls to ensure that test results are valid and attributable to the
correct individual.
(e) Offsite collection facilities. At the licensee's or other
entity's discretion, specimen collections and alcohol testing may be
conducted at a local hospital or other facility licensed to conduct
specimen collections and perform alcohol testing and audited by the
State or a State-designated entity. The licensee or other entity must
audit these facilities, if used, before their initial use and then on a
biennial basis to confirm that the facility procedures are comparable
to those described in subpart E of this part or the HHS Guidelines for
urine and oral fluid.
(f) Initial testing. A licensee or other entity subject to this
subpart performing an initial test must use an immunoassay, or an
alternative technology established in its FFD program through Sec.
26.603(e), that satisfies the requirements of the U.S. Food and Drug
Administration (FDA) for commercial distribution. Specimens that yield
positive, adulterated, substituted, or invalid initial validity or drug
test results or discrepant biological markers must be subject to
confirmatory testing by an HHS-certified laboratory, certified for that
biological specimen, except for invalid specimens that cannot be
tested.
(g) Oral fluid testing. If the licensee or other entity elects to
use oral fluid for drug or alcohol testing, the collection, packaging,
and temporary storage of the drug or alcohol test device, and shipment
of an oral fluid specimen to an HHS-certified laboratory or the
collection of an oral fluid specimen for alcohol testing must be
performed in accordance with licensee- or other entity-established
procedures based either on the requirements in part 26 or the
procedures in HHS Guidelines identified by the licensee or other entity
in Sec. 26.606(b)(1)(iv). The device must have received premarket
approval from the FDA and must not expire before laboratory testing.
The drugs, drug metabolites, initial and confirmatory testing cutoffs,
and biological markers, if applicable, must be those established by HHS
for oral fluid testing and the alcohol cutoffs in this part or, if not
established by HHS or the NRC for the panel of drugs and drug
metabolites to be tested, as determined and documented by a forensic
toxicologist review conducted pursuant to Sec. 26.31(d)(1)(i)(D).
(h) Point of collection testing and assessment. (1) If the licensee
or other entity elects to use a POCTA device, then it may only be used
for pre-access and random drug and alcohol initial testing in paragraph
(b) of this section, the alcohol testing process in paragraph (c)(3) of
this section, and the portal area screening process in paragraph (j) of
this section. Before the licensee or other entity uses a POCTA device,
a forensic toxicologist must review and document
[[Page 87035]]
their evaluation that the validity and accuracy of the device for
alcohol and/or the drugs and drug metabolites listed in Sec. 26.31(d)
are comparable to the performance achieved by initial testing conducted
using a similar technology at an HHS-certified laboratory. For initial
testing of drugs and drug metabolites using a POCTA device, this review
must include a documented evaluation of POCTA device performance
against the requirements in Sec. 26.161(b) for a urine specimen or the
procedures in the HHS Guidelines for urine or oral fluid, as
implemented by the licensee or other entity through its procedures.
(2) If the performance of the POCTA device is not comparable to
that achieved from initial testing conducted by an HHS-certified
laboratory as determined by the forensic toxicologist, then the
licensee or other entity must implement a mitigating strategy to
maintain program effectiveness under Sec. 26.603(e)(2), as applicable.
(3) The licensee and other entity must implement procedures for the
use of a POCTA that ensures the effectiveness of the collection
process, assessment of the screening results, and prevention of
subversion attempts.
(4) If the use of a POCTA device indicates a discrepant biological
marker or that a test result exceeds the initial test cutoff, the
specimen is invalid, or the individual subverted the drug or alcohol
test, then the individual must be immediately removed from duties,
responsibilities, and access making the individual subject to this
subpart.
(i) The individual must be subject to an immediate drug and alcohol
test using the alcohol testing process in paragraph (c)(3) of this
section for a positive alcohol screen and either oral fluid or urine by
a collection kit that is not a POCTA device, but of the same type of
biological specimen collected by the POCTA, for validity, if required,
and initial and confirmatory testing by an HHS-certified laboratory.
(ii) If this individual shows any signs of impairment, the
individual's authorization must be temporarily removed until the MRO
reviews the laboratory test result(s), interviews the individual, and
performs a determination of fitness under Sec. 26.189 or Sec. 26.619,
as applicable, that enables the restoration of authorization.
(i) Hair testing. The testing of hair specimens may only be used to
inform a licensee's or other entity's determination of whether the
individual is trustworthy and reliable under the test condition in
paragraph (b)(1) of this section to supplement the information gained
from a pre-access test using oral fluid or urine as the test specimen
and must be conducted at an HHS-certified laboratory certified for hair
specimens.
(1) If used, this process must be described in the licensee's or
other entity's FFD policy and described in detail in its procedure. The
panel of drugs and drug metabolites to be evaluated must only include
those listed as Schedule I or II of section 202 of the Controlled
Substances Act [21 U.S.C. 812]. The collection, packaging, and
temporary storage of a hair specimen and shipment of the specimen to an
HHS-certified laboratory must be conducted in accordance with the HHS
Guidelines. The test kit must be FDA cleared, and licensee- or other
entity-designated FFD program personnel must conduct the collection,
packaging, temporary storage, shipping, and custody and control of the
specimen.
(2) Before the licensee or other entity begins to conduct hair
testing, the initial and confirmatory testing cutoffs must be the
cutoffs established by HHS for hair testing or, if not established by
HHS or the NRC, as determined by a forensic toxicologist review
conducted pursuant to Sec. 26.31(d)(1)(i)(D).
(3) Confirmed positive test results must be considered potentially
disqualifying FFD information until proven otherwise by a review under
Sec. 26.613. Sanctions under this subpart must not be issued for any
FFD policy violation involving a drug test using a hair specimen unless
the licensee or other entity determines that the individual subverted,
as defined in Sec. 26.5, the hair test.
(j) Portal area screening. A non-invasive point of collection
testing instrument may be used to screen individuals for drugs, drug
metabolites, and alcohol before the individuals' entry into or exit
from a protected or vital area.
(1) If a licensee or other entity uses such an instrument, then
before such use, a forensic toxicologist must review the instrument and
document an evaluation that the instrument and setpoints used in the
instrument are acceptable for use for the detection and screening of
the drugs and drug metabolites selected for screening from the panel of
drugs and drug metabolites to be tested under the FFD program and
alcohol and its metabolites.
(2) The instrument must be operated in accordance with the
manufacturer's specifications. If screening detects the presence of
drugs, drug metabolites, or alcohol at or above the instrument set
point(s), the individual screened by the instrument must be subject to
a POCTA screening test using the process described in paragraph (h) of
this section or an oral fluid or urine test that is sent to an HHS-
certified laboratory.
(3) A part 26 sanction may not be issued to an individual based
solely on a portal area screening instrument detection that drugs or
alcohol exceed the instrument's established setpoint.
(k) Blood testing. The testing of blood specimens may only be
conducted under the order of the licensee- or other entity-designated
MRO for a valid medical reason as confirmed by the MRO pursuant to
Sec. 26.31(d)(5). This specimen must be subject to testing by a
laboratory that satisfies quality control requirements that are
comparable to those required for certification by the HHS.
(l) Custody-and-control form. For the collection and packaging of
urine, oral fluid, and hair specimens, the licensee or other entity
must use a custody-and-control form approved by the U.S. Office of
Management and Budget. For the use of a POCTA device, the licensee or
other entity must implement a licensee- or other entity-approved and -
maintained procedure that ensures the reliability of the tracking,
handling, and storage of a specimen from the point of specimen
collection to the final disposition of the specimen and the reliability
of an identification system to uniquely assign the specimen to the
donor.
(m) Medical Review Officer. Licensees or other entities must--
(1) Require their designated MRO to review positive, adulterated,
substituted, and dilute confirmatory drug and validity test results and
test results of questionable validity to determine whether the donor
has violated the FFD policy for urine and oral fluid specimens. The
review must be completed before reporting the results to the individual
designated by the licensee or other entity to assess authorization or
perform the suitability and fitness determinations required under Sec.
26.619, or, if required, that are described in subpart H of this part.
(2) Require their MRO to satisfy the requirements in Sec. 26.183
and, prior to conducting any activities under this part, attend and
pass a medical- or clinical-based training session to improve his/her
knowledge of MRO duties and responsibilities, drug and alcohol testing
processes and procedures, and evaluation of drug testing results. This
training session must be conducted by a nationally recognized MRO
training and certification organization that has been assessed by the
licensee's or other entity's FFD program personnel to include the
technical elements an MRO must implement under Sec. 26.185. An
[[Page 87036]]
MRO who performed the duties and responsibilities in Sec. Sec. 26.185
and 26.187 for at least 3 continuous years in the last 10 years prior
to being hired or contracted by the licensee or other entity satisfies
the requirements in this paragraph.
(3) Require their MRO to attend a medical- or clinical-based
training session on a triennial basis to improve his/her knowledge of
changes in drug and alcohol testing processes and procedures and
evaluation of drug testing results.
(4) Require their MRO to determine whether a biological specimen is
positive, adulterated, substituted, dilute or of questionable validity
by implementing the requirements in Sec. 26.185 or the HHS Guidelines
through the licensee's or other entity's procedures.
(i) If Sec. 26.185 or the HHS Guidelines, as used by the licensee
or other entity in its procedures, are insufficient to make this
determination, then guidance issued by a State agency in the state in
which the NRC-licensed facility is located, Federal agencies, or
nationally recognized MRO training and certification organizations may
be used to inform an MRO determination.
(ii) An MRO need not review a confirmed alcohol positive test
result determined by an evidentiary breath testing device under
paragraphs (c)(3)(vi) and (vii) of this section.
(5) Require their MRO to determine and approve the use of oral
fluid or urine as an alternative biological specimen when the donor
cannot provide a specimen for testing. This determination and the
retest must be documented and completed as soon as reasonably
practicable.
(6) Require the MRO to review all specimens screened and tested
associated with a drug-related FFD policy violation. This review
includes POCTA, split specimens, and all specimens taken to resolve a
discrepant condition, such as a possible subversion attempt, impairment
without a known cause, or a donor-requested or MRO-directed re-test. To
resolve a discrepant condition, the MRO is authorized to test a
specimen for a biological marker, adulterants, or additional drugs.
(n) Limitations of screening and testing. Specimens collected under
NRC regulations may only be designated or approved for screening and
testing as described in this part and may not be used to conduct any
other analysis or test without the written permission of the donor.
Analyses, screens, and tests that may not be conducted include, but are
not limited to, DNA testing, serological typing, or any other medical
or genetic test used for diagnostic or specimen identification
purposes. No biological specimens may be passively sampled and analyzed
in a manner different than described in this subpart.
(o) Specimen collectors. All onsite specimen collections, except a
collection by a portal area screening instrument in paragraph (j) of
this section, must be conducted by licensee- or other entity-designated
and -trained personnel.
Sec. 26.608 FFD program training.
(a) FFD program training. (1) Individuals must be trained in the
FFD policy and procedure, including fatigue management, and their FFD
program responsibilities. Individuals who collect specimens for testing
or screening must also be trained in specimen collector duties and
responsibilities, including, at a minimum, specimen collection, custody
and control, identification and response to subversion attempts, and
privacy. For licensees and other entities of commercial nuclear plants,
the FFD program training program must use a systems approach to
training as defined in Sec. 53.725 of this chapter and described in
Sec. 53.830 of this chapter for those individuals in Sec. 26.4.
(2) FFD program training must include training on the behavioral
observation program. The behavioral observation program training must
include the detection of physiological behaviors or conditions that may
indicate--
(i) Possible use, sale, or possession of illegal drugs or illicit
drugs, or substance abuse on- or offsite;
(ii) Use or possession of alcohol onsite or use while on duty
offsite;
(iii) Impairment from fatigue or any cause that, if left
unattended, could result in inattentiveness or human errors; and
(iv) Any individual's inability to safely and competently perform
assigned duties and responsibilities or act in a trustworthy and
reliable manner while having access to protected areas, SNM, or
sensitive information.
(3) Training must explain that an individual's FFD policy violation
will--
(i) Subject the individual to an FFD program-required sanction
designed to preclude recurrence of an FFD policy violation;
(ii) Contribute to the licensee's or other entity's assessment of
whether the individual can be trusted and relied upon to safely and
competently perform the assigned duties and responsibilities making the
individual subject to this subpart;
(iii) Be used to inform the licensee's or other entity's insider
mitigation and access authorization programs under Sec. Sec. 73.55,
73.56, 73.100 or 73.120 of this chapter; and
(iv) Be used to inform other NRC licensees and other entities
subject to part 26 when FFD program information is requested to support
authorization determinations under subpart C of this part or Sec. Sec.
73.56 or 73.120 of this chapter.
(b) Training and assessments. Training and a trainee assessment
must be conducted before pre-access testing, and refresher training and
trainee assessments must be conducted periodically thereafter.
(c) Training program review. The licensee or other entity must
periodically evaluate its FFD training program and revise it as
appropriate to reflect industry experience as well as applicable
changes to the regulations in this part, the HHS Guidelines, if used,
and specimen collection and testing processes implemented by the
licensee or other entity.
Sec. 26.609 Behavioral observation.
(a) Licensees and other entities must ensure that the individuals
who are subject to this subpart are subject to behavioral observation
and that behavioral observation is performed by all individuals subject
to this subpart.
(b) Licensees and other entities must require all individuals
subject to the FFD program to report to the licensee- or other entity-
designated representative any onsite or offsite behaviors or activities
by individuals subject to this part that may constitute an unreasonable
risk to the safety or security of the NRC-licensed facility or SNM or
may cause harm to others. This reporting must include any information
relating to character or reputation of the individual indicating that
the individual cannot be trusted or relied upon to perform those duties
and responsibilities or maintain access to NRC-licensed facilities,
SNM, or sensitive information that makes them subject to part 26.
(c) Behavioral observation must be performed visually, in-person,
and, when necessary, remotely by live video and audible streaming and
capture, to observe the behavior of individuals in the workforce
subject to the requirements in this subpart.
(d) Not withstanding paragraph (c) of this section, for a reactor
facility where individual task loading does not allow for the effective
conduct of behavior observation in addition to assigned operational
tasks, the licensee or other entity must implement a live video and
audible streaming and capture system to conduct behavioral observation
of
[[Page 87037]]
persons licensed to operate under 10 CFR part 53 who manipulate the
controls of any commercial nuclear plant licensed under 10 CFR part 53.
Sec. 26.610 Sanctions.
Licensees and other entities that implement an FFD program under
this subpart must establish sanctions for FFD policy violations that,
at a minimum, prohibit the individuals specified in Sec. 26.4 from
being assigned to perform or direct those duties and responsibilities
or maintaining authorization making them subject to this subpart. The
severity of the sanction must escalate with the number of occurrences
and severity of the FFD policy violation. The sanction must be long
enough to act as a deterrent and, if the individual is retained as a
licensee employee or contractor/vendor, facilitate the individual to
complete counseling or treatment. The sanctions must include a minimum
5-year denial of access to the NRC-licensed facility for any individual
who is determined to have been involved in the sale, use, or possession
of illegal drugs or the consumption of alcohol within a protected area
of any facility licensed under part 53 of this chapter or within a
transporter's facility or vehicle used in the conveyance of formula
quantities of strategic SNM while the individual is subject to this
subpart, and a permanent denial of access to the NRC-licensed facility
for three FFD policy violations or any subversion attempt of any drug
or alcohol test or screening process, including subversion attempts at
any licensee or other entity subject to this part.
Sec. 26.611 Protection of information.
(a) Licensees and other entities that collect personal information
about an individual for the purpose of complying with this subpart must
establish and maintain a system of files and procedures to prevent
unauthorized disclosure.
(b) Licensees and other entities must obtain a signed consent that
documents the individual's acceptance of being subject to the FFD
program and authorizes the disclosure of the personal information
collected and maintained under this subpart, except for disclosures to
the individuals and entities specified in Sec. 26.37(b)(1) through
(b)(6), (b)(8), and persons deciding matters under review in Sec.
26.613. This signed and dated consent must be obtained before making
the individual subject to the FFD program.
Sec. 26.613 Appeals process.
Licensees and other entities that implement an FFD program under
this subpart must establish and implement procedures for the review of
a determination that an individual in Sec. 26.4 has violated the FFD
policy. The procedure must provide for an objective and impartial
review of the facts related to the determination that the individual
has violated the FFD policy and a schedule for the completion of the
review.
Sec. 26.615 Audits.
(a) Licensees and other entities that implement an FFD program
under this subpart must audit their programs at a frequency that
ensures the continuing effectiveness of their FFD program, FFD program
elements that are provided by C/Vs, and the FFD programs of C/Vs that
are accepted by the licensee or other entity. Corrective actions must
be as soon as reasonably practicable to resolve any problems identified
in an audit and preclude recurrence.
(b) The subject matter, scope, and frequency of audits must be
revised as necessary to improve or maintain program performance based
on findings resulting from licensee or other entity implementation of
its FFD PMRP in Sec. 26.603(d).
(c) Licensees and other entities may conduct joint audits or accept
audits of C/Vs so long as the audit addresses the relevant services of
the C/Vs.
(d) Licensees and other entities must audit HHS-certified
laboratories unless the licensee's or other entity's panel of drugs and
drug metabolites to be tested is equivalent to the panel by which the
laboratory is certified by HHS or is subject to the standards and
procedures for drug testing and evaluation used by the laboratory under
the HHS Guidelines. Licensees and other entities must audit any
hospital or other facility licensed by the State (or State-designated
entity) if used to conduct specimen collections and perform alcohol
testing under this part on a biennial basis to confirm that the
facility procedures are comparable to those described in subpart E of
this part, for urine and oral fluid.
Sec. 26.617 Recordkeeping and reporting.
(a) Licensees and other entities that implement FFD programs under
this subpart must ensure that records pertaining to the administration
of their program, which may be stored and archived electronically, are
maintained so that they are available for NRC inspection purposes and
for any legal proceedings resulting from the administration of the
program. Records pertaining to the administration of the FFD program
and FFD performance data required by Sec. 26.717 must be retained
until license termination.
(b) Licensees and other entities must make the following reports:
(1) Reports to the NRC Operations Center by telephone within 24
hours after the licensee or other entity discovers any intentional act
that casts doubt on the integrity of the FFD program and any
programmatic failure, degradation, or discovered vulnerability of the
FFD program that may permit undetected drug or alcohol use or abuse by
individuals who are subject to this subpart. These events must be
reported under this subpart, rather than under the provisions of Sec.
73.1200 of this chapter; and
(2) Annual program performance reports for the FFD program,
including the FFD program performance data listed in Sec. 26.717(b),
as applicable. Licensees and other entities must submit FFD program
performance data (for January through December) to the NRC annually,
before March 1 of the following year and must use unexpired NRC-
provided forms for the electronic submission of FFD information to the
NRC.
(c) Licensees and other entities subject to this subpart must
describe in sufficient detail to support an authorization
determination, an individual's FFD policy violation (while protecting
privacy information under Sec. 26.611) and FFD program weakness to
NRC, licensees, and other entities subject to this part when requested
to support authorization determinations under subpart C of this part or
Sec. 73.120 of this chapter, as applicable, or to support licensee or
other entity performance monitoring.
Sec. 26.619 Suitability and fitness determinations.
Licensees and other entities that implement FFD programs under this
subpart must develop, implement, and maintain procedures for evaluating
whether to assign individuals to perform or direct those duties and
responsibilities making them subject to this subpart. A suitability or
fitness determination conducted for cause must be performed face to
face. A suitability or fitness determination conducted for cause may be
performed remotely using electronic communications only when supported
by someone who is present in-person with the individual being assessed,
and that supporting person must be trained in accordance with the
requirements of either Sec. Sec. 26.29 or 26.608.
0
94. Revise Sec. 26.709 to read as follows:
[[Page 87038]]
Sec. 26.709 Applicability.
(a) The requirements of this subpart apply to the FFD programs of
licensees and other entities specified in Sec. 26.3(a) through (d),
except for FFD programs that are implemented under subpart K of this
part.
(b) The requirements in this subpart apply to the FFD programs of
licensees and other entities specified in Sec. 26.3(f) that elect not
to implement the requirements in subpart M or elect to implement the
requirements in Sec. 26.605(b).
Sec. 26.711 [Amended]
0
95. In Sec. 26.711, in paragraphs (c) and (d), remove the phrase ``(c)
and (d),'' and add in its place the phrase ``(c), (d), and (f),''.
Sec. 26.825 [Amended]
0
96. In Sec. 26.825, in paragraph (b) add remove the phrase
``Sec. Sec. 26.1, 26.3, 26.5, 26.7, 26.8, 26.9, 26.11, 26.51, 26.81,
26.121, 26.151, 26.181, 26.201, 26.823, and 26.825'' and add in its
place the phrase ``Sec. Sec. 26.1, 26.3, 26.5, 26.7, 26.8, 26.9,
26.11, 26.51, 26.81, 26.121, 26.151, 26.181, 26.201, 26.601, 26.823,
and 26.825''.
PART 30--RULES OF GENERAL APPLICABILITY TO DOMESTIC LICENSING OF
BYPRODUCT MATERIAL
0
97. The authority citation for part 30 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 11, 81, 161, 181,
182, 183, 184, 186, 187, 223, 234, 274 (42 U.S.C. 2014, 2111, 2201,
2231, 2232, 2233, 2234, 2236, 2237, 2273, 2282, 2021); Energy
Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C.
5841, 5842, 5846, 5851); 44 U.S.C. 3504 note.
0
98. In Sec. 30.4, revise the definition for ``Utilization facility''
to read as follows:
Sec. 30.4 Definitions.
* * * * *
Utilization facility means a utilization facility as defined in the
regulations contained in part 50 or part 53 of this chapter;
0
99. In Sec. 30.50, revise paragraph (c)(3) to read as follows:
Sec. 30.50 Reporting requirements.
* * * * *
(c) * * *
(3) The provisions of this section do not apply to licensees
subject to the notification requirements in Sec. Sec. 50.72 or 53.1630
of this chapter. They do apply to those part 50 licensees possessing
material licensed under this part, who are not subject to the
notification requirements in Sec. 50.72 of this chapter.
PART 40--DOMESTIC LICENSING OF SOURCE MATERIAL
0
100. The authority citation for part 40 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 62, 63, 64, 65, 69,
81, 83, 84, 122, 161, 181, 182, 183, 184, 186, 187, 193, 223, 234,
274, 275 (42 U.S.C. 2092, 2093, 2094, 2095, 2099, 2111, 2113, 2114,
2152, 2201, 2231, 2232, 2233, 2234, 2236, 2237, 2243, 2273, 2282,
2021, 2022); Energy Reorganization Act of 1974, secs. 201, 202, 206,
211 (42 U.S.C. 5841, 5842, 5846, 5851); Uranium Mill Tailings
Radiation Control Act of 1978, sec. 104 (42 U.S.C. 7914); 44 U.S.C.
3504 note.
0
101. In Sec. 40.60, revise paragraph (c)(3) to read as follows:
Sec. 40.60 Reporting requirements.
* * * * *
(c) * * *
(3) The provisions of this section do not apply to licensees
subject to the notification requirements in Sec. Sec. 50.72 or 53.1630
of this chapter. They do apply to those part 50 licensees possessing
material licensed under this part who are not subject to the
notification requirements in Sec. 50.72 of this chapter.
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
0
102. The authority citation for part 50 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 11, 101, 102, 103,
104, 105, 108, 122, 147, 149, 161, 181, 182, 183, 184, 185, 186,
187, 189, 223, 234 (42 U.S.C. 2014, 2131, 2132, 2133, 2134, 2135,
2138, 2152, 2167, 2169, 2201, 2231, 2232, 2233, 2234, 2235, 2236,
2237, 2239, 2273, 2282); Energy Reorganization Act of 1974, secs.
201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste
Policy Act of 1982, sec. 306(42 U.S.C. 10226); National
Environmental Policy Act of 1969 (42 U.S.C. 4332); 44 U.S.C. 3504
note; Sec. 109, Pub. L. 96-295, 94 Stat. 783.
0
103. In Sec. 50.47, revise paragraphs (a)(1) and (e) to read as
follows:
Sec. 50.47 Emergency plans.
(a)(1)(i) Except as provided in paragraph (d) of this section, no
initial operating license for a nuclear power reactor will be issued
under this part or under part 53 of this chapter unless a finding is
made by the NRC that there is reasonable assurance that adequate
protective measures can and will be taken in the event of a
radiological emergency. No finding under this section is necessary for
issuance of a renewed nuclear power reactor operating license.
(ii) No initial combined license under parts 52 or 53 of this
chapter will be issued unless a finding is made by the NRC that there
is reasonable assurance that adequate protective measures can and will
be taken in the event of a radiological emergency. No finding under
this section is necessary for issuance of a renewed combined license.
(iii) If an application for an early site permit under subpart A of
part 52 of this chapter includes complete and integrated emergency
plans under Sec. 52.17(b)(2)(ii) of this chapter or an application for
an early site permit under subpart H of part 53 of this chapter
includes complete and integrated emergency plans under Sec.
53.1146(b)(2)(ii) of this chapter, no early site permit will be issued
unless a finding is made by the NRC that the emergency plans provide
reasonable assurance that adequate protective measures can and will be
taken in the event of a radiological emergency.
(iv) If an application for an early site permit proposes major
features of the emergency plans under Sec. Sec. 52.17(b)(2)(i) or
53.1146(b)(2)(i) of this chapter, no early site permit will be issued
unless a finding is made by the NRC that the major features are
acceptable in accordance with the applicable standards of either Sec.
50.47 and appendix E to this part or the applicable requirements of
Sec. 50.160, within the scope of emergency preparedness matters
addressed in the major features.
* * * * *
(e) Notwithstanding the requirements of paragraph (b) of this
section and the provisions of Sec. 52.103 or Sec. 53.1452 of this
chapter, a holder of a combined license under part 52 or part 53 of
this chapter, as applicable, that is complying with the requirements of
Sec. 50.47(b) and appendix E to this part may not load fuel or operate
except as provided in accordance with appendix E to this part and Sec.
50.54(gg), and a holder of a combined license under part 52 or part 53
of this chapter that is complying with the requirements of Sec. 50.160
may not load fuel or operate except as provided in accordance with
Sec. 50.160(c)(2) and Sec. 50.54(gg).
* * * * *
0
104. In Sec. 50.54, revise paragraphs (q)(2), (q)(4), and (gg)(1)
introductory text to read as follows:
Sec. 50.54 Conditions of licenses.
* * * * *
(q) * * *
(2)(i) Except as provided in paragraph (q)(2)(ii) of this section,
a holder of a license under this part, or a combined license under
parts 52 or 53 of this chapter after the Commission makes the finding
under Sec. Sec. 52.103(g) or 53.1452(g)
[[Page 87039]]
of this chapter, as applicable, shall follow and maintain the
effectiveness of an emergency plan that meets the requirements in
appendix E to this part and, for nuclear power reactor licensees, the
planning standards of Sec. 50.47(b).
(ii) A holder of a license under this part for a non-power
production or utilization facility, a holder of a license under this
part or part 53 of this chapter for a small modular reactor or a non-
light-water reactor, or a holder of a combined license under parts 52
or 53 of this chapter after the Commission makes the finding under
Sec. Sec. 52.103(g) or 53.1452(g) of this chapter, as applicable, for
a small modular reactor or a non-light-water reactor, shall follow and
maintain the effectiveness of either an emergency plan that meets the
requirements in Sec. 50.160 or an emergency plan that meets the
requirements in appendix E to this part and, for nuclear power reactor
licensees, the planning standards of Sec. 50.47(b).
* * * * *
(4) The changes to a licensee's emergency plan that reduce the
effectiveness of the plan as defined in paragraph (q)(1)(iv) of this
section may not be implemented without prior approval by the NRC. A
licensee desiring to make such a change shall submit an application for
an amendment to its license. In addition to the filing requirements of
Sec. Sec. 50.90 and 50.91 or Sec. Sec. 53.1510 and 53.1515 of this
chapter, as applicable, the request must include all emergency plan
pages affected by that change and must be accompanied by a forwarding
letter identifying the change, the reason for the change, and the basis
for concluding that the licensee's emergency plan, as revised, will
continue to meet either the requirements in Sec. 50.160 or the
requirements in appendix E to this part and, for nuclear power reactor
licensees, the planning standards of Sec. 50.47(b).
* * * * *
(gg)(1) Notwithstanding Sec. Sec. 52.103 or 53.1452 of this
chapter, if following the conduct of the exercise required by paragraph
IV.f.2.a of appendix E to this part or Sec. 50.160(c)(2), as
applicable, FEMA identifies one or more deficiencies in the state of
offsite emergency preparedness, the holder of a combined license under
10 CFR part 52 or under 10 CFR part 53, as applicable, may operate at
up to 5 percent of rated thermal power only if the Commission finds
that the state of onsite emergency preparedness provides reasonable
assurance that adequate protective measures can and will be taken in
the event of a radiological emergency. The NRC will base this finding
on its assessment of the applicant's onsite emergency plans against the
pertinent standards in either Sec. 50.47 and appendix E to this part,
or Sec. 50.160, as applicable. Review of the applicant's emergency
plans will include the following standards with offsite aspects:
* * * * *
0
105. In Sec. 50.160, revise paragraphs (b)(3) and (c)(2) to read as
follows:
Sec. 50.160 Emergency preparedness for small modular reactors, non-
light-water reactors, and non-power production or utilization
facilities.
* * * * *
(b) * * *
(3) Emergency planning zone. For an applicant whose analysis
required by Sec. 50.33(g)(2) or Sec. 53.1109(g)(2) of this chapter
meets the criteria in Sec. 50.33(g)(2)(i) or Sec. 53.1109(g)(2)(i) of
this chapter, as applicable, determine and describe the boundary and
physical characteristics of the EPZ in the emergency plan.
* * * * *
(c) * * *
(2) A holder of a combined license issued under parts 52 or 53 of
this chapter before the Commission has made the finding under
Sec. Sec. 52.103(g) or 53.1452(g) of this chapter, as applicable, must
establish, implement, and maintain an emergency preparedness program
that meets the requirements of paragraph (b) of this section, as
described in the approved emergency plan and license, and conduct an
initial exercise to demonstrate this compliance within 2 years before
the scheduled date for initial loading of fuel (or, for a fueled
manufactured reactor, within 2 years before the scheduled date for
initiating the physical removal of any one of the independent physical
mechanisms to prevent criticality required under Sec. 53.620(d)(1) of
this chapter).
0
106. In appendix B to part 50, revise the first paragraph in the
Introduction section, the first paragraph of section III, Design
Control, and section IV, Procurement Document Control, to read as
follows:
Appendix B to Part 50--Quality Assurance Criteria for Nuclear Power
Plants and Fuel Reprocessing Plants
Introduction. Every applicant for a construction permit is
required by the provisions of Sec. 50.34 or Sec. 53.1309 of this
chapter to include in its preliminary safety analysis report a
description of the quality assurance program to be applied to the
design, fabrication, construction, and testing of the structures,
systems, and components of the facility. Every applicant for an
operating license is required by the provisions of Sec. 50.34 or
Sec. 53.1369 of this chapter to include, in its final safety
analysis report, information pertaining to the managerial and
administrative controls to be used to assure safe operation. Every
applicant for a combined license is required by the provisions of
Sec. Sec. 52.79 or 53.1416 of this chapter to include in its final
safety analysis report a description of the quality assurance
applied to the design, and to be applied to the fabrication,
construction, and testing of the structures, systems, and components
of the facility and to the managerial and administrative controls to
be used to assure safe operation. For applications submitted after
September 27, 2007, every applicant for an early site permit is
required by the provisions of Sec. Sec. 52.17 or 53.1146 of this
chapter to include in its site safety analysis report a description
of the quality assurance program applied to site activities related
to the design, fabrication, construction, and testing of the
structures, systems, and components of a facility or facilities that
may be constructed on the site. Every applicant for a design
approval is required by the provisions of Sec. Sec. 52.137 or
53.1209 of this chapter to include in its final safety analysis
report a description of the quality assurance program applied to the
design of the structures, systems, and components of the facility.
Every applicant for a design certification is required by the
provisions of Sec. Sec. 52.47 or 53.1239 of this chapter to include
in its final safety analysis report a description of the quality
assurance program applied to the design of the structures, systems,
and components of the facility. Every applicant for a manufacturing
license is required by the provisions of Sec. Sec. 52.157 or
53.1279 of this chapter to include in its final safety analysis
report a description of the quality assurance program applied to the
design, and to be applied to the manufacture of, the structures,
systems, and components of the reactor. Nuclear power plants and
fuel reprocessing plants include structures, systems, and components
that prevent or mitigate the consequences of postulated accidents
that could cause undue risk to the health and safety of the public.
This appendix establishes quality assurance requirements for the
design, manufacture, construction, and operation of those
structures, systems, and components. The pertinent requirements of
this appendix apply to all activities affecting the safety-related
functions of those structures, systems, and components; these
activities include designing, purchasing, fabricating, handling,
shipping, storing, cleaning, erecting, installing, inspecting,
testing, operating, maintaining, repairing, refueling, and
modifying.
* * * * *
III. Design Control
Measures shall be established to assure that applicable
regulatory requirements and the design bases, as defined in Sec.
50.2 and as specified in the license application, or the functional
design criteria, as defined in Sec. 53.020 of this chapter and as
specified in the license application, for those structures, systems,
and components to which this appendix applies are correctly
translated into specifications, drawings, procedures, and
instructions. These measures shall include provisions to assure that
appropriate quality
[[Page 87040]]
standards are specified and included in design documents and that
deviations from such standards are controlled. Measures shall also
be established for the selection and review for suitability of
application of materials, parts, equipment, and processes that are
essential to the safety-related functions of the structures, systems
and components.
* * * * *
IV. Procurement Document Control
Measures shall be established to assure that applicable
regulatory requirements, design bases or functional design criteria,
and other requirements which are necessary to assure adequate
quality are suitably included or referenced in the documents for
procurement of material, equipment, and services, whether purchased
by the applicant or by its contractors or subcontractors. To the
extent necessary, procurement documents shall require contractors or
subcontractors to provide a quality assurance program consistent
with the pertinent provisions of this appendix.
* * * * *
PART 51--ENVIRONMENTAL PROTECTION REGULATIONS FOR DOMESTIC
LICENSING AND RELATED REGULATORY FUNCTIONS
0
107. The authority citation for part 51 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 161, 193 (42 U.S.C.
2201, 2243); Energy Reorganization Act of 1974, secs. 201, 202 (42
U.S.C. 5841, 5842); National Environmental Policy Act of 1969 (42
U.S.C. 4332, 4334, 4335); Nuclear Waste Policy Act of 1982, secs.
144(f), 121, 135, 141, 148 (42 U.S.C. 10134(f), 10141, 10155, 10161,
10168); 44 U.S.C. 3504 note.
0
108. In Sec. 51.20, revise paragraphs (b)(1) and (2) to read as
follows:
Sec. 51.20 Criteria for and identification of licensing and
regulatory actions requiring environmental impact statements.
* * * * *
(b) * * *
(1) Issuance of a limited work authorization or a permit to
construct a nuclear power reactor, testing facility, or fuel
reprocessing plant under part 50 of this chapter, issuance of an early
site permit under part 52 of this chapter, or issuance of a limited
work authorization, construction permit, or early site permit under
part 53 of this chapter.
(2) Issuance or renewal of a full power or design capacity license
to operate a nuclear power reactor, testing facility, or fuel
reprocessing plant under parts 50 or 53 of this chapter, or a combined
license under parts 52 or 53 of this chapter.
* * * * *
0
109. In Sec. 51.22, revise paragraphs (c)(3) introductory text, (c)(9)
introductory text, (c)(12) introductory text, (c)(17), (c)(22) and (23)
to read as follows:
Sec. 51.22 Criterion for categorical exclusion; identification of
licensing and regulatory actions eligible for categorical exclusion or
otherwise not requiring environmental review.
* * * * *
(c) * * *
(3) Amendments to parts 20, 30, 31, 32, 33, 34, 35, 37, 39, 40, 50,
51, 52, 53, 54, 60, 61, 63, 70, 71, 72, 73, 74, 81, and 100 of this
chapter which relate to--
* * * * *
(9) Issuance of an amendment to a permit or license for a reactor
under part 50, part 52, or part 53 of this chapter that changes a
requirement or issuance of an exemption from a requirement, with
respect to installation or use of a facility component located within
the restricted area, as defined in part 20 of this chapter; or the
issuance of an amendment to a permit or license for a reactor under
part 50, part 52, or part 53 of this chapter that changes an inspection
or a surveillance requirement; provided that:
* * * * *
(12) Issuance of an amendment to a license under parts 50, 52, 53,
60, 61, 63, 70, 72, or 75 of this chapter relating solely to safeguards
matters (i.e., protection against sabotage or loss or diversion of
special nuclear material) or issuance of an approval of a safeguards
plan submitted under parts 50, 52, 53, 70, 72, and 73 of this chapter,
provided that the amendment or approval does not involve any
significant construction impacts. These amendments and approvals are
confined to--
* * * * *
(17) Issuance of an amendment to a permit or license under part 30,
part 40, part 50, part 52, part 53, or part 70 of this chapter which
deletes any limiting condition of operation or monitoring requirement
based on or applicable to any matter subject to the provisions of the
Federal Water Pollution Control Act.
* * * * *
(22) Issuance of a standard design approval under part 52 or part
53 of this chapter.
(23) The Commission finding for a combined license under Sec.
52.103(g) or Sec. 53.1452(g) of this chapter.
* * * * *
Sec. 51.26 [Amended]
0
110. In Sec. 51.26, in paragraph (d) remove the phrase ``under part
52'' and add in its place the phrase ``under 10 CFR parts 52 or 53,''.
0
111. In Sec. 51.30, revise paragraph (a) introductory text and
paragraphs (d) and (e) to read as follows:
Sec. 51.30 Environmental assessment.
(a) An environmental assessment for proposed actions, other than
those for a standard design certification under 10 CFR parts 52 or 53,
or a manufacturing license under 10 CFR parts 52 or 53, shall identify
the proposed action and include:
* * * * *
(d) An environmental assessment for a standard design certification
under subpart B of part 52 of this chapter, or under subpart H of part
53 of this chapter must identify the proposed action and will be
limited to the consideration of the costs and benefits of severe
accident mitigation design alternatives and the bases for not
incorporating severe accident mitigation design alternatives in the
design certification. An environmental assessment for an amendment to a
design certification will be limited to the consideration of whether
the design change which is the subject of the proposed amendment
renders a severe accident mitigation design alternative previously
rejected in the earlier environmental assessment to become cost
beneficial, or results in the identification of new severe accident
mitigation design alternatives, in which case the costs and benefits of
new severe accident mitigation design alternatives and the bases for
not incorporating new severe accident mitigation design alternatives in
the design certification must be addressed.
(e) An environmental assessment for a manufacturing license under
subpart F of part 52 of this chapter or under subpart H of part 53 of
this chapter must identify the proposed action and will be limited to
the consideration of the costs and benefits of severe accident
mitigation design alternatives and the bases for not incorporating
severe accident mitigation design alternatives in the manufacturing
license. An environmental assessment for an amendment to a
manufacturing license will be limited to consideration of whether the
design change which is the subject of the proposed amendment either
renders a severe accident mitigation design alternative previously
rejected in an environmental assessment to become cost beneficial, or
results in the identification of new severe accident mitigation design
alternatives, in which case the costs and benefits of new severe
accident mitigation design alternatives and the bases for not
incorporating new severe accident mitigation design alternatives in the
manufacturing license must be addressed. In either case, the
environmental assessment will not address the environmental impacts
[[Page 87041]]
associated with manufacturing the reactor under the manufacturing
license.
Sec. 51.31 [Amended]
0
112. In Sec. 51.31, in paragraph (a) remove the phrase ``under part
52'' and add in its place the phrase ``under parts 52 or 53''.
Sec. 51.32 [Amended]
0
113. In Sec. 51.32, in paragraphs (b)(1) and (3) remove the phrase
``of part 52 of this chapter'' and add in its place the phrase ``of
part 52 of this chapter or subpart H of part 53 of this chapter''.
Sec. 51.49 [Amended]
0
114. In Sec. 51.49, in paragraph (c) introductory text, remove the
phrase ``of part 52 of this chapter'' and add in its place the phrase
``of part 52 of this chapter or under subpart H of part 53 of this
chapter''.
Sec. 51.50 [Amended]
0
115. In Sec. 51.50, wherever it appears, remove the phrase ``in
accordance with Sec. 50.36b of this chapter'' and add in its place the
phrase ``in accordance with Sec. Sec. 50.36b or 53.1112 of this
chapter''.
Sec. 51.53 [Amended]
0
116. In Sec. 51.53, in paragraph (d) remove the phrase ``under Sec.
50.82 of this chapter'' and add in its place the phrase ``under
Sec. Sec. 50.82 or 53.1080 of this chapter''.
Sec. 51.54 [Amended]
0
117. In Sec. 51.54, in paragraph (a), remove the phrase ``of part 52
of this chapter'' and add in its place the phrase ``of part 52 of this
chapter or under subpart H of part 53 of this chapter''.
Sec. 51.55 [Amended]
0
118. In Sec. 51.55, in paragraph (a) remove the phrase ``of part 52 of
this chapter'' and add in its place the phrase ``of part 52 of this
chapter or under subpart H of part 53 of this chapter''.
0
119. In Sec. 51.58, revise paragraph (b) to read as follows:
Sec. 51.58 Environmental report--number of copies; distribution.
* * * * *
(b) Each applicant for a license to manufacture a nuclear power
reactor, or for an amendment to a license to manufacture, seeking
approval of the final design of the nuclear power reactor under subpart
F of part 52 of this chapter or under subpart H of part 53 of this
chapter, shall submit to the Commission an environmental report or any
supplement to an environmental report in the manner specified in
Sec. Sec. 52.3 or 53.040 of this chapter. The applicant shall maintain
the capability to generate additional copies of the environmental
report or any supplement to the environmental report for subsequent
distribution to parties and Boards in the NRC proceeding; Federal,
State, and local officials; and any affected Indian Tribes, in
accordance with written instructions issued by the Director, Office of
Nuclear Reactor Regulation.
0
120. In Sec. 51.77, revise paragraph (a) introductory text to read as
follows:
Sec. 51.77 Distribution of draft environmental impact statement.
(a) In addition to the distribution authorized by Sec. 51.74, a
copy of a draft environmental statement for a licensing action for a
production or utilization facility, except an action authorizing
issuance, amendment, or renewal of a license to manufacture a nuclear
power reactor pursuant to 10 CFR part 52, subpart F or 10 CFR part 53,
subparts H or I will also be distributed to:
* * * * *
Sec. 51.92 [Amended]
0
121. In Sec. 51.92, in paragraph (b), wherever it may appear, remove
the phrase ``10 CFR part 52'' and add in its place the phrase ``10 CFR
parts 52 or 53''.
Sec. 51.95 [Amended]
0
122. In Sec. 51.95, in paragraph (c) introductory text remove the
phrase ``under 10 CFR parts 52 or 54'' and add in its place the phrase
``under 10 CFR parts 52, 53, or 54''.
0
123. In Sec. 51.101, revise paragraph (a)(2) to read as follows:
Sec. 51.101 Limitations on actions.
(a) * * *
(2) Any action concerning the proposal taken by an applicant which
would--
(i) Have an adverse environmental impact, or
(ii) Limit the choice of reasonable alternatives that may be
grounds for denial of the license. In the case of an application
covered by Sec. Sec. 30.32(f), 40.31(f), 50.10(c), 53.1130, 70.21(f),
or 72.16 and 72.34 of this chapter, the provisions of this paragraph
will be applied in accordance with Sec. 30.33(a)(5), 40.32(e),
50.10(c), 53.1130, 70.23(a)(7), or 72.40(b) of this chapter, as
appropriate.
* * * * *
Sec. 51.103 [Amended]
0
124. In Sec. 51.103, in paragraph (a)(6) remove the phrase ``under 10
CFR 50.10'' and add in its place the phrase ``under Sec. Sec. 50.10 or
53.1130 of this chapter''.
0
125. In Sec. 51.105, revise paragraph (c)(1) introductory text to read
as follows:
Sec. 51.105 Public hearings in proceedings for issuance of
construction permits or early site permits; limited work
authorizations.
* * * * *
(c)(1) In addition to complying with the applicable provisions of
Sec. 51.104, in any proceeding for the issuance of a construction
permit for a nuclear power plant or an early site permit under parts 52
or 53 of this chapter, where the applicant requests a limited work
authorization under Sec. Sec. 50.10(d) or 53.1130 of this chapter, the
presiding officer will--
* * * * *
0
126. In Sec. 51.107, revise paragraphs (a) introductory text, (b)
introductory text, and (d)(1) introductory text to read as follows:
Sec. 51.107 Public hearings in proceedings for issuance of combined
licenses; limited work authorizations.
(a) In addition to complying with the applicable requirements of
Sec. 51.104, in a proceeding for the issuance of a combined license
for a nuclear power reactor under parts 52 or 53 of this chapter, the
presiding officer will:
* * * * *
(b) If a combined license application references an early site
permit, then the presiding officer in the combined license hearing must
not admit any contention proffered by any party on environmental issues
that have been accorded finality under Sec. Sec. 52.39 or 53.1188 of
this chapter, unless the contention:
* * * * *
(d)(1) In any proceeding for the issuance of a combined license
where the applicant requests a limited work authorization under
Sec. Sec. 50.10(d) or Sec. 53.1130(a) of this chapter, the presiding
officer, in addition to complying with any applicable provision of
Sec. 51.104, will:
* * * * *
0
127. Revise Sec. 51.108 to read as follows:
Sec. 51.108 Public hearings on Commission findings that inspections,
tests, analyses, and acceptance criteria of combined licenses are met.
In any public hearing requested under Sec. Sec. 52.103(b) or
53.1452(b) of this chapter, the Commission will not admit any
contentions on environmental issues, the adequacy of the environmental
impact statement for the combined license issued under subpart C of
part 52 of this chapter or under
[[Page 87042]]
subpart H of part 53 of this chapter, or the adequacy of any other
environmental impact statement or environmental assessment referenced
in the combined license application. The Commission will not make any
environmental findings in connection with the finding under Sec.
52.103(g) or Sec. 53.1452(g) of this chapter.
0
128. Add part 53, consisting of Sec. Sec. 53.000 through 53.9010, to
read as follows:
PART 53--RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK
FOR COMMERCIAL NUCLEAR PLANTS
Sec.
53.000 Purpose.
Subpart A--General Provisions
53.015 Scope.
53.020 Definitions.
53.030 Reserved.
53.040 Written communications.
53.050 Deliberate misconduct.
53.060 Employee protection.
53.070 Completeness and accuracy of information.
53.080 Specific exemptions.
53.090 Standards for review.
53.100 Jurisdictional limits.
53.110 Attacks and destructive acts.
53.115 Rights related to special nuclear material.
53.117 License suspension and rights of recapture.
53.120 Information collection requirements: OMB approval.
Subpart B--Technology-Inclusive Safety Requirements
53.210 Safety criteria for design-basis accidents.
53.220 Safety criteria for licensing-basis events other than design-
basis accidents.
53.230 Safety functions.
53.240 Licensing-basis events.
53.250 Defense in depth.
53.260 Normal operations.
53.270 Protection of plant workers.
Subpart C--Design and Analysis Requirements
53.400 Design features for licensing-basis events.
53.410 Functional design criteria for design-basis accidents.
53.415 Protection against external hazards.
53.420 Functional design criteria for licensing-basis events other
than design-basis accidents.
53.425 Design features and functional design criteria for normal
operations.
53.430 Design features and functional design criteria for protection
of plant workers.
53.440 Design requirements.
53.450 Analysis requirements.
53.460 Safety categorization and treatments.
53.470 Maintaining analytical safety margins used to justify
operational flexibilities.
53.480 Earthquake engineering.
Subpart D--Siting Requirements
53.500 General siting and siting assessment.
53.510 External hazards.
53.520 Site characteristics.
53.530 Population-related considerations.
53.540 Siting interfaces.
Subpart E--Construction and Manufacturing Requirements
53.600 Construction and manufacturing--scope and purpose.
53.605 Reporting of defects and noncompliance.
53.610 Construction.
53.620 Manufacturing.
Subpart F--Requirements for Operation
53.700 Operational objectives.
53.710 Maintaining capabilities and availability of structures,
systems, and components.
53.715 Maintenance, repair, and inspection programs.
53.720 Response to seismic events.
53.725 General staffing, training, personnel qualifications, and
human factors requirements.
53.726 Communications.
53.728 Completeness and accuracy of information.
53.730 Defining, fulfilling, and maintaining the role of personnel
in ensuring safe operations.
53.735 General exemptions.
53.740 Facility licensee requirements--General.
53.745 Operator license requirements.
53.760 Operator licensing.
53.765 Medical requirements.
53.770 Incapacitation because of disability or illness.
53.775 Applications for operators and senior operators.
53.780 Training, examination, and proficiency program.
53.785 Conditions of operator and senior operator licenses.
53.790 Issuance, modification, and revocation of operator and senior
operator licenses.
53.795 Expiration and renewal of operator and senior operator
licenses.
53.800 Facility licensees for self-reliant-mitigation facilities.
53.805 Facility licensee requirements related to generally licensed
reactor operators.
53.810 Generally licensed reactor operators.
53.815 Generally licensed reactor operator training, examination,
and proficiency programs.
53.820 Cessation of individual applicability.
53.830 Training and qualification of commercial nuclear plant
personnel.
53.845 Programs.
53.850 Radiation protection.
53.855 Emergency preparedness.
53.860 Security programs.
53.865 Quality assurance.
53.870 Integrity assessment programs.
53.875 Fire protection.
53.880 Inservice inspection and inservice testing.
53.910 Procedures and guidelines.
Subpart G--Decommissioning Requirements
53.1000 Scope and purpose.
53.1010 Financial assurance for decommissioning.
53.1020 Cost estimates for decommissioning.
53.1030 Annual adjustments to cost estimates for decommissioning.
53.1040 Methods for providing financial assurance for
decommissioning.
53.1045 Limitations on the use of decommissioning trust funds.
53.1050 NRC oversight.
53.1060 Reporting and recordkeeping requirements.
53.1070 Termination of license.
53.1075 Program requirements during decommissioning.
53.1080 Release of part of a commercial nuclear plant or site for
unrestricted use.
Subpart H--Licenses, Certifications, and Approvals
53.1100 Filing of application for licenses, certifications, or
approvals; oath or affirmation.
53.1101 Requirement for license.
53.1103 Combining applications and licenses.
53.1106 Elimination of repetition.
53.1109 Contents of applications; general information.
53.1112 Environmental conditions.
53.1115 Agreement limiting access to classified information.
53.1118 Ineligibility of certain applicants.
53.1120 Exceptions and exemptions from licensing requirements.
53.1121 Public inspection of applications.
53.1124 Relationship between sections.
53.1130 Limited work authorizations.
53.1140 Early site permits.
53.1143 Filing of applications.
53.1144 Contents of applications for early site permits; general
information.
53.1146 Contents of applications for early site permits; technical
information.
53.1149 Review of applications.
53.1155 Referral to the Advisory Committee on Reactor Safeguards.
53.1158 Issuance of early site permit.
53.1161 Extent of activities permitted.
53.1164 Duration of permit.
53.1167 Limited work authorization after issuance of early site
permit.
53.1170 Transfer of early site permit.
53.1173 Application for renewal.
53.1176 Criteria for renewal.
53.1179 Duration of renewal.
53.1182 Use of site for other purposes.
53.1188 Finality of early site permit determinations.
53.1200 Standard design approvals.
53.1203 Filing of applications.
53.1206 Contents of applications for standard design approvals;
general information.
53.1209 Contents of applications for standard design approvals;
technical information.
53.1210 Contents of applications for standard design approvals;
other application content.
53.1212 Standards for review of applications.
[[Page 87043]]
53.1215 Referral to the Advisory Committee on Reactor Safeguards.
53.1218 Staff approval of design.
53.1221 Finality of standard design approvals; information requests.
53.1230 Standard design certifications.
53.1233 Filing of applications.
53.1236 Contents of applications for standard design certifications;
general information.
53.1239 Contents of applications for standard design certifications;
technical information.
53.1241 Contents of applications for standard design certifications;
other application content.
53.1242 Review of applications.
53.1245 Referral to the Advisory Committee on Reactor Safeguards.
53.1248 Issuance of standard design certification.
53.1251 Duration of certification.
53.1254 Application for renewal.
53.1257 Criteria for renewal.
53.1260 Duration of renewal.
53.1263 Finality of standard design certifications.
53.1270 Manufacturing licenses.
53.1273 Filing of applications.
53.1276 Contents of applications for manufacturing licenses; general
information.
53.1279 Contents of applications for manufacturing licenses;
technical information.
53.1282 Contents of applications for manufacturing licenses; other
application content.
53.1285 Review of applications.
53.1286 Referral to the Advisory Committee on Reactor Safeguards.
53.1287 Issuance of manufacturing licenses.
53.1288 Finality of manufacturing licenses.
53.1291 Duration of manufacturing licenses.
53.1293 Transfer of manufacturing licenses.
53.1295 Renewal of manufacturing licenses.
53.1300 Construction permits.
53.1306 Contents of applications for construction permits; general
information.
53.1309 Contents of applications for construction permits; technical
information.
53.1312 Contents of applications for construction permits; other
application content.
53.1315 Review of applications.
53.1318 Finality of referenced NRC approvals, permits, and
certifications.
53.1324 Referral to the Advisory Committee on Reactor Safeguards.
53.1327 Authorization to conduct limited work authorization
activities.
53.1330 Exemptions, departures, and variances.
53.1333 Issuance of construction permits.
53.1336 Finality of construction permits.
53.1342 Duration of construction permits.
53.1345 Transfer of construction permits.
53.1348 Termination of construction permits.
53.1360 Operating licenses.
53.1366 Contents of applications for operating licenses; general
information.
53.1369 Contents of applications for operating licenses; technical
information.
53.1372 Contents of applications for operating licenses; other
application content.
53.1375 Review of applications.
53.1381 Referral to the Advisory Committee on Reactor Safeguards.
53.1384 Exemptions, departures, and variances.
53.1387 Issuance of operating licenses.
53.1390 Backfitting of operating licenses.
53.1396 Duration of operating licenses.
53.1399 Transfer of an operating license.
53.1402 Application for renewal.
53.1405 Continuation of an operating license.
53.1410 Combined licenses.
53.1413 Contents of applications for combined licenses; general
information.
53.1416 Contents of applications for combined licenses; technical
information.
53.1419 Contents of applications for combined licenses; other
application content.
53.1422 Review of applications.
53.1425 Finality of referenced NRC approvals.
53.1431 Referral to the Advisory Committee on Reactor Safeguards.
53.1434 Authorization to conduct limited work authorization
activities.
53.1437 Exemptions, departures, and variances.
53.1440 Issuance of combined licenses.
53.1443 Finality of combined licenses.
53.1449 Inspection during construction.
53.1452 Operation under a combined license.
53.1455 Duration of combined license.
53.1456 Transfer of a combined license.
53.1458 Application for renewal.
53.1461 Continuation of combined license.
53.1470 Standardization of commercial nuclear plant designs:
licenses to construct and operate nuclear power reactors of
identical design at multiple sites.
Subpart I--Maintaining and Revising Licensing-Basis Information
53.1500 Licensing-basis information.
53.1502 Specific terms and conditions of licenses.
53.1505 Changes to licensing-basis information requiring prior NRC
approval.
53.1510 Application for amendment of license.
53.1515 Public notices; State consultation.
53.1520 Issuance of amendment.
53.1525 Revising certification information within a design
certification rule.
53.1530 Revising design information within a manufacturing license.
53.1535 Amendments during construction.
53.1540 Updating licensing-basis information and determining the
need for NRC approval.
53.1545 Updating Final Safety Analysis Reports.
53.1550 Evaluating changes to facility as described in Final Safety
Analysis Reports.
53.1560 Updating program documents included in licensing-basis
information.
53.1565 Evaluating changes to programs included in licensing-basis
information.
53.1570 Transfer of licenses.
53.1575 Termination of licenses.
53.1580 Information requests.
53.1585 Revocation, suspension, modification of licenses and
approvals for cause.
53.1590 Backfitting.
53.1595 Renewal.
Subpart J--Reporting and Other Administrative Requirements
53.1600 General information.
53.1610 Unfettered access for inspections.
53.1620 Maintenance of records, making of reports.
53.1630 Immediate notification requirements for operating commercial
nuclear plants.
53.1640 Licensee event report system.
53.1645 Reports of radiation exposure to members of the public.
53.1650 Facility information and verification.
53.1660 Financial requirements.
53.1670 Financial qualifications.
53.1680 Annual financial reports.
53.1690 Licensee's change of status; financial qualifications.
53.1700 Creditor regulations.
53.1710 Financial protection.
53.1720 Insurance required to stabilize and decontaminate plant
following an accident.
53.1730 Financial protection requirements.
Subpart M--Enforcement
53.9000 Violations.
53.9010 Criminal penalties.
Authority: Atomic Energy Act of 1954, secs. 11, 101, 103, 108,
122, 147, 161, 181, 182, 183, 184, 185, 186, 187, 189, 223, 234 (42
U.S.C. 2014, 2131, 2132, 2133, 2134, 2135, 2138, 2152, 2167, 2169,
2201, 2231, 2232, 2233, 2234, 2235, 2236, 2237, 2239, 2273, 2282);
Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42
U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste Policy Act of 1982,
sec. 306 (42 U.S.C. 10226); National Environmental Policy Act of
1969 (42 U.S.C. 4332); 44 U.S.C. 3504 note; Sec. 109, Pub. L. 96-
295, 94 Stat. 783; Pub. L. 115-439, 132 Stat. 5571.
Sec. 53.000 Purpose.
This part provides an optional technology-inclusive, performance-
based framework for the issuance, amendment, renewal, and termination
of licenses, permits, certifications, and approvals for commercial
nuclear plants licensed under section 103 of the Atomic Energy Act of
1954, as amended (the Act)(68 Stat. 919), and Title II of the Energy
Reorganization Act of 1974, as amended (88 Stat. 1242). Also, this part
gives notice to all persons who knowingly provide to any holder of or
applicant for an approval, certification, permit, or license, or to a
contractor, subcontractor, or consultant of any of
[[Page 87044]]
them, components, equipment, materials, or other goods or services that
relate to the activities of a holder of or applicant for an approval,
certification, permit, or license, subject to this part, that they may
be individually subject to U.S. Nuclear Regulatory Commission
enforcement action for violation of the provisions in Sec. 53.050.
Subpart A--General Provisions
Sec. 53.015 Scope.
Subpart A provides general provisions applicable to all applicants
and licensees subject to the rules of this part.
Sec. 53.020 Definitions.
For the purpose of this part:
Anticipated event sequence means event sequences expected to occur
one or more times during the life of a commercial nuclear plant.
Anticipated event sequences take into account the expected response of
all structures, systems, and components (SSCs) within the plant,
regardless of safety classification.
Applicant means a person applying for a license, permit, or other
form of Commission permission or approval under this part.
Certified fuel handler means, for a commercial nuclear plant,
either--
(1) A non-licensed operator who has qualified in accordance with a
fuel handler training program approved by the Commission; or
(2) A non-licensed operator who demonstrates compliance with the
following criteria:
(i) Has qualified in accordance with a fuel handler training
program that demonstrates compliance with the same requirements as
training programs for non-licensed operators required by Sec. 53.830,
and
(ii) Is responsible for decisions on--
(A) Safe conduct of decommissioning activities,
(B) Safe handling and storage of spent fuel, and
(C) Appropriate response to plant emergencies.
Combined license (COL) means a combined construction permit (CP)
and operating license (OL) with conditions for a commercial nuclear
plant issued under this part.
Commercial nuclear plant means a facility consisting of one or more
commercial nuclear reactors and associated co-located support
facilities, including the collection of buildings, radionuclide
sources, and SSCs for which a license, certification, or approval is
being sought under this part, that is or will be used for producing
power for commercial electric power or other commercial purposes. For
the purposes of requirements in this part that reference requirements
in part 50 of this chapter, a commercial nuclear plant is equivalent to
a nuclear power plant.
Commercial nuclear reactor means an apparatus, other than an atomic
weapon, designed or used to sustain nuclear fission. For the purposes
of requirements in this part that reference requirements in 10 CFR part
50, a commercial nuclear reactor is equivalent to a nuclear reactor as
defined in 10 CFR 50.2.
Commission means the U.S. Nuclear Regulatory Commission (NRC) or
its duly authorized representatives.
Consensus code or standard means any technical standard that is--
(1) Developed or adopted by a voluntary consensus standard body
under procedures that assure that persons having interests within the
scope of the standard that are affected by the provisions of the
standard have reached substantial agreement on its adoption;
(2) Formulated in a manner that afforded an opportunity for diverse
views to be considered; and
(3) Designated by the standards body as a consensus code or
standard.
Construction means the activities in paragraph (1) below and does
not mean the activities in paragraph (2) below.
(1) Activities constituting construction are those activities
credited or relied upon for demonstrating compliance with the safety
criteria defined in subpart B of this part which are conducted on-site
to build the commercial nuclear plant, including the driving of piles;
subsurface preparation; placement of backfill, concrete, or permanent
retaining walls within an excavation; installation of foundations; or
in-place assembly, erection, fabrication, or testing, which are for--
(i) Safety-related (SR) and non-safety-related but safety-
significant (NSRSS) SSCs of a facility;
(ii) SSCs necessary to comply with 10 CFR part 73; or
(iii) Onsite emergency facilities necessary to comply with Sec.
53.855.
(2) Construction does not include--
(i) Changes for temporary use of the land for public recreational
purposes;
(ii) Site exploration, including necessary borings to determine
foundation conditions or other preconstruction monitoring to establish
background information related to the suitability of the site, the
environmental impacts of construction or operation, or the protection
of environmental values;
(iii) Preparation of a site for construction of a facility,
including clearing of the site, grading, installation of drainage,
erosion, and other environmental mitigation measures, and construction
of temporary roads and borrow areas;
(iv) Erection of fences and other access control measures;
(v) Excavation;
(vi) Erection of support buildings (such as construction equipment
storage sheds, warehouse and shop facilities, utilities, concrete
mixing plants, docking and unloading facilities, and office buildings)
for use in connection with the construction of the facility;
(vii) Building of service facilities (such as paved roads, parking
lots, railroad spurs, exterior utility and lighting systems, potable
water systems, sanitary sewage treatment facilities, and transmission
lines);
(viii) Procurement or fabrication of components or portions of the
proposed facility occurring at locations other than the final, in-place
location at the facility; or
(ix) Manufacture of a nuclear power reactor under a manufacturing
license (ML) under subpart H of this part to be installed at the
proposed site and to be part of the proposed facility.
Custom combined license (custom COL) means a COL that does not
reference a standard design approval or design certification.
Decommission or decommissioning means to remove a plant or site
safely from service and reduce residual radioactivity to a level that
permits--
(1) Release of the property for unrestricted use and termination of
the license; or
(2) Release of the property under restricted conditions and
termination of the license.
Defense in depth means inclusion of two or more independent and
redundant layers of defense in the design of a facility and its
operating procedures to compensate for uncertainties such that no
single layer of defense, no matter how robust, is exclusively relied
upon. Defense in depth includes, but is not limited to, the use of
access controls, physical barriers, redundant and diverse safety
functions, and emergency response measures.
Design-basis accidents (DBAs) means postulated event sequences that
are used to set functional design criteria and performance objectives
for the design of SR SSCs through deterministic analyses. Design-basis
accidents are a type of licensing-basis event and are based on the
capabilities and reliabilities of SR SSCs needed to mitigate and
prevent event sequences, respectively.
Design-basis external hazard level means the level of severity or
intensity
[[Page 87045]]
of an external hazard for which the SR SSCs are protected against or
designed to withstand without losing their capability to perform their
safety functions.
Design features means the active and passive SSCs and the inherent
characteristics of those SSCs that contribute to limiting the total
effective dose equivalent to individual members of the public during
normal operations and prevent or mitigate the consequences of event
sequences.
Electric utility means any entity that generates or distributes
electricity and that recovers the cost of this electricity, either
directly or indirectly, through rates established by the entity itself
or by a separate regulatory authority. Investor-owned utilities,
including generation or distribution subsidiaries, public utility
districts, municipalities, rural electric cooperatives, and State and
Federal agencies, including associations of any of the foregoing, are
included within the meaning of ``electric utility.''
Event sequence means a postulated initiating event defined for a
set of initial plant conditions followed by system, safety function,
and operator successes or failures, and terminating in a specified end
state depending on the system, safety function, and operator successes
and failures (e.g., prevention of release of radioactive material or
release in one of the reactor-specific release categories). An event
sequence may include many unique variations of events that are similar
in terms of results or end states.
Exclusion area means that area surrounding the reactor, in which
the reactor licensee has the authority to determine all activities
including exclusion or removal of personnel and property from the area.
This area may be traversed by a highway, railroad, or waterway,
provided these are not so close to the facility as to interfere with
normal operations of the facility and provided appropriate and
effective arrangements are made to control traffic on the highway,
railroad, or waterway, in case of emergency, to protect the public
health and safety. Residence within the exclusion area must normally be
prohibited. In any event, residents must be subject to ready removal in
case of necessity. Activities unrelated to operation of the reactor may
be permitted in an exclusion area under appropriate limitations,
provided that no significant hazards to the public health and safety
will result.
Fission product release means the amount and composition of
radioactive material released to the environment, after accounting for
any retention of radionuclides provided by reactor design features.
Fuel means special nuclear material (SNM) or source material,
discrete elements that physically contain SNM or source material, and
homogeneous mixtures that contain SNM or source material, intended to
or used to create power in a commercial nuclear plant.
Functional design criteria means metrics for the performance of
SSCs. For SR SSCs, these criteria define performance metrics necessary
to demonstrate compliance with the safety criteria in Sec. 53.210. For
NSRSS SSCs, these criteria define performance metrics necessary to
demonstrate compliance with the safety criteria in Sec. 53.220.
License, when used in the context of a facility, means a limited
work authorization, CP, OL, early site permit, COL, or ML under this
part, or a renewed license issued by the Commission under this part.
When used in the context of a license authorizing an individual to
manipulate the controls of a facility, license means a license issued
by the Commission to perform the function of an operator, senior
operator, or generally licensed reactor operator as defined in this
part.
Licensee means a person who is authorized to conduct activities
under a license issued under this part by the Commission.
Licensing-basis events means a collection of event sequences
considered in the design and licensing of the commercial nuclear plant.
Licensing-basis events are unplanned events and include anticipated
event sequences, unlikely event sequences, very unlikely event
sequences, and DBAs.
Licensing-basis information means the information contained in
regulations, orders, licenses, certifications, or approvals issued by
the NRC for a commercial nuclear plant licensed under this part and
that information submitted to the NRC by an applicant or licensee in a
Safety Analysis Report, program description, or other licensing-related
document required under this part.
Low-population zone means the area immediately surrounding the
exclusion area which contains residents, the total number and density
of which are such that there is a reasonable probability that
appropriate protective measures could be taken on their behalf in the
event of a serious accident. A permissible population density or total
population within this zone is not included in this definition because
the situation may vary from case to case. Whether a specific number of
people can, for example, be evacuated from a specific area or
instructed to take shelter on a timely basis, will depend on many
factors such as location, number and size of highways, scope and extent
of advance planning, and actual distribution of residents within the
area.
Major decommissioning activity means, for a commercial nuclear
plant, any activity that results in permanent removal of major
radioactive components, permanently modifies the structure of the
containment, if applicable, or results in dismantling components for
shipment containing greater than class C waste in accordance with 10
CFR 61.55.
Major feature of the emergency plans means an aspect of those plans
necessary to:
(1) Address in whole or part either one or more of the 16 standards
in 10 CFR 50.47(b) or the requirements of 10 CFR 50.160(b), as
applicable; or
(2) Describe the emergency planning zones as required in Sec.
53.1109(g).
Manufactured reactor means the essential portions of a nuclear
reactor that are manufactured under an ML and subsequently transported
and incorporated into a commercial nuclear plant under a COL.
Manufacturing license means a license issued under this part that
authorizes the manufacture of manufactured reactors but not its
construction, installation, or operation.
Non-Safety-Related but Safety-Significant (NSRSS) SSCs means those
SSCs which are not SR but are relied on to achieve adequate defense in
depth or perform risk-significant functions and warrant special
treatment.
Non-Safety-Significant SSCs means those SSCs that are not SR or
NSRSS, are not relied on to achieve adequate defense in depth or to
perform risk-significant functions, and do not warrant special
treatment.
Person means--
(1) any individual, corporation, partnership, firm, association,
trust, estate, public or private institution, group, government agency
other than the Commission or the Department, except that the Department
shall be considered a person to the extent that its facilities are
subject to the licensing and related regulatory authority of the
Commission pursuant to section 202 of the ERA, any State or any
political subdivision of, or any political entity within a State, any
foreign government or nation or any political subdivision of any such
government or nation, or other entity; and
(2) any legal successor, representative, agent, or agency of the
foregoing.
[[Page 87046]]
Population center distance means the distance from the reactor to
the nearest boundary of a densely populated center containing more than
about 25,000 residents.
Probabilistic risk assessment means a quantitative assessment of
the risk associated with plant operation and maintenance that is
measured in terms of event sequence occurrence frequencies and
consequences.
Programmatic controls means administrative procedures that govern
human action in implementing programs and operating, monitoring, and
maintaining SSCs and equipment of a commercial nuclear plant.
Programmatic controls considered to be licensing basis information are
specified in an application for a requested activity of the Commission.
Quality assurance (QA) means all those planned and systematic
actions necessary to ensure that a structure, system, or component will
perform satisfactorily in service. Quality assurance includes quality
control, which comprises those QA actions related to the physical
characteristics of a material, structure, component, or system which
provide a means to control the quality of the material, structure,
component, or system to predetermined requirements.
Safety criteria means performance-based metrics that establish a
level of safety provided in requirements in Sec. Sec. 53.210 and
53.220.
Safety-related structures, systems, or components means those SSCs
that are relied upon to demonstrate compliance with the safety criteria
in Sec. 53.210 and warrant special treatment.
Small modular reactor means a power reactor, which may be of
modular design as defined in 10 CFR 52.1, licensed under this part to
produce heat energy up to 1,000 megawatts thermal per module.
Site characteristics means the actual physical, environmental, and
demographic features of a site. Site characteristics are specified in
an early site permit or in a Preliminary or Final Safety Analysis
Report for a limited work authorization, CP, or COL, as applicable.
Site parameters are the postulated physical, environmental, and
demographic features of an assumed site. Site parameters are specified
in a standard design approval, standard design certification, or ML.
Source material means source material as defined in subsection 11z.
of the Atomic Energy Act of 1954, as amended, (the Act) and in the
regulations contained in part 40 of this chapter.
Special nuclear material (SNM) means:
(1) Plutonium, uranium-233, uranium enriched in the isotope-233 or
in the isotope-235, and any other material which the Commission,
pursuant to the provisions of section 51 of the Act, determines to be
SNM, but does not include source material; or
(2) Any material artificially enriched by any of the foregoing, but
does not include source material.
Special treatment means those requirements, such as QA and
programmatic controls, that ensure that SR and NSRSS SSCs will provide
defense in depth or perform risk-significant functions. The
requirements also ensure that the SSCs will perform under the service
conditions and with the reliability assumed in the analysis performed
under Sec. 53.450 to demonstrate compliance with the safety criteria
in Sec. Sec. 53.210 and 53.220.
Standard design means a design which is sufficiently detailed and
complete to support certification or approval in accordance with
subpart H of this part, and which is usable under of this part for a
multiple number of units or at a multiple number of sites without
reopening or repeating the review.
Standard design approval or design approval means an NRC staff
approval, issued under subpart H of this part, of a final standard
design for a commercial nuclear plant. The approval may be for either
the final design for the entire reactor facility or the final design of
major portions thereof.
Standard design certification or design certification means a
Commission approval, issued under subpart H of this part, of a final
standard design for a nuclear power facility. This design may be
referred to as a certified standard design.
Total effective dose equivalent means the sum of the effective dose
equivalent (for external exposures) and the committed effective dose
equivalent (for internal exposures).
Utilization facility means any commercial nuclear reactor other
than one designed or used primarily for the formation of plutonium or
uranium-233.
Unlikely event sequences means event sequences that are not
expected to occur in the life of a commercial nuclear plant and are
less likely than anticipated event sequences, but are infrequent rather
than rare. Unlikely event sequences take into account the expected
response of all SSCs within the plant regardless of safety
classification.
Very unlikely event sequences means event sequences that are not
expected to occur in the life of a commercial nuclear plant, are less
likely than an unlikely event sequence, and are rare. Very unlikely
event sequences take into account the expected response of all SSCs
within the plant regardless of safety classification.
Sec. 53.030 [Reserved]
Sec. 53.040 Written communications.
(a) General requirements. All correspondence, reports,
applications, and other written communications from the applicant or
licensee to the NRC concerning the regulations in this part or
individual license conditions must be sent either by mail addressed:
ATTN: Document Control Desk, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001; by hand delivery to the NRC's offices at
11555 Rockville Pike, Rockville, Maryland, between the hours of 8:15
a.m. and 4 p.m. eastern time; or, where practicable, by electronic
submission, for example, via Electronic Information Exchange, email, or
CD-ROM. Electronic submissions must be made in a manner that enables
the NRC to receive, read, authenticate, distribute, and archive the
submission, and process and retrieve it a single page at a time.
Detailed guidance on making electronic submissions can be obtained by
visiting the NRC's website at https://www.nrc.gov/site-help/e-submittals.html; by email to [email protected]; or by writing the
Office of the Chief Information Officer, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001. The guidance discusses, among
other topics, the formats the NRC can accept, the use of electronic
signatures, and the treatment of nonpublic information. If the
communication is on paper, the signed original must be sent. If a
submission due date falls on a Saturday, Sunday, or Federal holiday,
the next Federal working day becomes the official due date.
(b) Distribution requirements. Copies of all correspondence,
reports, and other written communications concerning the regulations in
this part or individual license conditions, or the terms and conditions
of an early site permit or standard design approval, must be submitted
to the persons listed below (addresses for the NRC Regional Offices are
listed in appendix D to 10 CFR part 20).
(1) Applications for amendment of permits and licenses, reports,
and other communications. All written communications (including
responses to generic letters, bulletins, information notices,
regulatory information summaries, inspection reports, and
[[Page 87047]]
miscellaneous requests for additional information) that are required of
holders of licenses, permits, and design approvals issued pursuant to
this part, must be submitted as follows, except as otherwise specified
in paragraphs (b)(2) through (7) of this section: to the NRC's Document
Control Desk (if on paper, the signed original), with a copy to the
appropriate Regional Office, and a copy to the appropriate NRC Resident
Inspector if one has been assigned to the site of the facility or the
place of manufacture of a reactor licensed under this part.
(2) Applications for permits and licenses, and amendments to
applications. Applications for licenses, permits, and design approvals
and amendments to any of these types of applications must be submitted
to the NRC's Document Control Desk, with a copy to the appropriate
Regional Office, and a copy to the appropriate NRC Resident Inspector
if one has been assigned to the facility or the place of manufacture of
a reactor licensed under this part, except as otherwise specified in
paragraphs (b)(3) through (9) of this section. If the application or
amendment is on paper, the submission to the Document Control Desk must
be the signed original.
(3) Acceptance review application. Written communications required
for an application for determination of suitability for docketing must
be submitted to the NRC's Document Control Desk, with a copy to the
appropriate Regional Office. If the communication is on paper, the
submission to the Document Control Desk must be the signed original.
(4) Security plan and related submissions. Written communications,
as defined in paragraphs (b)(4)(i) through (v) of this section, must be
submitted to the NRC's Document Control Desk, with a copy to the
appropriate Regional Office. If the communication is on paper, the
submission to the Document Control Desk must be the signed original.
Submissions should include the following as appropriate:
(i) Physical security plan;
(ii) Safeguards contingency plan;
(iii) Cybersecurity plan;
(iv) Change to security plan, guard training and qualification
plan, safeguards contingency plan, or cybersecurity plan made without
prior Commission approval under Sec. 53.1565; and
(v) Application for amendment of physical security plan, guard
training and qualification plan, safeguards contingency plan, or
cybersecurity plan under Sec. 53.1510.
(5) Emergency plan and related submissions. Written communications
as defined in paragraphs (b)(5)(i) through (iii) of this section must
be submitted to the NRC's Document Control Desk, with a copy to the
appropriate Regional Office, and a copy to the appropriate NRC Resident
Inspector if one has been assigned to the site of the facility. If the
communication is on paper, the submission to the Document Control Desk
must be the signed original. Submissions should include the following
as appropriate:
(i) Emergency plan;
(ii) Change to an emergency plan under Sec. 53.1565; and
(iii) Emergency implementing procedures under Sec. 53.855.
(6) Updated Final Safety Analysis Report. An Updated Final Safety
Analysis Report or replacement pages under Sec. 53.1545 must be
submitted to the NRC's Document Control Desk, with a copy to the
appropriate Regional Office, and a copy to the appropriate NRC Resident
Inspector if one has been assigned to the site of the facility or the
place of manufacture of a reactor licensed under this part. Paper copy
submissions may be made using replacement pages; however, if a licensee
chooses to use electronic submission, all subsequent updates or
submissions must be performed electronically on a total replacement
basis. If the communication is on paper, the submission to the Document
Control Desk must be the signed original. If the communications are
submitted electronically, see Guidance for Electronic Submissions to
the Commission.
(7) Quality assurance related submissions. (i) A change to the
Safety Analysis Report QA program description under Sec. 53.1565, or a
change to a licensee's NRC-accepted QA topical report under Sec.
53.1565, must be submitted to the NRC's Document Control Desk, with a
copy to the appropriate Regional Office, and a copy to the appropriate
NRC Resident Inspector if one has been assigned to the site of the
facility or the place of manufacture of a reactor licensed under this
part. If the communication is on paper, the submission to the Document
Control Desk must be the signed original.
(ii) A change to an NRC-accepted QA topical report from non-
licensees (i.e., architect/engineers, nuclear steam supply system
suppliers, fuel suppliers, constructors, etc.) must be submitted to the
NRC's Document Control Desk. If the communication is on paper, the
signed original must be sent.
(8) Certification of permanent cessation of operations. The
licensee's certification of permanent cessation of operations, under
subpart G of this part, must state the date on which operations have
ceased or will cease, and must be submitted to the NRC's Document
Control Desk. This submission must be under oath or affirmation.
(9) Certification of permanent fuel removal. The licensee's
certification of permanent fuel removal, under subpart G of this part,
must state the date on which the fuel was removed from the reactor
vessel and the disposition of the fuel, and must be submitted to the
NRC's Document Control Desk. This submission must be under oath or
affirmation.
(c) Form of communications. All paper copies submitted to
demonstrate compliance with the requirements set forth in paragraph (b)
of this section must be typewritten, printed, or otherwise reproduced
in permanent form on unglazed paper. Exceptions to these requirements
imposed on paper submissions may be granted for the submission of
micrographic, photographic, or similar forms.
(d) Regulation governing submission. Licensees, applicants, and
holders of standard design approvals submitting correspondence,
reports, and other written communications under the regulations of this
part are requested but not required to cite whenever practical, in the
upper right corner of the first page of the submission, the specific
regulation or other basis requiring submission.
Sec. 53.050 Deliberate misconduct.
(a) Any licensee or applicant for a license; holder of or applicant
for a standard design approval; applicant for a standard design
certification; employee of a licensee, holder of a standard design
approval, or applicant for a license, standard design approval, or
standard design certification; or any contractor (including a supplier
or consultant), subcontractor, employee of a contractor or
subcontractor of any licensee or applicant for a license, holder of or
applicant for a standard design approval, or applicant for a standard
design certification, who knowingly provides to any licensee,
applicant, contractor, or subcontractor, any components, equipment,
materials, or other goods or services that relate to a licensee's or
applicant's activities in this part, may not--
(1) Engage in deliberate misconduct that causes or would have
caused, if not detected, a licensee or applicant to be in violation of
any rule, regulation, or order; or any term, condition, or
[[Page 87048]]
limitation of any license issued by the Commission; or
(2) Deliberately submit to the NRC, a licensee, an applicant, or a
licensee's or applicant's contractor or subcontractor, information that
the person submitting the information knows to be incomplete or
inaccurate in some respect material to the NRC.
(b) A person who violates paragraph (a)(1) or (2) of this section
may be subject to enforcement action in accordance with the procedures
in subpart B of 10 CFR part 2.
(c) For the purposes of paragraph (a)(1) of this section,
deliberate misconduct by a person means an intentional act or omission
that the person knows--
(1) Would cause a licensee or applicant to be in violation of any
rule, regulation, or order; or any term, condition, or limitation, of
any license issued by the Commission; or
(2) Constitutes a violation of a requirement, procedure,
instruction, contract, purchase order, or policy of a licensee,
applicant, contractor, or subcontractor.
Sec. 53.060 Employee protection.
(a) Discrimination by a Commission licensee, holder of a standard
design approval, an applicant for a license, standard design
certification, or standard design approval, a contractor or
subcontractor of a Commission licensee, holder of a standard design
approval, applicant for a license, standard design certification, or
standard design approval, against an employee for engaging in certain
protected activities is prohibited. Discrimination includes discharge
and other actions that relate to compensation, terms, conditions, or
privileges of employment. The protected activities are established in
section 211 of the Energy Reorganization Act of 1974, as amended, and
in general are related to the administration or enforcement of a
requirement imposed under the Act or the Energy Reorganization Act of
1974, as amended.
(1) The protected activities include but are not limited to--
(i) Providing the Commission or his or her employer information
about alleged violations of either of the statutes named in paragraph
(a) of this section or possible violations of requirements imposed
under either of those statutes;
(ii) Refusing to engage in any practice made unlawful under either
of the statutes named in paragraph (a) of this section or under these
requirements if the employee has identified the alleged illegality to
the employer;
(iii) Requesting the NRC to institute action against his or her
employer for the administration or enforcement of these requirements;
(iv) Testifying in any Commission proceeding, or before Congress,
or at any Federal or State proceeding regarding any provision (or
proposed provision) of either of the statutes named in paragraph (a) of
this section; and
(v) Assisting or participating in, or being about to assist or
participate in, these activities.
(2) These activities are protected even if no formal proceeding is
actually initiated as a result of the employee assistance or
participation.
(3) This section has no application to any employee alleging
discrimination prohibited by this section who, acting without direction
from his or her employer (or the employer's agent), deliberately causes
a violation of any requirement of the Energy Reorganization Act of
1974, as amended, or the Act.
(b) Any employee who believes that they have been discharged or
otherwise discriminated against by any person for engaging in protected
activities specified in paragraph (a)(1) of this section may seek a
remedy for the discharge or discrimination through an administrative
proceeding in the Department of Labor. The administrative proceeding
must be initiated within 180 days after an alleged violation occurs.
The employee may do this by filing a complaint alleging the violation
with the Department of Labor, Wage and Hour Division. The Department of
Labor may order reinstatement, back pay, and compensatory damages.
(c) A violation of paragraph (a), (e), or (f) of this section by a
Commission licensee, a holder of a standard design approval, an
applicant for a Commission license, standard design certification, or a
standard design approval, or a contractor or subcontractor of a
Commission licensee, holder of a standard design approval, or any
applicant may be grounds for--
(1) Denial, revocation, or suspension of the license or standard
design approval;
(2) Withdrawal or revocation of a proposed or final standard design
certification;
(3) Imposition of a civil penalty on the licensee, holder of a
standard design approval, or applicant (including an applicant for a
standard design certification under this part following Commission
adoption of final design certification rule) or a contractor or
subcontractor of the licensee, holder of a standard design approval, or
applicant; or
(4) Other enforcement action.
(d) Actions taken by an employer, or others, which adversely affect
an employee may be predicated upon nondiscriminatory grounds. The
prohibition applies when the adverse action occurs because the employee
has engaged in protected activities. An employee's engagement in
protected activities does not automatically render him or her immune
from discharge or discipline for legitimate reasons or from adverse
action dictated by nonprohibited considerations.
(e)(1) Each holder or applicant for a license or design approval,
must prominently post the revision of NRC Form 3, ``Notice to
Employees,'' referenced in Sec. 19.11(e)(1) of this chapter. This form
must be posted at locations sufficient to permit employees protected by
this section to observe a copy on the way to or from their place of
work. Premises must be posted no later than 30 days after an
application is docketed and remain posted while the application is
pending before the Commission, during the term of the license, and for
30 days following license termination.
(2) Copies of NRC Form 3 may be obtained by writing to the Regional
Administrator of the appropriate NRC Regional Office listed in appendix
D to 10 CFR part 20, via email to [email protected], or by
visiting the NRC's online library at https://www.nrc.gov/reading-rm/doc-collections/forms/.
(f) No agreement affecting the compensation, terms, conditions, or
privileges of employment, including an agreement to settle a complaint
filed by an employee with the Department of Labor pursuant to section
211 of the Energy Reorganization Act of 1974, as amended, may contain
any provision which would prohibit, restrict, or otherwise discourage
an employee from participating in protected activity as defined in
paragraph (a)(1) of this section including, but not limited to,
providing information to the NRC or to his or her employer on potential
violations or other matters within NRC's regulatory responsibilities.
(g) Part 19 of 10 CFR sets forth requirements and regulatory
provisions applicable to licensees, holders of a standard design
approval, applicants for a license, standard design certification, or
standard design approval, and contractors or subcontractors of a
Commission licensee, or holder of a standard design approval, and are
in addition to the requirements in this section.
[[Page 87049]]
Sec. 53.070 Completeness and accuracy of information.
(a) Information provided to the Commission by a holder of a
license, permit, design certification, or standard design approval
under this part or an applicant for a license, permit, design
certification, or standard design approval under this part, and
information required by statute or by the Commission's regulations,
orders, license conditions, or terms and conditions of a standard
design approval to be maintained by the applicant or the licensee must
be complete and accurate in all material respects.
(b) Each applicant or licensee, each holder of a standard design
approval under this part, and each applicant for a standard design
certification under this part following Commission adoption of a final
design certification regulation, must notify the Commission of
information identified by the applicant or licensee as having for the
regulated activity a significant implication for public health and
safety or common defense and security. An applicant, licensee, or
holder violates this paragraph only if the applicant, licensee, or
holder fails to notify the Commission of information that the
applicant, licensee, or holder has identified as having a significant
implication for public health and safety or common defense and
security. Notification must be provided to the Administrator of the
appropriate Regional Office within 2 working days of identifying the
information. This requirement is not applicable to information which is
already required to be provided to the Commission by other reporting or
updating requirements.
Sec. 53.080 Specific exemptions.
(a) The Commission may, upon application by any interested person
or upon its own initiative, grant exemptions from the requirements of
the regulations of this part, which are authorized by law, will not
present an undue risk to the public health and safety, and are
consistent with the common defense and security.
(b) The Commission will not consider granting an exemption unless
special circumstances are present. Special circumstances are present
whenever--
(1) Application of the regulation in the particular circumstances
conflicts with other rules or requirements of the Commission;
(2) Application of the regulation in the particular circumstances
would not serve the underlying purpose of the rule or is not necessary
to achieve the underlying purpose of the rule;
(3) Compliance would result in undue hardship or other costs that
are significantly in excess of those contemplated when the regulation
was adopted, or that are significantly in excess of those incurred by
others similarly situated;
(4) The exemption would result in benefit to the public health and
safety that compensates for any decrease in safety that may result from
the grant of the exemption;
(5) The exemption would provide only temporary relief from the
applicable regulation and the licensee or applicant has made good faith
efforts to comply with the regulation; or
(6) There is present any other material circumstance not considered
when the regulation was adopted for which it would be in the public
interest to grant an exemption. If such condition is relied on
exclusively for demonstrating compliance with paragraph (b) of this
section, the exemption may not be granted until the Executive Director
for Operations has consulted with the Commission.
(c) Any person may request an exemption permitting the conduct of
construction activities prior to the issuance of a CP. The Commission
may grant such an exemption upon considering and balancing the
following factors:
(1) Whether conduct of the proposed activities will give rise to a
significant adverse impact on the environment and the nature and extent
of such impact, if any;
(2) Whether redress of any adverse environment impact from conduct
of the proposed activities can reasonably be effective should such
redress be necessary;
(3) Whether conduct of the proposed activities would foreclose
subsequent adoption of alternatives; and
(4) The effect of delay in conducting such activities on the public
interest, including whether the power needs to be used by the proposed
facility, the availability of alternative sources, if any, to meet
those needs on a timely basis and delay costs to the applicant and to
consumers.
(d) Issuance of such an exemption must not be deemed to constitute
a commitment to issue a CP. During the period of any exemption granted
pursuant to paragraph (c) of this section, any activities conducted
must be carried out in such a manner as will minimize or reduce their
environmental impact.
(e) The Commission's consideration of requests for exemptions from
requirements of the regulations of other parts in this chapter that are
applicable by virtue of this part must be governed by the exemption
requirements of those parts.
Sec. 53.090 Standards for review.
(a) Common standards. In determining that a CP, OL, early site
permit, COL, or ML under this part will be issued to an applicant, the
Commission will be guided by the following considerations:
(1) Except for an early site permit or ML, the processes to be
performed, the operating procedures, the facility and equipment, the
use of the facility, and other technical specifications, or the
proposals, in regard to any of the foregoing, collectively provide
reasonable assurance that the applicant will comply with the
regulations in this chapter, including the regulations in 10 CFR part
20, and that the health and safety of the public will not be
endangered.
(2) The applicant for a CP, OL, COL, or ML is technically and
financially qualified to engage in the proposed activities in
accordance with the regulations in this chapter. However, no
consideration of financial qualification is necessary for an electric
utility applicant for an OL for a utilization facility of the type
described in paragraph (d) of this section or for an applicant for an
ML.
(3) The issuance of a CP, OL, early site permit, COL, or ML to the
applicant will not, in the opinion of the Commission, be inimical to
the common defense and security or to the health and safety of the
public.
(4) Any applicable requirements of subpart A of 10 CFR part 51 have
been satisfied.
(b) Additional standards for licenses. In determining whether a
license will be issued to an applicant, the Commission will, in
addition to applying the standards set forth in paragraph (a) of this
section, consider whether the proposed activities will serve a useful
purpose proportionate to the quantities of SNM or source material to be
utilized.
(c) Additional standards and provisions affecting licenses for
commercial power. In addition to applying the standards set forth in
paragraphs (a) and (b) of this section, paragraphs (c)(1) through
(c)(4) of this section apply in the case of a license for a facility
for the generation of commercial power.
(1) The NRC will--
(i) Give notice in writing of each application to the regulatory
agency or State as may have jurisdiction over the rates and services
incident to the proposed activity;
(ii) Publish notice of the application in trade or news
publications as it
[[Page 87050]]
deems appropriate to give reasonable notice to municipalities, private
utilities, public bodies, and cooperatives which might have a potential
interest in the utilization or production facility; and
(iii) Publish notice of the application once each week for four
consecutive weeks in the Federal Register. No license will be issued by
the NRC prior to the giving of these notices and until four weeks after
the last notice is published in the Federal Register.
(2) If there are conflicting applications for a limited opportunity
for such license, the Commission will give preferred consideration in
the following order: first, to applications submitted by public or
cooperative bodies for facilities to be located in high cost power
areas in the United States; second, to applications submitted by others
for facilities to be located in such areas; third, to applications
submitted by public or cooperative bodies for facilities to be located
in areas other than high cost power areas; and, fourth, to all other
applicants.
(3) The licensee who transmits electric energy in interstate
commerce, or sells it at wholesale in interstate commerce, must be
subject to the regulatory provisions of the Federal Power Act.
(4) Nothing shall preclude any government agency, now or hereafter
authorized by law to engage in the production, marketing, or
distribution of electric energy, if otherwise qualified, from obtaining
a CP, OL, or COL under this part for a utilization facility for the
primary purpose of producing electric energy for disposition for
ultimate public consumption.
(d) Licenses for commercial nuclear plants. A license will be
issued, to an applicant who qualifies, for any one or more of the
following: to transfer or receive in interstate commerce, or
manufacture, produce, transfer, acquire, possess, or use a utilization
facility for industrial or commercial purposes.
Sec. 53.100 Jurisdictional limits.
No permit, license, standard design approval, or standard design
certification under this part shall be deemed to have been issued for
activities that are not under or within the jurisdiction of the United
States.
Sec. 53.110 Attacks and destructive acts.
Licensees, applicants for licenses, permits, certifications, and
design approvals, and applicants for an amendment to any license,
permit, certification, or design approval under this part are not
required to provide for design features or other measures for the
specific purpose of protection against the effects of--
(a) Attacks and destructive acts, including sabotage, directed
against the facility by an enemy of the United States, whether a
foreign government or other person; or
(b) Use or deployment of weapons incident to U.S. defense
activities.
Sec. 53.115 Rights related to special nuclear material.
(a) No right to the SNM will be conferred by a license issued under
this part except as may be defined by the license.
(b) Neither a license issued under this part, nor any right
thereunder, nor any right to utilize or produce SNM may be transferred,
assigned, or disposed of in any manner, either voluntarily or
involuntarily, directly or indirectly, through transfer of control of
the license to any person, unless the Commission, after securing full
information, finds that the transfer is in accordance with the
provisions of the Act and gives its consent in writing.
Sec. 53.117 License suspension and rights of recapture.
Any license issued under this part must be subject to suspension
and to the rights of recapture of the material or control of the
facility reserved to the Commission under section 108 of the Act in a
state of war or national emergency declared by Congress.
Sec. 53.120 Information collection requirements: OMB approval.
(a) The NRC has submitted the information collection requirements
contained in this part to the Office of Management and Budget (OMB) for
approval as required by the Paperwork Reduction Act (44 U.S.C. 3501 et
seq.). The NRC may not conduct or sponsor, and a person is not required
to respond to, a collection of information unless it displays a
currently valid OMB control number. OMB has approved the information
collection requirements contained in this part under control number
3150-XXXX.
(b) The approved information collection requirements contained in
this part appear in Sec. Sec. 53.070, 53.080, 53.240, 53.410, 53.420,
53.425, 53.430, 53.440, 53.450, 53.480, 53.500, 53.540, 53.605, 53.610,
53.620, 53.700, 53.710, 53.715, 53.720, 53.730, 53.780, 53.785, 53.805,
53.810, 53.815, 53.830, 53.850, 53.855, 53.865, 53.870, 53.875, 53.880,
53.910, 53.1010, 53.1020, 53.1030, 53.1045, 53.1060, 53.1070, 53.1075,
53.1080, 53.1100, 53.1109, 53.1115, 53.1130, 53.1140, 53.1144, 53.1146,
53.1173, 53. 1182, 53.1188, 53.1200, 53.1206, 53.1209, 53.1210,
53.1221, 53.1230, 53.1236, 53.1239, 53.1241, 53.1254, 53.1257, 53,1263,
53.1270, 53.1276, 53.1279, 53.1282, 53.1288, 53.1295, 53.1300, 53.1306,
53.1309, 53.1312, 53.1327, 53.1330, 53.1333, 53.1336, 53.1348, 53.1360,
53.1366, 53.1369, 53.1372, 53.1384, 53.1410, 53.1413, 53.1416, 53.1419,
53.1437, 53.1449, 53.1452, 53.1458, 53.1470, 53.1505, 53.1510, 53.1515,
53.1525, 53.1530, 53.1535, 53.1540, 53.1545, 53.1550, 53.1560, 53.1565,
53.1570, 53.1575, 53.1580, 53.1620, 53.1630, 53.1645, 53.1680, 53.1690,
53.1720.
(c) This part contains information collection requirements in
addition to those approved under the control number specified in
paragraph (a) of this section. The information collection requirement
and the control numbers under which it is approved are as follows:
(1) In Sec. Sec. 53.765, 53.770, 53.780, and 53.795, NRC Form 396
is approved under control number 3150-0024.
(2) In Sec. Sec. 53.775 and 53.795, NRC Form 398 is approved under
control number 3150-0090.
(3) In Sec. 53.1640, NRC Form 366 is approved under control number
3150-0104.
(4) In Sec. 53.1630, NRC Form 361 is approved under control number
3150-0238.
(5) In Sec. 53.1650, International Atomic Energy Agency Design
Information Questionnaire forms are approved under control number 3150-
0056.
(6) In Sec. 53.1650, DOC/NRC Form AP-A and associated forms are
approved under control numbers 0694-0135.
Subpart B--Technology-Inclusive Safety Requirements
Sec. 53.210 Safety criteria for design-basis accidents.
Design features and programmatic controls must be provided for each
commercial nuclear plant such that identification and analyses of
design-basis accidents (DBAs) in accordance with Sec. 53.240
demonstrate the following:
(a) An individual located at any point on the boundary of the
exclusion area for any 2-hour period following the onset of the
postulated fission product release would not receive a radiation dose
in excess of 25 rem (250 millisieverts) total effective dose equivalent
(TEDE); and
(b) An individual located at any point on the outer boundary of the
low-population zone who is exposed to the radioactive cloud resulting
from the postulated fission product release (during the entire period
of its passage) would not receive a radiation dose in
[[Page 87051]]
excess of 25 rem (250 millisieverts) TEDE.\1\
\1\ The use of 25 rem TEDE is not intended to imply that this
number constitutes an acceptable limit for an emergency dose to the
public under accident conditions. Rather, this dose value has been
set forth in this section as a reference value, which can be used in
the evaluation of plant design features with respect to postulated
reactor accidents, to assure that these designs provide assurance of
low risk of public exposure to radiation, in the event of an
accident.
Sec. 53.220 Safety criteria for licensing-basis events other than
design-basis accidents.
Design features and programmatic controls must be provided for each
commercial nuclear plant such that identification and analysis of
licensing-basis events (LBEs) other than DBAs in accordance with Sec.
53.240 demonstrate the following:
(a) Plant SSCs, personnel, and programs provide the necessary
capabilities and maintain the necessary reliability to address LBEs
other than DBAs in accordance with Sec. Sec. 53.240 and 53.450(e), and
provide measures for defense in depth in accordance with Sec. 53.250;
and
(b) The analysis of risks to public health and safety resulting
from LBEs other than DBAs under Sec. 53.450(e) includes comprehensive
risk metrics that satisfy associated risk performance objectives that
are acceptable to the NRC and provide an appropriate level of safety.
Sec. 53.230 Safety functions.
(a) The primary safety function is limiting the release of
radioactive materials from the facility and must be maintained during
normal operation and for LBEs over the life of the plant.
(b) Additional safety functions needed to support the retention of
radioactive materials during LBEs--such as controlling reactivity, heat
generation, heat removal, and chemical interactions--must be identified
for each commercial nuclear plant.
(c) The primary and additional safety functions are required to
satisfy the safety criteria defined in Sec. Sec. 53.210 and 53.220, or
more restrictive alternative criteria adopted under Sec. 53.470, and
must be fulfilled by the design features, human actions, and
programmatic controls specified throughout this part.
Sec. 53.240 Licensing-basis events.
(a) Licensing-basis events must be identified for each commercial
nuclear plant and analyzed under Sec. 53.450 to demonstrate that the
safety requirements in this subpart have been satisfied.
(b) The identified LBEs, ranging from anticipated event sequences
to very unlikely event sequences, must collectively address
combinations of malfunctions of plant SSCs, human errors, facility
hazards, and the effects of external hazards.
(c) The analysis of LBEs must--
(1) Include analysis of one or more DBAs under Sec. 53.450(f);
(2) Confirm the adequacy of design features and programmatic
controls needed to satisfy the safety criteria defined in Sec. Sec.
53.210 and 53.220, or more restrictive alternative criteria adopted
under Sec. 53.470, and
(3) Establish related functional requirements for plant SSCs,
personnel, and programs.
Sec. 53.250 Defense in depth.
(a) Measures must be taken for each commercial nuclear plant to
ensure appropriate defense in depth is provided to compensate for
uncertainties in the analysis of the safety criteria such that there is
reasonable assurance that the safety criteria in this subpart are met
over the life of the plant.
(b) The uncertainties that must be addressed under paragraph (a) of
this section include those related to the state of knowledge and
modeling capabilities, the ability of barriers to limit the release of
radioactive materials from the facility during LBEs other than DBAs,
the reliability and performance of plant SSCs and personnel, and the
effectiveness of programmatic controls.
(c) The safety analysis may not rely upon a single engineered
design feature, human action, or programmatic control, no matter how
robust, to address the range of LBEs other than DBAs.
Sec. 53.260 Normal operations.
Holders of licenses to operate commercial nuclear plants under this
part must control public doses and dose rates in unrestricted areas
from normal plant operations to meet the requirements in 10 CFR part
20.
Sec. 53.270 Protection of plant workers.
Holders of licenses to operate commercial nuclear plants under this
part must control occupational doses to meet the requirements in 10 CFR
part 20.
Subpart C--Design and Analysis Requirements
Sec. 53.400 Design features for licensing-basis events.
(a) Design features must be provided for each commercial nuclear
plant such that, when combined with corresponding human actions and
programmatic controls, the plant will satisfy the safety criteria
defined in Sec. Sec. 53.210 and 53.220, or more restrictive
alternative criteria adopted under Sec. 53.470.
(b) Design features must ensure that the safety functions
identified in Sec. 53.230 are fulfilled during licensing-basis events
(LBEs).
Sec. 53.410 Functional design criteria for design-basis accidents.
(a) Functional design criteria must be defined for each design
feature required by Sec. 53.400 and relied upon to demonstrate
compliance with the safety criteria defined in Sec. 53.210.
(b) Corresponding human actions and programmatic controls must be
identified and implemented in accordance with this and other subparts
to achieve and maintain the reliability and capability of structures,
systems, and components (SSCs) relied upon to satisfy the defined
functional design criteria and the safety criteria required in Sec.
53.210, and to maintain consistency with analyses required by Sec.
53.450(f).
Sec. 53.415 Protection against external hazards.
Safety-related (SR) SSCs must be protected against or must be
designed to withstand the effects of natural phenomena (e.g.,
earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches) and
constructed hazards (e.g., dams, transportation routes, military and
industrial facilities) considering an event severity up to the design-
basis external hazard levels as determined under Sec. 53.510 without
losing the capability to perform the safety functions identified under
Sec. 53.230. Specific requirements for earthquake engineering are
included in Sec. 53.480.
Sec. 53.420 Functional design criteria for licensing-basis events
other than design-basis accidents.
(a) Functional design criteria must be defined for each design
feature required by Sec. 53.400 and relied upon to--
(1) Demonstrate compliance with the safety criteria in Sec. 53.220
or more restrictive alternative criteria adopted under Sec. 53.470;
and
(2) Demonstrate compliance with the evaluation criteria in Sec.
53.450(e) or more restrictive alternative criteria adopted under Sec.
53.470.
(b) Corresponding human actions and programmatic controls must be
identified and implemented in accordance with this and other subparts
to achieve and maintain the reliability and capability of SSCs relied
upon to--
[[Page 87052]]
(1) Satisfy the safety criteria in Sec. 53.220 or more restrictive
alternative criteria adopted under Sec. 53.470; and
(2) Satisfy the evaluation criteria in Sec. 53.450(e) or more
restrictive alternate criteria adopted under Sec. 53.470.
Sec. 53.425 Design features and functional design criteria for normal
operations.
(a) Design features must be provided for each commercial nuclear
plant to support the Radiation Protection Program required in Sec.
53.850.
(b) Functional design criteria must be defined for each design
feature relied upon to demonstrate compliance with Sec. 53.850.
(c) Functional design criteria, including design objectives for
dose to the maximally exposed member of the public, must be defined for
design features to show that plant design features and corresponding
programmatic controls, including monitoring programs, control liquid,
gaseous, and solid wastes, as required under part 20 of this
chapter.\1\
\1\ A guide for keeping doses to the public as low as is
reasonably achievable is that the estimated annual dose to the
maximally exposed member of the public does not exceed 10 mrem total
effective dose equivalent. A design objective of maintaining doses
below 10 mrem/year should not be construed as a radiation protection
standard.
Sec. 53.430 Design features and functional design criteria for
protection of plant workers.
(a) Design features must be provided for each commercial nuclear
plant such that, when combined with corresponding programmatic
controls, the requirements in Sec. 53.270 can be met.
(b) Functional design criteria must be defined for each design
feature relied upon to demonstrate compliance with Sec. 53.270.
Sec. 53.440 Design requirements.
(a)(1) Analysis, appropriate test programs, prototype testing,
operating experience, or a combination thereof must demonstrate that
each design feature required by Sec. 53.400 meets the defined
functional design criteria required by Sec. Sec. 53.410 and 53.420.
This demonstration must consider interdependent effects throughout the
commercial nuclear plant and the range of conditions under which the
design features required by Sec. 53.400 must function throughout the
plant's lifetime.
(2) The design processes for SR and non-safety-related but safety-
significant (NSRSS) SSCs under this part must include administrative
procedures for evaluating operating, design, and construction
experience and for considering applicable important industry
experiences in the design of those SSCs.
(b) The design features required by Sec. 53.400 must, wherever
applicable, be designed using generally accepted consensus codes and
standards that have been endorsed or otherwise found acceptable by the
U.S. Nuclear Regulatory Commission (NRC).
(c) The materials used for each SR and NSRSS SSC must be qualified
for their service conditions over the design life of the SSC.
(d) Possible degradation mechanisms related to aging, fatigue,
chemical interactions, operating temperatures, effects of irradiation,
and other environmental factors that may affect the performance of SR
and NSRSS SSCs must be evaluated and used to inform the design and the
development of integrity assessment programs under Sec. 53.870.
(e)(1) Safety-related and NSRSS SSCs must be designed and located
to minimize, consistent with other safety requirements in this part,
the probability and effect of fires and explosions.
(2) Noncombustible and fire-resistant materials must be used
wherever practical throughout the facility, particularly in locations
with SR and NSRSS SSCs.
(3) Fire detection and fire suppression systems of appropriate
capacity and capability must be provided and designed to minimize the
adverse effects of fires on SR and NSRSS SSCs.
(4) Fire suppression systems must be designed to ensure that their
rupture or inadvertent operation does not significantly impair the
ability of SR and NSRSS SSCs to perform their safety functions to
satisfy Sec. 53.230.
(f) Safety and security must be considered together in the design
process such that, where possible, security issues are effectively
resolved through design and engineered security features.
(g) The reactor system and waste stores for each commercial nuclear
plant must be capable of achieving and maintaining a subcritical
condition during normal operations and following any LBE identified in
accordance with Sec. 53.240.
(h) Each commercial nuclear plant must have a capability to provide
long-term cooling of the reactor fuel and waste stores during normal
operations and following any LBE identified in accordance with Sec.
53.240.
(i) The design, analysis, staffing, and programmatic controls for
each commercial nuclear plant must consider the number of reactors,
waste stores, and other significant inventories of radioactive
materials and the associated operating configurations, common systems,
system interfaces, and system interactions.
(j)(1) Design features must be provided and related functional
design criteria defined such that, with limited use of operator
actions, one or more physical barriers are maintained to limit the
release of radionuclides from reactor systems, waste stores, or other
significant inventories of radioactive materials assuming the impact of
a large, commercial aircraft.
(2) The functional design criteria for those design features
provided to address the requirements in paragraph (j)(1) of this
section must be based on an assessment of the impact of a large,
commercial aircraft used for long distance flights in the United
States, with aviation fuel loading typically used in such flights, and
an impact speed and angle of impact considering the ability of both
experienced and inexperienced pilots to control large, commercial
aircraft at low altitude representative of a commercial nuclear plant's
low profile.\1\
\1\ Changes to the detailed parameters on aircraft impact
characteristics set forth in guidance must be approved by the
Commission.
(k) Design features and related functional design criteria must be
defined such that analyses demonstrate a low risk of permanent injury
to the public due to the health effects of the chemical hazards of
licensed material.
(l) Measures must be taken during the design of commercial nuclear
plants to minimize, to the extent practicable, contamination of the
facility and the environment, facilitate eventual decommissioning, and
minimize, to the extent practicable, the generation of radioactive
waste in accordance with Sec. 20.1406 of this chapter.
(m)(1) Each commercial nuclear plant must include criticality
monitoring capabilities meeting the requirements of either Sec. 70.24
of this chapter or paragraph (m)(2) of this section.
(2) In lieu of maintaining a monitoring system capable of detecting
criticality as described in Sec. 70.24 of this chapter, criticality
accident requirements may be satisfied by--
(i) Demonstrating the sub-criticality of special nuclear material,
except when it is inside the reactor and the reactor is being operated,
by maintaining k-effective below 0.95 at a 95 percent probability, 95
percent confidence level, under conditions that maximize reactivity for
the applicable storage and handling configurations, and
(ii) Providing radiation monitors for fuel storage and associated
handling
[[Page 87053]]
areas when fuel is present to detect excessive radiation levels and to
support initiating appropriate safety actions.
(3) While a spent fuel transportation package approved under 10 CFR
part 71 of this chapter or spent fuel storage cask approved under 10
CFR part 72 is in the special nuclear material handing or storage area,
the requirements in 10 CFR parts 71 or 72, as applicable, and the
requirements of the certificate of compliance for that package or cask,
are the applicable requirements for the fuel within that package or
cask.
(n)(1) The design of each commercial nuclear plant must reflect
state-of-the-art human factors principles for safe and reliable
performance in all locations that human activities are expected for
performing or supporting the continued availability of plant safety or
emergency response functions.
(2) The design must provide for the capabilities described in Sec.
53.730(b) to ensure the plant staff are able to monitor plant
conditions and respond to events.
(3) The means by which the design and human actions together will
achieve the safety requirements of subpart B of this part must be
evaluated and used to inform the design and the development of the
concept of operations required by Sec. 53.730(c).
(4) A functional requirements analysis and function allocation must
be used to ensure that plant design features address how safety
functions and functional safety criteria are satisfied, and how the
safety functions will be assigned to appropriate combinations of human
action, automation, active safety features, passive safety features, or
inherent safety characteristics.
Sec. 53.450 Analysis requirements.
(a) Requirement to have a probabilistic risk assessment (PRA). A
PRA of each commercial nuclear plant must be performed to identify
potential failures, susceptibility to internal and external hazards,
and other contributing factors to event sequences that might challenge
the safety functions identified in Sec. 53.230 and to support
demonstrating that each commercial nuclear plant meets the safety
criteria of Sec. 53.220, or more restrictive alternative criteria
adopted under Sec. 53.470.
(b) Specific uses of analyses. The PRA in combination with other
generally accepted approaches for systematically evaluating engineered
systems must be used--
(1) In informing the selection of the LBEs, as described in Sec.
53.240, which must be considered in the design to determine compliance
with the safety criteria in subpart B of this part.
(2) For informing the classification of SSCs according to their
safety significance in accordance with Sec. 53.460 and for identifying
the environmental conditions under which the SSCs and operating staff
must perform their safety functions.
(3) In evaluating the adequacy of defense-in-depth measures
required in accordance with Sec. 53.250.
(4) To identify and assess all plant operating states where there
is the potential for the uncontrolled release of radioactive material
to the environment.
(5) To identify and assess events that challenge plant control and
safety systems whose failure could lead to the uncontrolled release of
radioactive material to the environment. These include internal events,
such as human errors and equipment failures, and external events
identified in accordance with subpart D of this part.
(c) Maintenance and upgrade of analyses. The PRA must be maintained
at least every 5 years until the permanent cessation of operations
under Sec. 53.1070 and upgraded in conformance with generally accepted
methods, standards, and practices that have been endorsed or otherwise
found acceptable by the NRC.
(d) Qualification of analytical codes. The analytical codes used in
modeling plant behavior in analyses of licensing-basis events
(including but not limited to thermodynamics, reactor physics, fuel
performance, and mechanistic source term codes) must be qualified for
the range of conditions for which they are to be used.
(e) Analyses of licensing-basis events other than design-basis
accidents.
(1) Analyses must be performed for LBEs other than design-basis
accidents (DBAs). These LBEs must be identified using insights from a
PRA in combination with other generally accepted approaches for
systematically evaluating engineered systems to identify and analyze
equipment failures and human errors.
(2) The analysis of LBEs other than DBAs must include definition of
evaluation criteria for each event or specific categories of LBEs to
determine the acceptability of the plant response to the challenges
posed by internal and external hazards to provide an appropriate level
of safety.
(3) The analyses of LBEs other than DBAs must address event
sequences from initiation to a defined end state and be used in
combination with other engineering analyses to demonstrate that the
functional design criteria required by Sec. 53.420 provide sufficient
barriers to the unplanned release of radionuclides to satisfy the
evaluation criteria defined for each LBE other than DBAs, to satisfy
the safety criteria specified in accordance with Sec. 53.220 and
provide defense in depth as required by Sec. 53.250.
(4) The methodology used to identify, categorize, and analyze LBEs
must include a means to identify event sequences deemed significant for
controlling the risks posed to public health and safety.
(f) Analysis of design-basis accidents. (1) The analysis of LBEs
required by Sec. 53.240 must include analysis of DBAs that address
possible challenges to the safety functions identified under Sec.
53.230. The events selected as DBAs must be those that, if not
terminated, have the potential for exceeding the safety criteria in
Sec. 53.210.
(2) The DBAs selected must be analyzed using deterministic methods
that address event sequences from initiation to a safe stable end state
and assume only the SR SSCs identified under Sec. 53.460 and human
actions addressed by the requirements of subpart F of this part are
available to perform the safety functions identified in accordance with
Sec. 53.230.
(3) The analysis must conservatively demonstrate compliance with
the safety criteria in Sec. 53.210.
(g) Other required analyses. Analyses must be performed to assess--
(1) Fire protection. Fire protection measures to demonstrate,
through inclusion of fires in the analysis of LBEs or by separate
analyses, that a fire or explosion in any plant area would not--
(i) Prevent equipment from fulfilling the safety functions
identified in accordance with Sec. 53.230, or
(ii) Challenge the safety criteria in Sec. Sec. 53.210 and 53.220.
(2) Aircraft impact. Measures provided to protect against aircraft
impacts under Sec. 53.440(j).
(3) Dose to members of the public. Measures taken under Sec.
53.425, including estimating--
(i) The quantity of each of the principal radionuclides expected to
be released annually to unrestricted areas in liquid effluents produced
during normal reactor operations and the dose to the maximally exposed
member of the public in unrestricted areas.
(ii) The quantities of each of the principal radionuclides of the
gases, halides, and particulates expected to be released annually to
unrestricted areas in gaseous effluents produced during normal reactor
operations and the dose to the maximally exposed member of the public
in unrestricted areas.
(iii) The annual external radiation dose in unrestricted areas and
the maximally exposed member of the
[[Page 87054]]
public in unrestricted areas due to direct radiation from contained
radiation sources from the commercial nuclear plant during normal
reactor operations.
Sec. 53.460 Safety categorization and special treatments.
(a) Structures, systems, and components must be classified
according to their safety significance. The SSC categories must include
``Safety-Related,'' ``Non-Safety-Related but Safety-Significant,'' and
``Non-Safety-Significant,'' as defined in subpart A of this part.
(b) For SR and NSRSS SSCs, the conditions under which they must
perform their safety function in Sec. 53.230 must be identified.
Special treatments must be established in accordance with this and
other subparts to provide confidence that the SSCs will perform under
the service conditions and with reliability consistent with the
analysis performed under Sec. 53.450 to demonstrate meeting the safety
criteria in Sec. Sec. 53.210 and 53.220, or more restrictive
alternative criteria adopted under Sec. 53.470.
(1) The special treatments for SR SSCs must include meeting the
applicable quality assurance requirements from appendix B of part 50 of
this chapter.
(2) The special treatments for NSRSS SSCs and special treatments
for SR SSCs beyond those required under (b)(1) of this section may
include meeting selected quality assurance requirements from appendix B
of part 50 of this chapter when such treatment is needed to address
performance requirements, equipment reliability, or uncertainties.
(c) Human actions needed to prevent or mitigate LBEs must be
identified, be able to be performed reliably under the postulated
environmental conditions, and be addressed by programs established in
accordance with subpart F of this part to provide confidence that those
actions will be performed as assumed in the analysis performed in
accordance with Sec. 53.450 to demonstrate meeting the criteria in
Sec. Sec. 53.210, 53.220, and 53.450(e), or more restrictive
alternative criteria adopted under Sec. 53.470.
Sec. 53.470 Maintaining analytical safety margins used to justify
operational flexibilities.
Where an applicant or licensee so chooses, alternative criteria
more restrictive than those defined in Sec. Sec. 53.220 and 53.450(e)
may be adopted to support operational flexibilities. In such cases,
applicants and licensees must ensure that the functional design
criteria of Sec. 53.420, the analysis requirements of Sec. 53.450(e),
and identification of special treatment of SSCs and human actions under
Sec. 53.460 reflect and support the use of alternative criteria to
justify operational flexibilities. Licensees must ensure that measures
taken to provide the analytical margins supporting operational
flexibilities are incorporated into design features and programmatic
controls and are maintained within programs required in other subparts.
Sec. 53.480 Earthquake engineering.
(a) Effects of earthquakes. Structures, systems, and components
classified as SR or NSRSS must be able to withstand the effects of
earthquakes, commensurate with the safety significance of the SSC,
without loss of capability to perform their role in fulfilling the
safety functions required by Sec. 53.230.
(b) Definitions. For the purpose of this section--
Design-Basis Ground Motions (DBGMs) are the vibratory ground
motions for which certain SSCs must be designed to remain functional.
Operating basis earthquake (OBE) ground motion is the vibratory
ground motion for which those features of the commercial nuclear plant
necessary for continued operation without undue risk to the health and
safety of the public are designed to remain functional. The OBE ground
motion is used in Sec. 53.720.
Response spectrum is a plot of the maximum responses (acceleration,
velocity, or displacement) of idealized single-degree-of-freedom
oscillators as a function of the natural frequencies of the oscillators
for a given damping value. The response spectrum is calculated for a
specified vibratory motion input at the oscillators' supports.
Surface deformation is the distortion of geologic strata on or near
the ground surface that occurs because of tectonic forces that result
from earthquakes.
(c) Design considerations--(1) Design-Basis Ground Motions. (i) The
DBGMs must be derived from the Site Ground Motion Response Spectra
developed in accordance with Sec. 53.510(c), by taking into
consideration the functional design criteria of SSCs in accordance with
Sec. Sec. 53.410 and 53.420. The horizontal component of the DBGM(s)
in the free-field at the foundation level of the structures must be an
appropriate response spectrum that is determined based on the risk-
significance of SSCs and their safety functions. In view of the limited
data available on vibratory ground motion of strong earthquakes, it is
acceptable that the design response spectra be smoothed spectra.
(ii) The commercial nuclear plant must be designed so that, if the
DBGMs occur, the following SSCs remain functional and within applicable
stress, strain, and deformation limits:
(A) Structures, systems, and components for which functional design
criteria are established in accordance with Sec. 53.410 or Sec.
53.420; and
(B) Structures, systems, and components classified as SR or NSRSS
commensurate with safety significance in accordance with Sec. 53.460.
(iii) In addition to seismic loads, applicable concurrent normal
operating, functional, and accident-induced loads must be taken into
account in the design of the SR SSCs and, commensurate with safety
significance, NSRSS SSCs.
(iv) The design of the commercial nuclear plant must take into
account the possible effects of seismic-induced ground disruption, such
as fissuring, lateral spreads, differential settlement, liquefaction,
and landsliding, on the facility foundations.
(v) The SSCs fulfilling the safety functions required by Sec.
53.230 must be demonstrated through design, testing, or qualification
methods to be able to fulfill those safety functions during and after
the vibratory ground motion associated with the DBGMs.
(vi) The evaluation of SSCs required by this section to show they
are able to function during and after earthquake ground motion must
take into account soil-structure interaction effects and the expected
duration of vibratory motion. It is permissible to design for strain
limits in excess of yield strain in some of these SSCs during the DBGMs
and under the postulated concurrent loads, provided the necessary
safety functions are maintained.
(2) OBE Ground Motion. The OBE Ground Motion must be characterized
by response spectra. The value of the OBE Ground Motion must be set to
one-third or less of the DBGMs response spectra.
(3) [Reserved]
(4) Required seismic instrumentation. Suitable instrumentation must
be provided so that the seismic response of commercial nuclear plant SR
SSCs or NSRSS SSCs can be evaluated promptly after an earthquake.
(d) Surface deformation. (1) The potential for surface deformation
must be taken into account in the design of the commercial nuclear
plant by providing reasonable assurance that in the event of
deformation, SSCs classified as SR or NSRSS in accordance with Sec.
53.460 will remain functional.
(2) In addition to surface deformation induced loads, the design of
SSCs must take into account, commensurate with
[[Page 87055]]
safety significance, seismic loads and applicable concurrent functional
and accident-induced loads.
(3) The design provisions for surface deformation must be based on
its postulated occurrence in any direction and azimuth and under any
part of the commercial nuclear plant, unless evidence indicates this
assumption is not appropriate, and must take into account the estimated
rate at which the surface deformation may occur.
(e) Seismically induced floods and water waves and other design
conditions. Seismically induced floods and water waves from either
locally or distantly generated seismic activity and other design
conditions determined pursuant to subpart D of this part must be taken
into account in the design of the commercial nuclear plant so as to
prevent undue risk to the health and safety of the public.
(f) Analysis. The analyses required by Sec. 53.450 must address
seismic hazards and related SSC responses in determining that the
safety criteria defined in Sec. 53.220 will be met.
(g) Design criteria, human actions, and programmatic controls.
Functional design criteria, human actions, and programmatic controls
needed to address seismic events must be identified and implemented in
accordance with this and other subparts to achieve and maintain the
performance of SSCs relied upon to satisfy the safety criteria in Sec.
53.220 and to maintain consistency with analyses required by Sec.
53.450 when accounting for the site-specific frequencies and magnitudes
of earthquakes for a commercial nuclear plant.
Subpart D--Siting Requirements
Sec. 53.500 General siting and siting assessment.
(a) The siting of each commercial nuclear plant must be supported
by assessments of proposed sites such that the design, including design
features and programmatic controls corresponding to the site
characteristics, satisfies the safety criteria defined in Sec. Sec.
53.210 and 53.220 or more restrictive alternative criteria adopted
under Sec. 53.470. The siting assessment must ensure that site
characteristics that might contribute to the initiation, progression,
or consequences of licensing-basis events (LBEs) analyzed under
Sec. Sec. 53.450 and 53.480 are identified and mitigated by design
features or programmatic controls. The siting assessment must take into
consideration the potential adverse impacts that a commercial nuclear
plant may have on nearby populations as a result of normal operations
or LBEs.
(b) Activities performed to identify site characteristics or
otherwise needed to determine site-specific contributors to functional
design criteria or analysis assumptions under subpart C of this part
must satisfy the applicable special treatment requirements of Sec.
53.460, including, where applicable, the quality assurance requirements
from appendix B of part 50 of this chapter.
Sec. 53.510 External hazards.
(a) General external hazard requirements. The design-basis external
hazard level for the relevant external hazards for a site must be
identified and characterized based on site-specific assessments of
natural and constructed hazards with the potential to adversely affect
plant functions. The external hazard frequencies and magnitudes
determined from the site-specific assessments must take into account
uncertainties and variabilities in data, models, and methods relied on
to characterize the external hazards.
(b) Definitions. For the purpose of this section, the following
terms mean:
Geological siting factors are geological and seismic factors that
may affect the design and operation of the proposed commercial nuclear
plant.
Ground Motion Response Spectra (GMRS) are the site-specific GMRS
resulting from the geologic investigations and evaluations of the site
vicinity and region and used to determine design-basis ground motions
for structures, systems, and components under Sec. 53.480.
Probabilistic seismic hazard analysis is an analytical methodology
that incorporates uncertainty into estimates of an annual frequency of
exceedance for a certain ground motion parameter (e.g., peak ground
acceleration, peak ground velocity, response spectral values) at a
site.
(c) Geological investigations. The GMRS for the site must be
determined based on the results of investigations of the geological,
seismological, and engineering characteristics of the site and its
environs and must be characterized by both horizontal and vertical
free-field GMRS at the free ground surface. The size of the region to
be investigated and the type of data pertinent to the investigations
must be determined based on the nature of the region surrounding the
site. Data on vibratory ground motion, earthquake recurrence rates,
fault geometry and slip rates, and site subsurface material properties
must be obtained by reviewing pertinent literature and carrying out
field investigations. Uncertainties are inherent in the parameters and
models used to estimate the GMRS for the site. The site assessment must
reflect these uncertainties through an appropriate analysis, such as a
probabilistic seismic hazard analysis.
(d) Geologic and seismic siting factors. The geologic and seismic
siting factors considered for design under Sec. Sec. 53.415 and 53.480
must include, but are not limited to, determination of the potential
for surface tectonic and nontectonic deformations, the size and
character of seismically induced floods and water waves that could
affect a site from either locally or distantly generated seismic
activity, soil and rock stability, liquefaction potential, and natural
and artificial slope stability.
Sec. 53.520 Site characteristics.
Site characteristics that might contribute to the initiation,
progression, or consequences of LBEs analyzed under Sec. 53.450 must
be identified, assessed, and considered in the design and analyses
required by subpart C of this part.
Sec. 53.530 Population-related considerations.
Every site must have an exclusion area, a low-population zone, and
a population center distance as defined in Sec. 53.020.
(a) The offsite radiological consequences estimated by the analyses
required by Sec. 53.450(f) must be used to confirm that--
(1) An individual located at any point on the boundary of the
exclusion area for any 2-hour period following onset of the postulated
fission product release would not receive a radiation dose in excess of
25 rem (250 millisieverts) total effective dose equivalent.
(2) An individual located at any point on the outer boundary of the
low-population zone who is exposed to the radioactive cloud resulting
from the postulated fission product release (during the entire period
of its passage) would not receive a radiation dose in excess of 25 rem
(250 millisieverts) total effective dose equivalent.
(b) The population center distance must be at least one and one-
third times the distance from the reactor to the outer boundary of the
low-population zone. The boundary of the population center must be
determined upon consideration of population distribution. Political
boundaries are not controlling in the calculation of population center
distance.
(c) Reactor sites should be located away from very densely
populated centers. Areas of low-population density
[[Page 87056]]
are, generally, preferred. However, in determining the acceptability of
a particular site located away from a very densely populated center but
not in an area of low-population density, consideration will be given
to safety, environmental, economic, or other factors, which may result
in the site being found acceptable.
Sec. 53.540 Siting interfaces.
Site characteristics must be addressed by the design features,
programmatic controls, and supporting analyses used to demonstrate that
the safety criteria in Sec. Sec. 53.210 and 53.220 are met for each
commercial nuclear plant. Site characteristics must be such that
adequate emergency plans and security plans can be developed and
maintained.
Subpart E--Construction and Manufacturing Requirements
Sec. 53.600 Construction and manufacturing--scope and purpose.
This subpart applies to those construction and manufacturing
activities authorized by a construction permit (CP), combined license
(COL), manufacturing license (ML), or limited work authorization (LWA)
issued under this part.
Sec. 53.605 Reporting of defects and noncompliance.
Each CP and ML issued under this part is subject to the terms and
conditions in this section, and each COL issued under this part is
subject to the terms and conditions in this section until the date that
the Commission makes the finding under Sec. 53.1452(g).
(a) Definitions. The definitions in Sec. 21.3 of this chapter
apply to this section.
(b) Posting requirements. (1) Each individual, partnership,
corporation, dedicating entity, or other entity subject to the
regulations in this section must post current copies of this section
and the regulations in 10 CFR part 21; section 206 of the Energy
Reorganization Act of 1974, as amended; and procedures adopted under
these regulations. These documents must be posted in a conspicuous
position on any premises within the United States where the activities
subject to the license are conducted.
(2) If posting of these regulations or the procedures adopted under
them is not practical, the licensee may, in addition to posting section
206 of the Energy Reorganization Act of 1974, as amended, post a notice
that describes the regulations/procedures, including the name of the
individual to whom reports may be made, and states where they may be
examined.
(c) Procedures. The holder of a CP, COL, or ML subject to this
section must adopt appropriate procedures to--
(1) Evaluate deviations and failures to comply to identify defects
and failures to comply associated with substantial safety hazards as
soon as practicable, and, except as provided in paragraph (c)(2) of
this section, in all cases within 60 days of discovery, to identify a
reportable defect or failure to comply that could create a substantial
safety hazard, were it to remain uncorrected.
(2) Ensure that if an evaluation of an identified deviation or
failure to comply potentially associated with a substantial safety
hazard cannot be completed within 60 days from the discovery of the
deviation or failure to comply, an interim report is prepared and
submitted to the Commission through a director or responsible officer,
or designated person as discussed in paragraph (d)(5) of this section.
The interim report should describe the deviation or failure to comply
that is being evaluated and should also state when the evaluation will
be completed. This interim report must be submitted in writing within
60 days of discovery of the deviation or failure to comply.
(3) Ensure that a director or responsible officer of the holder of
a CP, COL, or ML subject to this section is informed as soon as
practicable, and, in all cases, within the 5 working days after
completion of the evaluation described in paragraph (c)(1) or (c)(2) of
this section, if the construction or manufacture of a facility or
activity, or a basic component supplied for such a facility or
activity--
(i) Fails to comply with the Atomic Energy Act of 1954, as amended,
or any applicable regulation, order, or license of the Commission
relating to a substantial safety hazard;
(ii) Contains a defect; or
(iii) Underwent any significant breakdown in any portion of the
quality assurance program (QAP) conducted under the requirements of
appendix B to part 50 of this chapter that could have produced a defect
in a basic component. These breakdowns in the QAP are reportable
whether or not the breakdown actually resulted in a defect in a design
approved and released for construction, installation, or manufacture.
(d) Reporting defects and noncompliance. (1) The holder of a CP,
COL, or ML subject to this section that obtains information reasonably
indicating that the facility or manufactured reactors fails to comply
with the Atomic Energy Act of 1954, as amended, or any applicable
regulation, order, or license of the Commission relating to a
substantial safety hazard must notify the Commission of the failure to
comply through a director, responsible officer, or designated person as
discussed in paragraph (d)(5) of this section.
(2) The holder of a CP, COL, or ML subject to this section that
obtains information reasonably indicating the existence of any defect
found in the construction or manufacture, or any defect found in the
final design of a facility as approved and released for construction or
manufacture, must notify the Commission of the defect through a
director, responsible officer, or designated person as discussed in
paragraph (d)(5) of this section.
(3) The holder of a CP, COL, or ML subject to this part, who
obtains information reasonably indicating that the QAP has undergone
any significant breakdown discussed in paragraph (c)(3)(iii) of this
section must notify the Commission of the breakdown in the QAP through
a director, responsible officer, or designated person as discussed in
paragraph (d)(5) of this section.
(4) When acting as a dedicating entity, the holder of a CP, COL, or
ML subject to this section is responsible for identifying and
evaluating deviations; reporting defects and failures to comply
associated with substantial safety hazards for dedicated items; and
maintaining auditable records for the dedication process.
(5) The notification requirements of this paragraph apply to all
defects and failures to comply associated with a substantial safety
hazard regardless of whether extensive evaluation, redesign, or repair
is required to conform to the criteria and bases stated in the Safety
Analysis Report, CP, COL, or ML. Evaluation of potential defects and
failures to comply and reporting of defects and failures to comply
under this section satisfies the CP holder's, COL holder's, and ML
holder's evaluation and notification obligations under 10 CFR part 21,
and satisfies the responsibility of individual directors or responsible
officers or holders of a CP, COL, or ML subject to this section to
report defects, and failures to comply associated with substantial
safety hazards under section 206 of the Energy Reorganization Act of
1974, as amended. The director or responsible officer may authorize an
individual to provide the notification required by this section.
However, this does not relieve the director or responsible officer of
his or her responsibility under this section.
[[Page 87057]]
(e) Notification--timing and where sent. The notification required
by paragraph (d) of this section must consist of--
(1) Initial notification by telephone, facsimile, or email
identified in appendix A to 10 CFR part 73 to the U.S. Nuclear
Regulatory Commission (NRC) Operations Center within 2 days following
receipt of information by the director or responsible corporate officer
under paragraph (c)(3) of this section, on the identification of a
defect or a failure to comply. If the CP, COL, or ML holder elects to
use facsimile, verification that the facsimile has been received should
be made by calling the NRC Operations Center. This paragraph does not
apply to interim reports described in paragraph (c)(2) of this section.
(2) Written notification submitted to the NRC Document Control Desk
by an appropriate method listed in Sec. 53.040, with a copy to the
appropriate NRC Regional Administrator at the address specified in
appendix D to 10 CFR part 20 and a copy to the appropriate NRC resident
inspector, if applicable, within 30 days following receipt of
information by the director or responsible corporate officer under
paragraph (c)(3) of this section, on the identification of a defect or
failure to comply.
(f) Content of notification. The written notification required by
paragraph (e)(2) of this section must clearly indicate that the written
notification is being submitted under this section and include the
following information, to the extent known.
(1) Name and address of the individual or individuals informing the
Commission.
(2) Identification of the facility, the activity, or the basic
component supplied for the facility or the activity within the United
States which contains a defect or fails to comply.
(3) Identification of the firm constructing or manufacturing the
facility or supplying the basic component which fails to comply or
contains a defect.
(4) Nature of the defect or failure to comply and the safety hazard
which is created or could be created by the defect or failure to
comply.
(5) The date on which the information of a defect or failure to
comply was obtained.
(6) In the case of a basic component that contains a defect or
failure to comply, the number and location of these components in use
at the facility subject to the regulations in this part.
(7) In the case of a completed reactor manufactured under this
part, the entities to which the reactor was supplied.
(8) The corrective action which has been, is being, or will be
taken; the name of the individual or organization responsible for the
action; and the length of time that has been or will be taken to
complete the action.
(9) Any advice related to the defect or failure to comply about the
facility, activity, or basic component that has been, is being, or will
be given to other entities.
(g) Procurement documents. Each holder of a CP, COL, or ML subject
to this section must ensure that each procurement document for a
facility or a basic component specifies the provisions of 10 CFR part
21 or this section that apply, as applicable.
(h) Coordination with 10 CFR part 21. The requirements of this
section are satisfied when the defect or failure to comply associated
with a substantial safety hazard has been previously reported under 10
CFR part 21, under Sec. 73.1205 of this chapter, under this section,
or under Sec. 53.1640.
(i) Records retention. The holder of a CP, COL, or ML subject to
this section must prepare and maintain records necessary to accomplish
the purposes of this section, specifically--
(1) Retain procurement documents, which define the requirements
that facilities or basic components must satisfy in order to be
considered acceptable, for the lifetime of the facility or basic
component.
(2) Retain records of evaluations of all deviations and failures to
comply under paragraph (c)(1) of this section for the longest of--
(i) Ten years from the date of the evaluation;
(ii) Five years from the date that an early site permit is
referenced in an application for a COL; or
(iii) Five years from the date of delivery of a manufactured
reactor.
(3) Retain records of all interim reports to the Commission made
under paragraph (c)(2) of this section, or notifications to the
Commission made under paragraph (d) of this section for the minimum
time periods stated in paragraph (i)(2) of this section;
(4) Suppliers of basic components must retain records of--
(i) All notifications sent to affected licensees or purchasers
under paragraph (d)(4) of this section for a minimum of 10 years
following the date of the notification;
(ii) The facilities or other purchasers to whom the basic
components or associated services were supplied for a minimum of 15
years from the delivery of the basic component or associated services.
(5) Maintaining reports in accordance with this section satisfies
the recordkeeping obligations under 10 CFR part 21 of the entities,
including directors or responsible officers thereof, subject to this
section.
Sec. 53.610 Construction.
(a) Management and control. Licensees must ensure that the
following plans, programs, and organizational units are developed and
implemented to manage and control the construction activities:
(1) Programs to ensure that the construction of a commercial
nuclear plant supports the eventual compliance with the design and
analysis requirements in subpart C of this part.
(2) An organization, headed by qualified personnel, responsible for
managing, controlling, and evaluating the adequacy of the construction
activities.
(3) Procedures describing the qualifications for personnel in key
positions in the licensee's management and control organization and the
organizational responsibilities, authority, and interfaces with other
parts of the licensee's organization.
(4) Procedures to evaluate the applicability of other national and
international construction experience to the planned and ongoing
construction activities and to ensure the applicable experience will be
provided to those constructing the plant.
(5) A fitness-for-duty program, under 10 CFR part 26.
(6)(i) A QAP meeting the requirements of appendix B of part 50 of
this chapter as required by Sec. 53.460(b).
(ii) Appropriate programmatic controls to provide special treatment
for non-safety-related but safety-significant structures, systems, and
components (SSCs).
(7) A radiation protection program, in accordance with 10 CFR part
20, that includes measures for monitoring the dose to individuals
working with radioactive materials brought onto the site, as
applicable.
(8) An information security program in accordance with Sec. Sec.
73.21, 73.22, and 73.23 of this chapter, as applicable.
(b) Construction activities. No person may begin the construction
of a commercial nuclear plant on a site on which the facility is to be
operated under this part until that person has been issued either a CP
or COL, an early site permit authorizing activities under Sec.
53.1130, or an LWA under this part.
(1) Licensees must satisfy the following requirements:
[[Page 87058]]
(i) As appropriate, considering the types and quantities of
radioactive materials being brought onto the site--
(A) The licensee must maintain and follow a special nuclear
material (SNM) material control and accounting program, a measurement
control program, and other material control procedures that include
corresponding record management requirements as required by the
provisions of Sec. 70.32 of this chapter. Prior to initial receipt of
SNM onsite, the licensee must implement an SNM material control and
accounting program in accordance with 10 CFR part 74.
(B) Procedures must be in place to receive, possess, use, and store
source, byproduct, and SNM in accordance with applicable portions of 10
CFR parts 30, 40, and 70.
(C) A plant staff training program associated with the receipt of
radioactive material must be approved and implemented prior to initial
receipt of byproduct, source or SNM (excluding exempt quantities as
described in Sec. 30.18 of this chapter).
(ii) For construction of a commercial nuclear plant involving
multiple reactor units, plans and procedures must be in place to
prevent or mitigate potential hazards to the SSCs of operating units
resulting from construction activities, including the managerial and
administrative controls to be used to provide assurance that the
limiting conditions for operation of the operating units are not
exceeded as a result of construction activities.
(iii) Procedures must be in place prior to the start of
construction activities that describe how construction will be
controlled so as not to impact other features important to the design,
such as dewatering, slope stability, backfill, compaction, and seepage.
(iv) For LWA holders, a plan must be developed for redress of
activities performed under the LWA should one of the following
situations arise:
(A) LWA work activities are terminated by the holder of the LWA;
(B) The LWA is revoked by the NRC; or
(C) The Commission denies the associated CP or COL application.
(2)(i) Onsite fresh fuel must be protected and stored in compliance
with Sec. 73.67 of this chapter.
(ii) Before initial fuel load into the reactor (or, for a fueled
manufactured reactor, before initiating the physical removal of any one
of the independent physical mechanisms to prevent criticality required
under Sec. 53.620(d)(1)), a cybersecurity program that meets the
requirements of Sec. Sec. 73.54 or 73.110 of this chapter, a physical
security program that meets the requirements of Sec. Sec. 73.55 or
73.100 of this chapter, and an access authorization program that meets
the requirements of Sec. Sec. 73.56 or 73.120 of this chapter must be
established, as applicable.
(iii) Fire protection measures must be implemented for work and
storage areas (including adjacent fire areas that could affect the work
or storage area) before initial receipt of byproduct, source, or non-
fuel SNM (excluding exempt quantities as described in Sec. 30.18 of
this chapter). The fire protection measures for areas associated with
new fuel (including all fuel handling, fuel storage, and adjacent fire
areas that could affect the new fuel) must be implemented before
receipt of fuel. Prior to the receipt of fuel, a formal letter of
agreement must be in place with the local fire department specifying
the nature of arrangements in support of the fire protection program.
(c) Inspection and acceptance. (1) The licensee must have a process
for accepting individual or groups of SSCs upon completion of
construction and protecting them from damage or tampering as other
construction activities continue.
(2) The post construction acceptance process must address the
inspections, tests, analyses, and acceptance criteria specified in the
COL under Sec. 53.1440 or the equivalent verifications needed to
support the issuance of an operating license under Sec. 53.1387.
Sec. 53.620 Manufacturing.
(a) Management and control. Holders of MLs must ensure that the
following plans, programs, and organizational units are developed and
implemented to manage and control the manufacturing activities within
the scope of the ML:
(1) Programs to ensure that the manufacturing of a manufactured
reactor or portions of a manufactured reactor complies with the design
and analysis requirements in subpart C of this part. The entity with
design authority for the manufactured reactor covered by the ML must be
identified in the license.
(2) An organizational and management structure responsible for
managing, controlling, and evaluating the adequacy of the reactor
design and manufacturing activities.
(3) Procedures describing the qualifications for personnel in key
positions in the licensee's management and control organization and the
organizational responsibilities, authority, and interfaces with other
parts of the licensee's organization.
(4) A program to evaluate the applicability of other national and
international design and manufacturing experience to the planned and
ongoing manufacturing activities.
(5) A fitness-for-duty program, in accordance with 10 CFR part 26.
(6)(i) A QAP meeting the requirements of appendix B to part 50 of
this chapter, to be applied to the design, fabrication, construction,
and testing of the SSCs of the manufactured reactor.
(ii) Appropriate programmatic controls to provide special treatment
measures for non-safety-related but safety-significant SSCs.
(7) A radiation protection program, in accordance with 10 CFR part
20, that includes measures for monitoring the dose to individuals if
the manufacturing activities include working with radioactive
materials.
(8) An information security program in accordance with Sec. Sec.
73.21, 73.22 and 73.23 of this chapter, as applicable.
(b) Manufacturing activities. Holders of MLs must satisfy the
following requirements:
(1) The manufacturing process must be conducted within facilities
for which the ML holder has the authority to establish controls on any
activity that might affect manufacturing. The licensee must establish
access controls to the portions of each facility involved in the
manufacturing processes governed by the ML.
(2) Manufacturing processes must be performed in accordance with
the ML and the referenced codes and standards that have been endorsed
or otherwise found acceptable by the NRC.
(3) A post-manufacturing inspection and acceptance process must be
established and implemented before transporting a manufactured reactor
or portions of a manufactured reactor for installation at a commercial
nuclear plant. The process must consider the results of inspections,
tests, and analyses that have been performed and the acceptance
criteria that are necessary and sufficient to conclude that
manufacturing activities have been completed in accordance with the ML.
(c) Control of radioactive materials. As appropriate considering
the types and quantities of radioactive materials being brought into
the manufacturing facility--
(1) Procedures must be in place to receive, transfer, possess, and
use source, byproduct, and SNM in accordance with the applicable
portions of 10 CFR parts 30, 40 and 70.
(2) A fire protection program must be established and implemented
before the initial receipt of byproduct, source, or
[[Page 87059]]
non-fuel SNM (excluding exempt quantities as described in Sec. 30.18
of this chapter).
(3) An emergency plan appropriate for responding to the facility-
specific hazards of an accidental release of radioactive material and
to limit the health effects of the associated chemical hazards of
licensed material must be approved and implemented prior to the receipt
of byproduct, source, or SNM (excluding exempt quantities as described
in Sec. 30.18 of this chapter).
(4) A plant staff training program associated with the receipt of
radioactive material must be approved and implemented before initial
receipt of byproduct, source, or SNM (excluding exempt quantities as
described in Sec. 30.18 of this chapter).
(5) Security requirements must be implemented for the protection of
SNM based on the type, enrichment, and quantity in accordance with 10
CFR part 73, as applicable, and for the protection of Category 1 and
Category 2 quantities of radioactive material in accordance with 10 CFR
part 37, as applicable.
(d) Fuel loading. (1)(i) An ML may authorize possession of a
manufactured reactor into which the licensee has loaded fresh
(unirradiated) fuel pursuant to a license issued under part 70 of this
chapter only if the manufactured reactor is configured during its
loading, storage, and transport with at least two independent physical
mechanisms in place, each of which is sufficient to prevent criticality
assuming optimum neutron moderation and neutron reflection conditions.
(ii) The ML applicant may file a separate, subsequent application
for the 10 CFR part 70 license or combine the application for the 10
CFR part 70 license with the application for an ML.
(iii) The Commission has determined that any such fueled
manufactured reactor in which the independent physical mechanisms to
prevent criticality have been installed is not in operation.
(iv) Upon installation of the fueled manufactured reactor in its
place of operation and a Commission finding that the acceptance
criteria in the COL that authorized reactor construction are met under
Sec. 53.1452(g), the independent physical mechanisms to prevent
criticality may be removed. Upon initiating the physical removal of any
one of the independent physical mechanisms to prevent criticality, the
fueled manufactured reactor has commenced operation.
(2) Holders of 10 CFR part 70 licenses authorizing the possession
and loading of fresh fuel into manufactured reactors must comply with
the requirements of 10 CFR part 70 for the facilities and activities
related to the storage, movement, and loading of fresh fuel in the
manufactured reactor. Holders of these 10 CFR part 70 licenses must
comply with the requirements of Subpart H to 10 CFR part 70, regardless
of whether their proposed activities meet the applicability criteria
found in 10 CFR 70.60. Procedures, equipment, and personnel required by
the 10 CFR part 70 license, must be in place before the receipt of SNM
at the manufacturing facility.
(i) Before the receipt of SNM, the licensee must have security
programs in place that meet the performance objectives of 10 CFR 73.67,
with the following additions and exceptions:
(A) A physical security plan describing the physical security
program must be maintained and a cybersecurity program must be
established for the possession and loading of fresh fuel into a
manufactured reactor authorized by a 10 CFR part 70 license, regardless
of fuel type, enrichment, and quantity.
(B) The physical security program must be designed to prevent
unintended and uncontrolled criticality events.
(C) The cybersecurity program must provide reasonable assurance
that a cyberattack would not adversely impact the functions performed
by digital assets used by the licensee for implementing the physical
security requirements of this section, or the radiation monitoring and
criticality requirements in this section or in 10 CFR part 70.
(D) All holders of a part 70 license that authorizes loading of
fresh fuel into a manufactured reactor must perform the screening
required in Sec. 73.67(d)(4) of this chapter to confirm the identity,
trustworthiness, and reliability of individuals prior to granting
unescorted access to special nuclear material; these determinations
must be documented.
(ii) [Reserved]
(3) The loading or unloading of fresh fuel into or from a
manufactured reactor and any changes to the configuration of reactivity
control and prevention systems for the fueled manufactured reactor must
be performed by a certified fuel handler meeting the requirements in
subpart F of this part.
(e) Transportation. (1) A holder of an ML may not transport or
allow to be removed from the places of manufacture the manufactured
reactor or portions thereof as defined in the ML except for transport
to a site for which the Commission has issued a COL that references the
subject ML.
(2) A holder of an ML must include in any contract governing the
transport of a manufactured reactor or portions thereof as defined in
the ML from the places of manufacture to any other location, a
provision requiring that the person transporting the manufactured
reactor comply with all shipping requirements in applicable NRC
regulations, certificates of compliance, and NRC-issued licenses.
(3) Procedures governing the preparation of the manufactured
reactor or portions thereof as defined in the ML for transport and the
conduct of the transport must be issued prior to transport. The
procedures must implement the protective measures and restrictions
described in NRC regulations and NRC-issued licenses to protect the
reactor from potential conditions that would adversely affect the safe
operation of a commercial nuclear plant.
(4) For a manufactured reactor that is to be loaded with fresh fuel
before transport to the place of operation, the ML must specify that
transportation will be in accordance with parts 71 and 73 of this
chapter.
(f) Acceptance and installation at the site for which the
Commission has issued a COL that references the subject ML. (1)
Installation at the site for which the Commission has issued a COL that
references the subject ML must follow the regulations in Sec. 53.610.
(2) Upon arrival at the site, the manufactured reactor or portions
of a manufactured reactor may not be installed in its place of
operation unless the COL holder performs inspections sufficient to
verify the reactor is in compliance with the ML and has not been
damaged in transit. The COL holder must perform these inspections in
accordance with documented procedures subject to quality assurance
measures commensurate with their importance to safety. In addition,
inspections must confirm that the interface requirements between the
manufactured reactor or portions of a manufactured reactor and the
remaining portions of the commercial nuclear plant are met.
Subpart F--Requirements for Operation
Sec. 53.700 Operational objectives.
(a) Each holder of an operating license (OL) or combined license
(COL) under this part must develop, implement, and maintain controls
for plant structures, systems, and components (SSCs), responsibilities
of plant personnel, and plant programs during the operating life of
each commercial nuclear plant such that the requirements defined in
subpart B are satisfied. More specifically:
[[Page 87060]]
(1) Each holder of an OL or COL under this part must maintain the
capabilities, availability, and reliability of plant SSCs to ensure
that the safety functions identified in Sec. 53.230 will be performed
if called upon during licensing-basis events (LBEs).
(2) Each holder of an OL or COL under this part must ensure that
plant personnel have adequate knowledge and skills to perform their
assigned duties that support the performance of the safety functions
identified in Sec. 53.230.
(3) Each holder of an OL or COL under this part must implement
plant programs sufficient to ensure that the safety functions
identified in Sec. 53.230 will be performed if called upon during
normal operations and LBEs.
(b) [Reserved]
Sec. 53.710 Maintaining capabilities and availability of structures,
systems, and components.
Controls must be provided for each commercial nuclear plant
licensed under this part such that the capabilities, availability, and
reliability of plant SSCs, when combined with corresponding
programmatic controls and human actions, provide that the safety
criteria defined in Sec. Sec. 53.210 and 53.220 will be met.
(a) Technical specifications must be developed, implemented, and
maintained that define conditions or limitations on plant operations
that are necessary to ensure that safety-related (SR) SSCs can fulfill
the safety functions identified under Sec. 53.230 and support meeting
the safety criteria of Sec. 53.210. The technical specifications must
describe the following requirements:
(1) Limits on the inventory of radioactive materials within the
reactor system and supporting systems with the potential, individually
or collectively, to cause a release exceeding the safety criteria in
Sec. 53.210 as a result of a design-basis accident analyzed in
accordance with Sec. 53.450(f).
(2) Operating limits for the facility that if exceeded could lead
to a failure to perform a required safety function necessary to
demonstrate compliance with the safety criteria in Sec. 53.210.
(3) For each SSC classified as SR in accordance with Sec. 53.460,
technical specifications must define--
(i) Limiting conditions for operation. Limiting conditions for
operation are the lowest functional capability or performance levels of
SR SSCs required to ensure that the design-basis accidents analyzed in
accordance with Sec. 53.450(f) satisfy the safety criteria of Sec.
53.210. When a limiting condition for operation is not met, the
licensee must shut down the plant or follow any remedial action
permitted by the technical specifications until the condition can be
met.
(ii) Surveillance requirements. Surveillance requirements are
requirements relating to test, calibration, or inspection to assure
that the necessary quality of systems and components is maintained and
that the limiting conditions for operation will be met.
(4) Design elements to be included are those elements of the plant
such as materials of construction and geometric arrangements, which, if
altered or modified, would have a significant effect on safety and are
not covered in categories described in paragraphs (a)(1) through (3) of
this section.
(5) Administrative controls are the provisions relating to
organization and management, procedures, recordkeeping, review and
audit, and reporting necessary to assure operation of the plant in a
safe manner. Each licensee must submit any reports to the Commission
pursuant to approved technical specifications under Sec. 53.040.
(b) Controls on plant operations, including availability controls,
must be developed and implemented to ensure that the configurations and
special treatments for SR SSCs and non-safety-related but safety-
significant (NSRSS) SSCs provide the capabilities, availability, and
reliability required to demonstrate compliance with the criteria of
Sec. Sec. 53.220 and 53.450(e).\1\ The controls must--
\1\ The comprehensive risk metrics and related risk performance
objectives established under Sec. 53.220 involve assessing and
averaging the risks over a defined period (e.g., plant year) and do
not constitute a real-time requirement that must be continuously
demonstrated by the licensee.
(1)(i) Identify who within the commercial nuclear plant has
authority to make configuration changes;
(ii) Establish processes to make configuration changes to NSRSS
SSCs; and
(iii) Establish processes to ensure that all organizations of the
commercial nuclear plant affected by the configuration changes are
formally notified and approve of the change.
(2) Describe how the special treatments for each NSRSS SSC and
special treatments for SR SSCs beyond those under paragraph (a) of this
section will be established and maintained over the operating life of
the commercial nuclear plant.
Sec. 53.715 Maintenance, repair, and inspection programs.
(a) A program to control maintenance activities and monitor the
performance or condition of SR and NSRSS SSCs must be developed,
implemented, and maintained.
(b) Whenever a licensee determines through activities related to
maintenance, repair, and inspection of SSCs, the activities under Sec.
53.710, or otherwise that the performance or condition of an SR or
NSRSS SSC does not demonstrate compliance with established special
treatments or performance goals related to capabilities, availability,
or reliability, the licensee must take appropriate corrective action.
(c) Performance and condition monitoring activities and associated
goals and preventive maintenance activities must be evaluated at least
every 24 months. The evaluations must take into account, where
practical, industry-wide operating experience. Adjustments must be made
where necessary to ensure that the objective of preventing failures of
SSCs through maintenance is appropriately balanced against the
objective of minimizing unavailability of SSCs due to monitoring or
preventive maintenance.
(d) Before performing maintenance activities (including but not
limited to surveillance, post-maintenance testing, and corrective and
preventive maintenance), the licensee must assess and manage the
increase in risk that may result from the proposed maintenance
activities.
Sec. 53.720 Response to seismic events.
If vibratory ground motion exceeding that of the operating basis
earthquake ground motion or significant plant damage due to vibratory
ground motion occurs, the licensee must shut down the commercial
nuclear plant. If structures, systems, or components necessary for the
safe shutdown of the commercial nuclear plant are not available after
the occurrence of this vibratory ground motion, the licensee must
consult with the Commission and must propose a plan for the timely,
safe shutdown of the commercial nuclear plant. Prior to resuming
operations, the licensee must demonstrate to the Commission that those
features necessary for continued operation without undue risk to the
health and safety of the public or necessary to maintain the licensing
basis of the commercial nuclear plant were either not functionally
damaged or have been repaired.
Sec. 53.725 General staffing, training, personnel qualifications, and
human factors requirements.
(a) Two classes of commercial nuclear plants. Commercial nuclear
plants licensed under this part are either of the
[[Page 87061]]
class, based upon the similarity of operating and technical
characteristics of the plants in the class, of self-reliant-mitigation
facilities or of interaction-dependent-mitigation facilities. A
commercial nuclear plant is a self-reliant-mitigation facility if the
U.S. Nuclear Regulatory Commission (NRC) determined as part of its
approval of the OL or COL for that plant that its design demonstrates
compliance with the criteria of Sec. 53.800(a)(1) through (a)(5).
Otherwise, the commercial nuclear plant is an interaction-dependent-
mitigation facility.
(b) Purpose and applicability. The regulations in Sec. Sec. 53.725
through 53.830 address areas related to staffing, training, personnel
qualifications, and human factors engineering for applicants for or
holders of OLs or COLs under this part. These regulations are organized
as follows:
(1) Sections 53.725 through 53.745 address general requirements for
staffing, training, personnel qualifications, and human factors
engineering. The regulations within these sections are applicable to
all applicants for or holders of OLs or COLs under this part, except
where specifically stated otherwise.
(2) Sections 53.760 through 53.795 address operator and senior
operator licensing requirements. The regulations within these sections
are applicable to those applicants for or holders of OLs or COLs under
this part for interaction-dependent-mitigation facilities that have not
yet certified the permanent cessation of operations and permanent
removal of fuel from the reactor vessel as described under Sec.
53.1070.
(3) Sections 53.800 through 53.820 address generally licensed
reactor operator requirements. The regulations within these sections
are in lieu of Sec. Sec. 53.760 through 53.795 for those applicants
for or holders of OLs or COLs under this part for self-reliant-
mitigation facilities that have not yet certified the permanent
cessation of operations and permanent removal of fuel from the reactor
vessel as described under Sec. 53.1070.
(4) Section 53.830 provides general personnel training
requirements. The regulations within this section are applicable to all
applicants for or holders of OLs or COLs under this part.
(c) Definitions. When used in Sec. Sec. 53.725 through 53.830:
Applicant refers to an applicant for an operator or senior operator
license; licensee refers to the holder of an operator, senior operator,
or generally licensed reactor operator license; and facility licensee
refers to the licensee for the commercial nuclear plant where the
applicant would be licensed or the licensee is licensed.
Automation means a device or system that accomplishes (partially or
fully) a function or task.
Auxiliary operator means any individual who operates components of
a commercial nuclear plant but does not manipulate controls or direct
the manipulation of controls of the plant and is not required to be
licensed under the provisions of this part.
Controls when used with respect to a nuclear reactor means
apparatus and mechanisms, the manipulation of which directly affects
the reactivity or power level of the reactor.
Generally licensed reactor operator means any individual licensed
under the provisions of Sec. 53.810 to manipulate controls of a self-
reliant-mitigation facility and to direct the licensed activities of
generally licensed reactor operators.
Interaction-dependent-mitigation facility means a commercial
nuclear plant design other than one that demonstrates compliance with
the operating and technical characteristics defined under Sec. 53.800.
Load following means a commercial nuclear plant automatically
changing its output to match expected demand in response to externally
originated instructions or signals.
Operator means any individual licensed under the provisions of
Sec. Sec. 53.760 through 53.795 to manipulate controls of an
interaction-dependent-mitigation facility.
Performance testing means testing conducted to verify a simulation
facility's performance as compared to actual or predicted reference
plant performance.
Reference plant means the specific commercial nuclear plant on
which a simulation facility's configuration, system control
arrangement, and design data are based. The reference plant may or may
not be constructed.
Self-reliant-mitigation facility means a commercial nuclear plant
design that demonstrates compliance with the operating and technical
characteristics defined under Sec. 53.800.
Senior operator means any individual licensed under the provisions
of Sec. Sec. 53.760 through 53.795 to manipulate controls of an
interaction-dependent-mitigation facility and to direct the licensed
activities of operators.
Simulation facility means an interface designed to provide a
realistic imitation of the operation of a commercial nuclear plant used
for the administration of examinations, for training, and/or to
demonstrate compliance with experience requirements for applicants or
licensees. A simulation facility may rely, in whole or part, upon the
physical utilization of the reference plant itself.
Systems approach to training means a training program that includes
the following five elements:
(1) Systematic analysis of the jobs to be performed.
(2) Learning objectives derived from the analysis which describe
desired performance after training.
(3) Training design and implementation based on the learning
objectives.
(4) Evaluation of trainee mastery of the objectives during
training.
(5) Evaluation and revision of the training based on the
performance of trained personnel in the job setting.
Sec. 53.726 Communications.
(a) An applicant or licensee or facility licensee must submit any
communication or report required by the regulations contained within
Sec. Sec. 53.725 through 53.830 and must submit any application filed
under these regulations to the Commission.
(b) Each licensee that is required to comply with the requirements
of Sec. Sec. 53.760 through 53.795 (i.e., interaction-dependent-
mitigation facilities) must notify the appropriate NRC contact within
30 days of the following in regard to a licensed operator or senior
operator:
(1) Permanent reassignment from the position for which the licensee
has certified the need for a licensed operator or senior operator under
Sec. 53.775(a)(1);
(2) Termination of any operator or senior operator; or
(3) Permanent disability or illness as required under Sec. 55.770
of this chapter.
Sec. 53.728 Completeness and accuracy of information.
Information provided to the Commission by an applicant for an
operator or senior operator license or by a licensee or information
required by statute or by the Commission's regulations, orders, or
license conditions to be maintained by the applicant or the licensee
must be complete and accurate in all material respects.
Sec. 53.730 Defining, fulfilling, and maintaining the role of
personnel in ensuring safe operations.
Each applicant for or holder of an OL or COL for a commercial
nuclear plant under this part must comply with the following:
(a) Human factors engineering design requirements. The plant design
must reflect state-of-the-art human factors engineering principles for
safe and reliable performance in all locations that
[[Page 87062]]
human activities are expected for performing or supporting the
continued availability of plant safety or emergency response functions.
(b) Human system interface design requirements. The plant design
must provide for the following to support operating personnel in
monitoring plant conditions and responding to plant events:
(1) Features for displaying to operating personnel a minimum set of
parameters that define the safety status of the plant and are capable
of displaying both the full range of important plant parameters and
data trends on demand, as well as indicating when process limits are
being approached or exceeded;
(2) Automatic indication of the bypassed and operable status of
safety systems;
(3) Direct indication of SSC status that relates to the ability of
the SSC to perform its safety function, such as relief and safety valve
position (i.e., open or closed) for barriers important to fulfilling
safety functions of with such devices, and ultimate heat sink and
cooling system status and availability;
(4) Instrumentation to measure, record, and display key plant
parameters related to the performance of SSCs and the integrity of
barriers important to fulfilling safety functions to support operators
in monitoring plant conditions and responding to plant events. Examples
include temperatures and pressures within important systems or
structures, core or fuel system conditions (including possible damage
states), temperatures and levels associated with cooling functions,
combustible gas concentrations, radiation levels in systems and within
structures, and radioactive effluent releases;
(5) Leakage control and detection in the design of systems that
pass through barriers important to fulfilling safety functions for the
release of radionuclides. An example is an SSC that penetrates a
containment structure that might contain radioactive materials that
could contribute to the source term during an accident;
(6) Monitoring of in-plant radiation and airborne radioactivity as
appropriate for a broad range of normal operating and accident
conditions; and
(7) For self-reliant-mitigation facilities, the plant design must
also provide the generally licensed reactor operators with the
capability to do the following:
(i) Receive plant operating data, including reactor parameters and
information needed for the evaluation of emergency conditions.
(ii) Immediately initiate a reactor shutdown from their location.
(iii) Promptly dispatch operations and maintenance personnel.
(iv) Immediately implement responsibilities under the facility
emergency plan, as applicable.
(c) Concept of operations. A concept of operations that is of
sufficient scope and detail to address the following must be provided:
(1) Plant goals;
(2) The roles and responsibilities of operating personnel and
automation (or any combination thereof) that are responsible for
completing plant functions;
(3) Staffing, qualifications, and training;
(4) The management of normal operations;
(5) The management of off-normal conditions and emergencies;
(6) The management of maintenance and modifications; and
(7) The management of tests, inspections, and surveillances.
(d) Functional requirements analysis and function allocation. A
functional requirements analysis and a function allocation must be
provided that are sufficient to demonstrate compliance with the
following:
(1) The functional requirements analysis must address how safety
functions and functional safety criteria are satisfied, and
(2) The function allocation must describe how the safety functions
will be assigned to human action, automation, active safety features,
passive safety features, and/or inherent safety characteristics.
(e) Operating experience. A program, during construction and during
operation, as applicable, for evaluating and applying operating
experience must be developed, implemented, and maintained.
(f) Staffing plan. A staffing plan must be developed and comply
with the following:
(1) The staffing plan must include a description of how engineering
expertise will be available to the on-shift operating personnel during
all plant conditions, to assist if they encounter a situation not
covered by procedures or training. Engineering expertise includes
familiarity with the operation of the plant for which the expertise is
provided and one of the following:
(i) A bachelor's degree in engineering, engineering technology, or
physical science from an institution accredited by a U.S. government
recognized accrediting body or equivalent; or
(ii) A Professional Engineer's license from a U.S. State or
territory.
(2) Applicants for or holders of OLs or COLs for interaction-
dependent-mitigation facilities must include within their staffing
plans a description of how the proposed numbers, positions, and
qualifications of operators and senior operators across all modes of
plant operations will be sufficient to ensure that plant safety
functions will be maintained. This description must be supported by
human factors engineering analyses and assessments.
(3) Applicants for or holders of OLs or COLs for self-reliant-
mitigation facilities must include within their staffing plans a
description of how generally licensed reactor operator staffing that is
both sufficient to continually monitor the operations of fueled
reactors and to provide for a continuity of responsibility for facility
operations at all times during the operating phase will be maintained.
(4) Applicants for or holders of OLs or COLs under this part must
include within their staffing plans a description of how the numbers,
positions, and responsibilities of personnel contained within those
plans will adequately support all necessary functions within areas such
as plant operations, equipment surveillance and maintenance,
radiological protection, chemistry control, fire brigades, engineering,
security, and emergency response.
(5) The staffing plan must be approved by the NRC as part of its
approval of the OL or COL for the plant. The approved staffing plan is
subject to the requirements of Sec. 53.1565.
(g) Training, examination, and proficiency programs. Develop,
implement, and maintain programs that comply with the following
requirements. These programs must be approved by the NRC as part of its
approval of the OL or COL for the plant:
(1) For those applicants for or holders of OLs or COLs for
interaction-dependent-mitigation facilities:
(i) The operator licensing initial training program required under
Sec. 53.780(a);
(ii) The operator licensing initial examination program required
under Sec. 53.780(b);
(iii) The operator licensing requalification program required under
Sec. 53.780(c); and
(iv) The operator proficiency program required under Sec.
53.780(g).
(2) For those applicants for or holders of OLs or COLs for self-
reliant-mitigation facilities, the generally licensed reactor operator
training, examination, and proficiency programs required under Sec.
53.815.
[[Page 87063]]
(3) The operator licensing requalification programs required under
Sec. 53.780(c) or Sec. 53.815(b) must be implemented upon commencing
the administration of initial examinations under the operator licensing
examination program required under Sec. 53.780(b) or Sec. 53.815(b),
respectively.
Sec. 53.735 General exemptions.
The regulations in Sec. Sec. 53.725 through 53.830 do not require
a license for an individual who--
(a) Under the direction and in the presence of an operator or
senior operator or a generally licensed reactor operator, as
appropriate, manipulates the controls of a commercial nuclear plant as
a part of the individual's training in a facility licensee's training
program as approved by the Commission to qualify for an operator or
senior operator license or a generally licensed reactor operator
license there, as appropriate, under these regulations; or
(b) Under the direction and in the presence of a senior operator or
generally licensed reactor operator, as appropriate, manipulates the
controls of a commercial nuclear plant to load or unload the fuel into,
out of, or within the reactor vessel while the reactor is not
operating.
Sec. 53.740 Facility licensee requirements--General.
(a) Facility licensees must demonstrate compliance with the
requirements of either Sec. Sec. 53.760 through 53.795 for
interaction-dependent-mitigation facilities or Sec. Sec. 53.800
through 53.820 for self-reliant-mitigation facilities.
(b) The facility licensee must maintain the staffing complement
described under its approved facility staffing plan until such time as
the permanent cessation of operations and permanent removal of fuel
from the reactor vessel has been certified as described under Sec.
53.1070. The approved staffing plan is subject to the requirements of
Sec. 53.1565.
(c) Except as provided under Sec. 53.735, the facility licensee
may not permit the manipulation of the controls of a commercial nuclear
plant by anyone who is not an operator or senior operator or generally
licensed reactor operator, as appropriate.
(d) Facility licensees for interaction-dependent-mitigation
facilities that have not yet certified the permanent cessation of
operations and permanent removal of fuel from the reactor vessel as
described under Sec. 53.1070 must designate senior operators to be
responsible for supervising the licensed activities of operators.
(e) Apparatus and mechanisms other than controls, the operation of
which may affect the reactivity or power level of a reactor, must be
manipulated only while plant conditions are being monitored by an
individual who is an operator or senior operator or a generally
licensed reactor operator, as appropriate.
(f)(1) Load following is permitted if at least one of the following
is immediately capable of refusing demands when they could challenge
the safe operation of the plant or when precluded by the plant
equipment conditions:
(i) The actuation of an automatic protection system that utilizes
setpoints more conservative than those otherwise credited for the
purposes of reactor protection; or
(ii) An automated control system; or
(iii) An operator or senior operator or a generally licensed
reactor operator, as appropriate.
(2) The provisions of paragraph (e) of this section do not apply
during load following operations.
(g)(1) Facility licensees for interaction-dependent-mitigation
facilities must have present during alteration of the core (including
fuel loading or transfer) an individual holding a senior operator
license, or a senior operator license limited to fuel handling to
directly supervise the activity and, during this time, the facility
licensee must not assign other duties to this person.
(2) Facility licensees for self-reliant-mitigation facilities must
have present during alteration of the core (including fuel loading or
transfer) an individual holding a generally licensed reactor operator
license to directly supervise the activity and, during this time, the
facility licensee must not assign other duties to this person.
(3) The provisions of paragraphs (g)(1) and (2) of this section do
not apply to core alterations performed as part of refueling operations
while a facility that is capable of online refueling is operating at
power.
(h) Facility licensees may take reasonable action that departs from
a license condition or a technical specification (contained in a
license issued under this part) in an emergency when this action is
immediately needed to protect the public health and safety and no
action consistent with license conditions and technical specifications
that can provide adequate or equivalent protection is immediately
apparent. Such facility licensee action must be approved, as a minimum,
by a senior operator or a generally licensed reactor operator, as
applicable, or, after certifying the permanent cessation of operations
and permanent removal of fuel from the reactor vessel as described
under Sec. 53.1070 by a certified fuel handler, senior operator, or
generally licensed reactor operator, as applicable, prior to taking the
action.
Sec. 53.745 Operator license requirements.
A person must be authorized by a license issued by the Commission
to perform the function of an operator, senior operator, or generally
licensed reactor operator as defined in this part.
Sec. 53.760 Operator licensing.
(a) Applicability. Sections 53.760 through 53.795 address operator
and senior operator licensing requirements. The regulations within
these sections are applicable to those applicants for or holders of OLs
or COLs under this part for interaction-dependent-mitigation facilities
that have not yet certified the permanent cessation of operations and
permanent removal of fuel from the reactor vessel as described under
Sec. 53.1070.
(b) Reserved.
Sec. 53.765 Medical requirements.
(a) An applicant for an operator or senior operator license must
have a medical examination by a physician. An operator or senior
operator must have a medical examination by a physician every 2 years.
(b) To certify the medical fitness of an applicant for an operator
or senior operator license, an authorized representative of the
facility licensee must complete and sign NRC Form 396, ``Certification
of Medical Examination by Facility Licensee,'' which can be obtained by
writing the Office of the Chief Information Officer, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, by calling 301-415-
7232, or by visiting the NRC's website at https://www.nrc.gov and
selecting forms from the index found on the home page, or by other
means provided by the NRC.
(1) Form NRC 396 must certify that a physician has conducted the
medical examination of the applicant as required in paragraph (a) of
this section.
(2) When the medical certification requests a conditional license
based on medical evidence, the medical evidence must be submitted on
NRC Form 396 to the Commission to enable the Commission to make a
determination in accordance with Sec. 53.775(b).
(c) The facility licensee must document and maintain the results of
medical qualifications data, test results,
[[Page 87064]]
and each operator's or senior operator's medical history for the
current license period and provide the documentation to the Commission
upon request. The facility licensee must retain this documentation
while an individual performs the functions of an operator or senior
operator.
Sec. 53.770 Incapacitation because of disability or illness.
If, during the term of the operator or senior operator license, the
licensee develops a permanent physical or mental condition that causes
the licensee to fail to demonstrate compliance with the requirements of
Sec. 53.775(b)(1)(i), the facility licensee must notify the Commission
within 30 days of learning of the diagnosis. For conditions for which a
conditional license (as described in Sec. 53.775(b)) is requested, the
facility licensee must provide medical certification on Form NRC 396 to
the Commission (as described in Sec. 53.765(b)).
Sec. 53.775 Applications for operators and senior operators.
(a) How to apply. (1) The applicant for an operator or senior
operator license must--
(i) Complete NRC Form 398, ``Personal Qualification Statement--
Licensee,'' which can be obtained by writing the Office of the Chief
Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, by calling 301-415-5877, or by visiting the NRC's website
at https://www.nrc.gov and selecting forms from the index found on the
home page, or by other means provided by the NRC;
(ii) File an original of NRC Form 398, or an equivalent electronic
submittal, together with the information required in paragraphs
(a)(1)(iii) and (a)(1)(iv) of this section, with the appropriate
Regional Administrator.
(iii) Provide evidence that the applicant, as a trainee, has
successfully demonstrated competence in manipulating the controls of
either the facility for which a license is sought or a simulation
facility that demonstrates compliance with the requirements of Sec.
53.780(e). For operators applying for a senior operator license,
certification that the operator has successfully operated the controls
of the facility as an operator will be accepted; and
(iv) Provide certification by the facility licensee of medical
condition and general health on Form NRC 396, to comply with Sec.
53.765.
(2) The Commission may at any time after the application has been
filed, and before the license has expired, require further information
under oath or affirmation to enable it to determine whether to grant or
deny the application or whether to revoke, modify, or suspend the
license.
(3) An applicant whose application has been denied because of a
medical condition or their general health may submit a further medical
report at any time as a supplement to the application.
(4) Each application and statement must contain complete and
accurate disclosure as to all matters required to be disclosed. The
applicant must sign statements required by paragraphs (a)(1)(i) and
(a)(1)(ii) of this section.
(b) Disposition of an initial application. (1) License approval.
The Commission will approve an initial application if it finds that the
following criteria are met:
(i) Health. The applicant's medical condition and general health
will not adversely affect the performance of assigned operator or
senior operator job duties or cause operational errors endangering
public health and safety. The Commission will base its finding upon the
certification by the facility licensee as detailed in Sec. 53.765(b).
(ii) Examination. The applicant has passed the requisite
examination in accordance with Sec. 53.780(b). The examination
determines whether the applicant for an operator's or senior operator's
license has learned to operate a facility competently and safely, and
additionally, in the case of a senior operator, whether the applicant
has learned to supervise the licensed activities of operators
competently and safely.
(2) Conditional license. If an applicant's general medical
condition does not demonstrate compliance with the minimum standards
under Sec. 53.775(b)(1)(i) of this section, the Commission may approve
the application and include conditions in the license to accommodate
the medical condition. The Commission will consider the recommendations
and supporting evidence of the facility licensee and of the examining
physician (provided on Form NRC 396) in arriving at its decision.
(c) Re-applications. (1) An applicant whose application for a
license has been denied because of failure to pass the examination may
file a new application. The application must be submitted on Form NRC
398 and include a statement signed by an authorized representative of
the facility licensee by whom the applicant will be employed that
states in detail the extent of the applicant's additional training and
remediation since the denial and certifies that the applicant is ready
for re-examination.
(2) An applicant who has passed a portion of the examination and
failed another may request in a new application on Form NRC 398 to be
excused from re-examination on the portions of the examination that the
applicant has passed. The Commission may in its discretion grant the
request if it determines that sufficient justification is presented.
Sec. 53.780 Training, examination, and proficiency program.
(a) Operator licensing initial training program. (1) A program that
is based upon a systems approach to training, as defined by Sec.
53.725(b), must be utilized for the training of applicants for operator
and senior operator licenses. The program must ensure that applicants
at the facility will possess the knowledge, skills, and abilities
necessary to protect the public health and maintain those plant safety
functions specific to the facility design. The program must be approved
by the Commission prior to its use for training applicants, as
described under Sec. 53.730(g). The approved operator licensing
initial training program is subject to the requirements of Sec.
53.1565.
(2) The facility licensee must maintain operator licensing initial
training program records documenting the initial operator licensing
training administered and completed by each applicant. The facility
licensee must retain these records during the period in which any
trainees subsequently remain licensed as operators or senior operators
at the facility.
(b) Operator licensing initial examination program. (1) The
facility licensee must establish and implement an examination program
for testing a representative sample of the knowledge, skills, and
abilities needed to safely perform operator and senior operator duties,
to include both the examination methods and criteria to be used to
assess passing performance. The program must provide for valid and
reliable examinations and be approved by the Commission prior to its
use for examining applicants, as described under Sec. 53.730(g). The
approved operator licensing initial examination program is subject to
the requirements of Sec. 53.1565.
(2) The facility licensee must submit prepared examinations to the
Commission for review and approval in advance of their administration.
(3) The Commission will either administer an approved examination
or allow the facility licensee to administer the examination. The
facility licensee must ensure that sufficient advance notification is
provided to the
[[Page 87065]]
Commission to either administer the examination or allow for a
representative of the Commission to be afforded the opportunity to be
present when the facility licensee administers the examination.
(4) Graded examination documentation for each applicant must be
promptly provided to the Commission for review in making operator
licensing decisions.
(5) The facility licensee must maintain operator licensing initial
examination program records documenting the participation of each
operator and senior operator applicant in the initial examination. The
records must contain copies of examinations administered, the answers
given by the applicant, documentation of the grading of examinations,
and documentation of any additional training administered in areas in
which an applicant exhibited deficiencies. The facility licensee must
retain these records during the period in which the associated
operators or senior operators remain licensed at the facility.
(c) Operator licensing requalification program. (1) A program based
upon a systems approach to training, as defined by Sec. 53.725(b),
must be utilized for the continuing training of operators and senior
operators.
(i) The program must ensure that operators and senior operators at
the facility maintain the knowledge, skills, and abilities necessary to
protect the public health and maintain those plant safety functions
specific to the facility design. The program must be conducted for a
continuous period not to exceed 24 months in duration.
(ii) The program must be approved by the Commission prior to its
use for continuing training, as described under Sec. 53.730(g). The
approved operator licensing requalification program is subject to the
requirements of Sec. 53.1565.
(2) The following requirements apply to operator licensing
requalification programs:
(i) The facility licensee must propose a requalification
examination program for testing, for each requalification period, a
sample of the topics included under the systems approach to training,
to include both the examination methods and criteria to be used to
assess passing performance. The program must provide for valid and
reliable examinations and be approved by the Commission prior to its
use for examining operators and senior operators, as described under
Sec. 53.730(g). The approved requalification examination program is
subject to the requirements of Sec. 53.1565.
(ii) The following requirements apply to the requalification
examination program:
(A) The facility licensee must make prepared requalification
examinations available to the Commission for review.
(B) The facility licensee must ensure that a representative of the
Commission is afforded the opportunity to be present during
requalification examination administration.
(C) The facility licensee must ensure that each operator and senior
operator is administered a complete requalification examination on a
periodicity not to exceed 24 months. Additionally, the facility
licensee must ensure that any licensed operator or senior licensed
operator who either demonstrates unsatisfactory performance on, or
fails to complete, the biennial requalification examination is removed
from the performance of licensed operator and senior licensed operator
duties until such time that any necessary remedial training has been
completed and a retake examination has been passed.
(D) The facility licensee must promptly provide a summary of
examination results for each operator and senior operator following the
completion of the requalification examination.
(3) The facility licensee must maintain operator licensing
requalification program records documenting the participation of each
operator and senior operator in the requalification program. The
records must contain copies of examinations administered, the answers
given by the operator or senior operator, documentation of the grading
of examinations, and documentation of any additional training
administered in areas in which an operator or senior operator exhibited
deficiencies. The facility licensee must retain these records until the
operator's or senior operator's license is renewed.
(d) Examination integrity. Applicants, operators and senior
operators, and facility licensees must not engage in any activity that
compromises the integrity of any application or examination required by
Sec. Sec. 53.760 through 53.795. The integrity of an examination is
considered compromised if any activity, regardless of intent, affected,
or, but for detection, could have affected the equitable and consistent
administration of the examination. This includes activities related to
the preparation and certification of applications and all activities
related to the preparation, administration, and grading of examinations
required by Sec. Sec. 53.760 through 53.795.
(e) Simulation facilities. (1) This section addresses the use of a
simulation facility for the administration of examinations, for
training, or to demonstrate compliance with experience requirements for
applicants for operator and senior operator licenses.
(2) Simulation facilities used for training purposes, for
demonstrating compliance with experience requirements, or for the
conduct of examinations under Sec. 53.780(b) and (c) must demonstrate
compliance with the following criteria as they relate to the facility
licensee's reference plant:
(i) The simulation facility must be of sufficient scope and
fidelity for individuals to acquire and demonstrate the necessary
knowledge, skills, and abilities to safely perform operator and senior
operator duties.
(ii) The simulation facility must utilize models relating to
nuclear, thermal-hydraulic, and other applicable design-specific
characteristics that either replicate the most recent fuel load in the
reference commercial nuclear plant or, prior to initial fuel load (or,
for a fueled manufactured reactor, prior to initiating the physical
removal of any one of the independent physical mechanisms to prevent
criticality required under Sec. 53.620(d)(1)), replicate the intended
initial fuel load for the reference commercial nuclear plant, with the
exception of those portions of the simulation facility that utilize the
reference plant itself.
(iii) Simulation facility fidelity must be demonstrated so that
significant control manipulations are completed without procedural
exceptions, simulator performance exceptions, or deviation from the
approved training scenario sequence.
(3) Facility licensees that maintain a simulation facility that has
been approved by the Commission for training purposes, demonstrating
compliance with experience requirements, or the conduct of examinations
under Sec. 53.780(b) and (c) for the facility licensee's reference
plant must:
(i) Conduct performance testing throughout the life of the
simulation facility in a manner sufficient to ensure that paragraph
(e)(2) of this section is met;
(ii) Retain the results of performance testing for 4 years after
the completion of each performance test or until superseded by updated
test results;
(iii) Promptly correct modeling and hardware discrepancies and
discrepancies identified from scenario validation and from performance
testing or provide justification as to why the
[[Page 87066]]
presence of such discrepancies will not adversely affect simulator
performance with respect to the criteria of paragraph (e)(2) of this
section;
(iv) Make the results of any uncorrected performance test failures
that may exist at the time of the initial license examination or
requalification examination available for NRC review, prior to or
concurrent with preparations for each initial license examination or
requalification examination; and
(v) Maintain the provisions for license application and examination
integrity consistent with Sec. 53.780(d).
(4) A simulation facility must demonstrate compliance with the
requirements of paragraphs (e)(2) and (e)(3) of this section for the
Commission to accept the simulation facility for conducting initial
examinations as described in Sec. 53.780(b), requalification training
as described in Sec. 53.780(c), or performing control manipulations
that affect reactivity to establish eligibility for an operator or
senior operator license as described in Sec. 53.775(a).
(f) Waiver of examination requirement. On application, the
Commission may waive any or all of the requirements for an examination
if it finds that the applicant has demonstrated the required knowledge,
skills, and abilities to safely operate the plant, and is capable of
continuing to do so. The Commission may make such a finding based on
demonstration of the following:
(1) Operating experience at a comparable facility;
(2) Proof of the applicant's past competent and safe performance;
and
(3) Proof of the applicant's current qualifications.
(g) Proficiency. The facility licensee must develop, implement, and
maintain a proficiency program to ensure that operators and senior
operators will actively perform the functions of an operator or senior
operator, respectively, as needed to maintain proficiency with on-shift
duties and familiarity with plant status. This program must include
those steps that will be taken to re-establish proficiency when it
cannot be maintained. This program must be approved by the Commission
as part of its approval of the OL or COL for the plant. The approved
proficiency program is subject to the requirements of Sec. 53.1565.
(h) Records. Each record required by this section must be legible
throughout the retention period specified by each Commission
regulation. The record may be the original, a reproduced copy, or an
electronic copy provided that the copy is authenticated by authorized
personnel.
Sec. 53.785 Conditions of operator and senior operator licenses.
Each operator and senior operator license contains and is subject
to the following conditions whether stated in the license or not:
(a) Neither the license nor any right under the license may be
assigned or otherwise transferred.
(b) The license is limited to the facility for which it is issued.
(c) The license is limited to those controls of the facility or
facilities specified in the license.
(d) The license is subject to, and the licensee must observe, all
applicable rules, regulations, and orders of the Commission.
(e) The licensee must maintain or re-establish proficiency in
accordance with the facility licensee's Commission-approved proficiency
program required under Sec. 53.780(g).
(f) The licensee must be subject to the facility's Commission-
approved operator licensing requalification and requalification
examination programs required under Sec. 53.780(c).
(g) The licensee must have a biennial medical examination as
described by Sec. 53.765.
(h) The licensee must notify the Commission within 30 days about a
conviction for a felony.
(i) The licensee must not consume or ingest alcoholic beverages
within the protected area of commercial nuclear plants. The licensee
must not use, possess, or sell any illegal drugs. The licensee must not
perform activities authorized by a license issued under this part while
under the influence of alcohol or any prescription, over-the-counter,
or illegal substance that could adversely affect his or her ability to
safely and competently perform his or her licensed duties. For the
purpose of this paragraph, with respect to alcoholic beverages and
drugs, the term ``under the influence'' means the licensee exceeded, as
evidenced by a confirmed test result, the lower of the cutoff levels
for drugs or alcohol contained in 10 CFR part 26, or as established by
the facility licensee. The term ``under the influence'' also means the
licensee could be mentally or physically impaired as a result of
substance use including prescription and over-the-counter drugs, as
determined under the provisions, policies, and procedures established
by the facility licensee for its fitness-for-duty program, in such a
manner as to adversely affect his or her ability to safely and
competently perform licensed duties.
(j) Each licensee must participate in the drug and alcohol testing
programs as required under 10 CFR part 26.
(k) The licensee must comply with any other conditions that the
Commission may impose to protect health or to minimize danger to life
or property.
Sec. 53.790 Issuance, modification, and revocation of operator and
senior operator licenses.
(a) Issuance of operator and senior operator licenses. If the
Commission determines that an applicant for an operator license or a
senior operator license demonstrates compliance with the requirements
of the Atomic Energy Act of 1954, as amended, (the Act) and its
regulations, it will issue a license in the form and containing any
conditions and limitations it considers appropriate and necessary.
(b) Modification and revocation of operator and senior operator
licenses. (1) The terms and conditions of all operator and senior
operator licenses are subject to amendment, revision, or modification
by reason of rules, regulations, or orders issued in accordance with
the Act or any amendments thereto.
(2) Any license may be revoked, suspended, or modified, in whole or
in part--
(i) For any material false statement in the application or in any
statement of fact required under section 182 of the Act;
(ii) Because of conditions revealed by the application or statement
of fact or any report, record, inspection, or other means that would
warrant the Commission to refuse to grant a license on an original
application;
(iii) For willful violation of, or failure to observe, any of the
terms and conditions of the Act or the license, or of any rule,
regulation, or order of the Commission;
(iv) For any conduct determined by the Commission to be a hazard to
safe operation of the facility; or
(v) For the sale, use, or possession of illegal drugs, or refusal
to participate in the facility drug and alcohol testing program, or a
confirmed positive test for drugs, drug metabolites, or alcohol in
violation of the conditions and cutoff levels established by Sec.
53.785(i) or the consumption of alcoholic beverages within the
protected area of commercial nuclear plants, or a determination of
unfitness for scheduled work as a result of the consumption of
alcoholic beverages.
Sec. 53.795 Expiration and renewal of operator and senior operator
licenses.
(a) Expiration. (1) Each operator license and senior operator
license expires 6 years after the date of
[[Page 87067]]
issuance, upon termination of employment with the facility licensee, or
upon determination by the facility licensee that the licensed
individual no longer needs to maintain a license.
(2) If a licensee files an application for renewal or an upgrade of
an existing license on Form NRC 398 at least 30 days before the
expiration of the existing license, it does not expire until
disposition of the application for renewal or for an upgraded license
has been finally determined by the Commission. Filing by mail will be
deemed to be complete at the time the application is postmarked
(b) Renewal. (1) The applicant for renewal of an operator license
or senior operator license must--
(i) Complete and sign Form NRC 398 and include the number of the
license for which renewal is sought.
(ii) File an original of NRC Form 398 as specified in Sec. 53.775.
(iii) Provide written evidence of the applicant's experience under
the existing license and the approximate number of hours that the
licensee has operated the facility.
(iv) Provide a statement by an authorized representative of the
facility licensee that during the effective term of the current license
the applicant has satisfactorily completed the requalification program
for the facility for which operator or senior operator license renewal
is sought.
(v) Provide evidence that the applicant has discharged the license
responsibilities competently and safely. The Commission may accept as
evidence of the applicant's having met this requirement a certificate
of an authorized representative of the facility licensee or holder of
an authorization by which the licensee has been employed.
(vi) Provide certification by the facility licensee of medical
condition and general health on Form NRC 396, to comply with Sec.
53.765.
(2) The license will be renewed if the Commission finds that--
(i) The medical condition and the general health of the licensee
continue to be such as not to cause operational errors that endanger
public health and safety. The Commission will base this finding upon
the certification by the facility licensee as described in Sec.
53.765(b).
(ii) The licensee--
(A) Is capable of continuing to competently and safely assume
licensed duties;
(B) Has successfully completed a requalification program that has
been approved by the Commission as required by Sec. 53.780(c); and
(C) Has passed the requalification examinations as required by
Sec. 53.780(c).
(iii) There is a continued need for an operator to operate or for a
senior operator to supervise operators at the facility designated in
the application.
(iv) The past performance of the licensee has been satisfactory to
the Commission. In making its finding, the Commission will include in
its evaluation information such as notices of violations or letters of
reprimand in the licensee's docket.
Sec. 53.800 Facility licensees for self-reliant-mitigation
facilities.
(a) A commercial nuclear plant is a self-reliant-mitigation
facility if the NRC determined as part of its approval of the OL or COL
for that plant that its design demonstrates compliance with criteria
(a)(1) though (a)(5) of this section. A self-reliant-mitigation
facility is of a class, based upon the similarity of operating and
technical characteristics of the plants in the class, such that its
licensee must comply with the requirements of Sec. Sec. 53.800 through
53.820 in lieu of those in Sec. Sec. 53.760 through 53.795.
(1) The safety performance criteria of Sec. Sec. 53.210 and 53.220
and, if applicable, any alternative criteria used in accordance with
Sec. 53.470, must be met without reliance upon human action for
credited event mitigation.
(2) The results of a probabilistic risk analysis must demonstrate
that the evaluation criteria for the events analyzed in accordance with
Sec. 53.450 will be met without reliance on human actions to achieve
acceptable event mitigation.
(3) The functional requirements analysis and function allocation
performed under Sec. 53.730(d) must demonstrate that functions
required for safety are not reliant upon credited human action.
(4) The plant response to events analyzed under Sec. 53.450 must
rely exclusively on safety features and characteristics that will
neither be rendered unavailable by credible human errors of commission
or omission nor credibly require manual human operation in response to
equipment failures. Compliance with this paragraph may be achieved
through the use of SSCs that function through inherent characteristics
or that have engineered protections against human failures.
(5) The plant design must provide for a layered defense-in-depth
approach that is not dependent upon any single barrier or credited
human action.
(b) [Reserved]
Sec. 53.805 Facility licensee requirements related to generally
licensed reactor operators.
(a) Licensees for self-reliant-mitigation facilities that have not
yet certified the permanent cessation of operations and permanent
removal of fuel from the reactor vessel as described under Sec.
53.1070 must demonstrate compliance with the following requirements:
(1) Ensure that, in addition to being qualified to perform those
items identified by the facility-specific systems approach to training
conducted under Sec. 53.815, generally licensed reactor operators are
qualified to safely and competently--
(i) Perform administrative tasks, including compliance with
technical specifications, and perform operability determinations;
(ii) Implement maintenance and configuration controls;
(iii) Comply with radioactive release limitations;
(iv) Understand plant operating data, including reactor parameters,
and evaluate emergency conditions;
(v) Initiate a reactor shutdown from necessary locations;
(vi) Dispatch and direct operations and maintenance personnel;
(vii) Implement any applicable responsibilities under the facility
emergency plan; and
(viii) Make required notifications to local, State, participating
Tribal and Federal authorities.
(2) Develop, implement, and maintain facility technical
specifications that provide the necessary administrative controls to
ensure the implementation of these requirements.
(3) Develop, implement, and maintain the generally licensed reactor
operator training, examination, and proficiency programs required under
Sec. 53.815.
(4) Ensure that generally licensed reactor operators are subject to
the facility's generally licensed reactor operator training,
examination, and proficiency programs required under Sec. 53.815.
Ensure that generally licensed reactor operators are subject to and
comply with the applicable programmatic requirements for plant
personnel required under 10 CFR parts 26 and 73. An individual that is
not in compliance with any of these programs is not qualified to be in
a position that may involve the manipulation of the controls of the
commercial nuclear plant.
(5) Report annually to the NRC the identity of all generally
licensed reactor operators at the commercial nuclear plant, including
all additions and deletions since the previous report.
[[Page 87068]]
(6) Ensure that the facility design continues to meet the criteria
of Sec. 53.800.
(b) [Reserved]
Sec. 53.810 Generally licensed reactor operators.
(a) A general license to manipulate the controls of a self-reliant-
mitigation facility and to direct the licensed activities of generally
licensed reactor operators is hereby issued to any individual employed
in a position that may involve the manipulation of the controls of that
self-reliant-mitigation facility and who observes the restrictions of
this section.
(b) A generally licensed reactor operator must comply with the
operating procedures and other conditions specified in the license
authorizing operation of the facility.
(c) The general license is limited to the facility or facilities at
which the operator is employed.
(d) The Commission will suspend the general license on an
individual basis for violations of any provision of the Act or any rule
or regulation issued thereunder whenever the Commission deems such
suspension desirable, including--
(1) For willful violation of, or failure to observe, any of the
terms and conditions of the Act or the general license, or of any rule,
regulation, or order of the Commission;
(2) For any conduct determined by the Commission to be a hazard to
safe operation of the facility; or
(3) For the sale, use, or possession of illegal drugs, or refusal
to participate in the facility drug and alcohol testing program, or a
confirmed positive test for drugs, drug metabolites, or alcohol in
violation of the conditions and cutoff levels established by Sec.
53.810(f) or the consumption of alcoholic beverages within the
protected area of commercial nuclear plants, or a determination of
unfitness for scheduled work as a result of the consumption of
alcoholic beverages.
(e) The Commission may require information from a generally
licensed reactor operator to determine whether a general license should
be revoked or suspended with respect to that operator.
(f) The generally licensed reactor operator must not consume or
ingest alcoholic beverages within the protected area of commercial
nuclear plants. The generally licensed reactor operator must not use,
possess, or sell any illegal drugs. The generally licensed reactor
operator must not perform activities requiring a general license while
under the influence of alcohol or any prescription, over-the-counter,
or illegal substance that could adversely affect his or her ability to
safely and competently perform these activities. For the purpose of
this paragraph, with respect to alcoholic beverages and drugs, the term
``under the influence'' means the generally licensed reactor operator
exceeded, as evidenced by a confirmed test result, the lower of the
cutoff levels for drugs or alcohol contained in 10 CFR part 26, or as
established by the facility licensee. The term ``under the influence''
also means the generally licensed reactor operator could be mentally or
physically impaired as a result of substance use including prescription
and over-the-counter drugs, as determined under the provisions,
policies, and procedures established by the facility licensee for its
fitness-for-duty program, in such a manner as to adversely affect his
or her ability to safely and competently perform generally licensed
reactor operator duties.
(g) The generally licensed reactor operator must notify the
Commission within 30 days about a conviction for a felony.
Sec. 53.815 Generally licensed reactor operator training,
examination, and proficiency programs.
(a) Applicability. The requirements of this section apply to each
licensee of a self-reliant-mitigation facility that has not yet
certified the permanent cessation of operations and permanent removal
of fuel from the reactor vessel as described under Sec. 53.1070.
(b) Requirements. (1) The licensee must develop, implement, and
maintain training and examination programs that demonstrate compliance
with the requirements of paragraphs (b)(2) and (3) of this section.
(2) The training program must provide for both the initial and
continuing training of generally licensed reactor operators and be
derived from a systems approach to training as defined in this part.
(3)(i) The training program must incorporate the instructional
requirements necessary to provide qualified generally licensed reactor
operators to operate and maintain the facility in a safe manner in all
modes of operation. The training program must comply with the facility
license, including all technical specifications and applicable
regulations. The facility licensee must periodically evaluate and
revise the training program as appropriate to reflect industry
experience and relevant changes, including changes to the facility,
procedures, regulations, and quality assurance (QA) requirements.
Facility licensee management must periodically review the training
program for effectiveness.
(ii) The training program must ensure that generally licensed
reactor operators have and maintain the necessary knowledge, skills,
and abilities.
(iii) The training program must include the generally licensed
reactor operator manipulating the controls of either the facility or a
simulation facility that demonstrates compliance with the requirements
of Sec. 53.815(e).
(iv) The training program must include an initial examination
program for testing a representative sample of the knowledge, skills,
and abilities needed to safely perform generally licensed reactor
operator duties, to include both the examination methods and criteria
to be used to assess passing performance. The facility licensee must
provide the opportunity for a representative of the Commission to be
present during initial examination administration.
(v) The training program must include a requalification examination
program for testing a sample of the topics included under the systems
approach to training, to include the examination methods and criteria
to be used to assess passing performance. The requalification
examination program must specify an appropriate periodicity for
administering a complete requalification examination to each generally
licensed reactor operator, and the facility licensee must provide the
opportunity for a representative of the Commission to be present during
requalification examination administration.
(A) The facility licensee must ensure that any generally licensed
reactor operator who either demonstrates unsatisfactory performance on,
or fails to complete, the requalification examination is removed from
the performance of generally licensed reactor operator duties until
such time that any necessary remedial training has been completed and a
retake examination has been passed.
(B) [Reserved]
(vi) The training program must be approved by the Commission prior
to its use, as described under Sec. 53.730(g). The examination program
must provide for valid and reliable examinations and must be approved
by the Commission prior to their use, as described under Sec.
53.730(g). The approved programs are subject to the requirements of
Sec. 53.1565.
(c) Records. The following is required regarding the documentation
of the generally licensed reactor operator training and examination
programs:
(1) Sufficient records must be maintained by the facility licensee
to
[[Page 87069]]
maintain the integrity of the programs and kept available for NRC
inspection to verify the adequacy of the programs.
(2) The facility licensee must maintain records documenting the
participation of each generally licensed reactor operator in the
training and examination programs. The records must contain copies of
examinations administered, the answers given by the generally licensed
reactor operator, documentation of the grading of examinations, and
documentation of any additional training administered in areas in which
a generally licensed reactor operator exhibited deficiencies. The
facility licensee must retain these records while the associated
generally licensed reactor operators remain employed at the facility.
(3) Each record required by this part must be legible throughout
the retention period. The record may be the original, a reproduced
copy, or an electronic copy provided that the copy is authenticated by
authorized personnel.
(d) Examination integrity. Generally licensed reactor operators and
facility licensees must not engage in any activity that compromises the
integrity of any examination conducted under the generally licensed
reactor operator training and examination programs. The integrity of an
examination is considered compromised if any activity, regardless of
intent, affected, or, but for detection, could have affected the
equitable and consistent administration of the examination. This
includes all activities related to the preparation, administration, and
grading of examinations.
(e) Simulation facilities. (1) Simulation facilities used for
training purposes, for maintaining proficiency, or for the conduct of
examinations must demonstrate compliance with the following criteria as
they relate to the facility licensee's reference plant:
(i) The simulation facility must be of sufficient scope and
fidelity for individuals to acquire and demonstrate the necessary
knowledge, skills, and abilities to safely perform generally licensed
reactor operator duties.
(ii) The simulation facility must utilize models relating to
nuclear, thermal-hydraulic, and other applicable design-specific
characteristics that either replicate the most recent fuel load in the
reference commercial nuclear plant or, prior to initial fuel load (or,
for a fueled manufactured reactor, prior to initiating the physical
removal of any one of the independent physical mechanisms to prevent
criticality required under Sec. 53.620(d)(1)), replicate the intended
initial fuel load for the reference commercial nuclear plant, with the
exception of those portions of the simulation facility that utilize the
reference plant itself.
(iii) Simulator fidelity must be demonstrated so that significant
control manipulations are completed without procedural exceptions,
simulator performance exceptions, or deviation from the approved
training scenario sequence.
(2) Facility licensees that maintain a simulation facility for
training purposes, for maintaining proficiency, or for the conduct of
examinations must--
(i) Conduct performance testing throughout the life of the
simulation facility in a manner sufficient to ensure that paragraph
(e)(1) of this section is met;
(ii) Retain the results of performance testing for 4 years after
the completion of each performance test or until superseded by updated
test results;
(iii) Promptly correct modeling and hardware discrepancies and
discrepancies identified from scenario validation and from performance
testing or provide justification for why the presence of such
discrepancies will not adversely affect the criteria of paragraph
(e)(1) of this section;
(iv) Make the results of any uncorrected performance test failures
that may exist at the time of an inspection available for NRC review;
and
(v) Maintain the provisions for examination integrity consistent
with Sec. 53.815(d).
(f) Waiver of examination requirement. The facility licensee may
waive any or all of the requirements for an examination in accordance
with the facility licensee's Commission-approved generally licensed
reactor operator training and examination programs.
(g) Proficiency. The facility licensee must develop, implement, and
maintain a proficiency program to allow generally licensed reactor
operators to maintain proficiency regarding position functions and
familiarity with plant status. This program must include those steps
that will be taken in order to re-establish proficiency when it cannot
be maintained.
Sec. 53.820 Cessation of individual applicability.
The general license ceases to be applicable on an individual basis
once a generally licensed reactor operator is no longer being employed
in a position that may involve the manipulation of the controls of the
self-reliant mitigation facility.
Sec. 53.830 Training and qualification of commercial nuclear plant
personnel.
(a) This section addresses personnel training requirements. The
regulations within this section are applicable to all applicants for or
holders of OLs or COLs under this part.
(b) Prior to initial fuel load (or, for a fueled manufactured
reactor, prior to initiating the physical removal of any one of the
independent physical mechanisms to prevent criticality required under
Sec. 53.620(d)(1)), each holder of an operating or COL under this part
must, with sufficient time to provide trained and qualified personnel
to operate the facility, establish, implement, and maintain a training
program that demonstrates compliance with the requirements of
paragraphs (c) and (d) of this section.
(c) The training program must be derived from a systems approach to
training as defined in this part and must provide, at a minimum, for
the training and qualification of the following categories of
commercial nuclear plant personnel:
(1) Supervisors (e.g., shift supervisors);
(2) Technicians (e.g., maintenance, chemistry, and radiological);
and
(3) Other appropriate operating personnel (e.g., auxiliary
operators, certified fuel handlers, and individuals who provide
engineering expertise to on-shift operating personnel).
(d) The training program must incorporate the instructional
requirements necessary to provide qualified personnel to operate
components of a commercial nuclear plant and maintain the facility in a
safe manner in all modes of operation. The training program must be
developed to be in compliance with the facility license, including all
technical specifications and applicable regulations.
(1) The training program must be periodically evaluated and revised
as appropriate to reflect industry experience and relevant changes,
including changes to the facility, procedures, regulations, and QA
requirements. The training program must be periodically reviewed by
facility licensee management for effectiveness.
(2) Sufficient records must be maintained by the facility licensee
to maintain program integrity and kept available for NRC inspection to
verify the adequacy of the training program.
Sec. 53.845 Programs.
(a) The required plant programs under this part must include but
are not necessarily limited to the programs
[[Page 87070]]
described in the following sections of this subpart. Licensees may
combine, separate, and otherwise organize programs and related
documents as appropriate for the technologies and organizations
associated with the commercial nuclear plant.
(b) In addition to the programs described in the following
sections, programs must be provided for each commercial nuclear plant,
if necessary, to ensure that the performance of design features and
human actions are consistent with the analyses performed under
Sec. Sec. 53.450 and 53.730 and that the plant will demonstrate
compliance with the safety criteria defined in Sec. Sec. 53.210 and
53.220.
Sec. 53.850 Radiation protection.
(a) Each holder of an OL or COL under this part must develop,
implement, and maintain a Radiation Protection Program for operations
that is commensurate with the scope and extent of licensed activities
under this part and includes measures for limiting and monitoring
radioactive plant effluents and limiting and monitoring the dose to
individuals working with radioactive materials in accordance with 10
CFR part 20.
(b) Each holder of an OL or COL under this part must develop,
implement, and maintain a program for the control of radioactive
effluents and for keeping the doses to members of the public from
radioactive effluents as low as is reasonably achievable and for
environmental monitoring. The program must be contained in an Offsite
Dose Calculations Manual, must be implemented by procedures, and must
include remedial actions to be taken whenever the program limits are
exceeded. The Offsite Dose Calculations Manual must--
(1) Contain the methodology and parameters used in the calculation
of offsite doses resulting from radioactive gaseous and liquid
effluents, in the calculation of gaseous and liquid effluent monitoring
alarm and trip setpoints, and in the conduct of the radiological
environmental monitoring program; and
(2) Contain the radioactive effluent controls and radiological
environmental monitoring activities, and descriptions of the
information that should be included in the Annual Radiological
Environmental Operating and Radioactive Effluent Release Reports
required by Sec. 53.1645.
(c) Each holder of an OL or COL under this part must develop,
implement, and maintain a Process Control Program that identifies the
administrative and operational controls for solid radioactive waste
processing, process parameters, and surveillance requirements
sufficient to ensure compliance with the requirements of 10 CFR part
20, 10 CFR part 61, and 10 CFR part 71.
Sec. 53.855 Emergency preparedness.
(a) Each holder of an OL or COL under this part must have an
emergency response plan that must contain information needed to
demonstrate compliance with either the requirements in Sec. 50.160 of
this chapter or the requirements in appendix E to part 50 and the
planning standards of Sec. 50.47(b) of this chapter.
(b) No initial OL, initial COL, or early site permit that includes
complete and integrated emergency plans will be issued under this part
unless a finding is made by the NRC, in accordance with Sec. 50.47 of
this chapter, that there is reasonable assurance that adequate
protective measures can and will be taken in the event of a
radiological emergency.
Sec. 53.860 Security programs.
(a) Physical Protection Program. Each holder of an OL or COL under
this part must develop, implement, and maintain a physical protection
program under the following requirements:
(1) The licensee must implement security requirements for the
protection of special nuclear material based on the type, enrichment,
and quantity in accordance with 10 CFR part 73, as applicable, and
implement security requirements for the protection of Category 1 and
Category 2 quantities of radioactive material in accordance with 10 CFR
part 37, as applicable; and
(2) The licensee must demonstrate compliance with the provisions
set forth in either Sec. Sec. 73.55 or 73.100 of this chapter, unless
the licensee demonstrates compliance with the following criterion:
(i) The radiological consequences from a design-basis threat-
initiated event involving the loss of engineered systems for decay heat
removal and possible breaches in physical structures surrounding the
reactor, spent fuel, and other inventories of radioactive materials
result in offsite doses below the values in Sec. 53.210.
(ii) The applicant must perform a site-specific analysis, including
identification of target sets, to demonstrate that the criterion in
Sec. 53.860(a)(2)(i) is satisfied. The analysis must assume that
licensee mitigation and recovery actions, including any operator
actions, are unavailable or ineffective. The licensee must maintain the
analysis until the permanent cessation of operations and permanent
removal of fuel from the reactor vessel as described under Sec.
53.1070.
(b) Fitness for Duty. Each holder of an OL or COL under this part
must develop, implement, and maintain a fitness for duty program under
10 CFR part 26.
(c) Access Authorization. Each holder of an OL or COL under this
part must develop, implement, and maintain an access authorization
program under Sec. 73.120 of this chapter if the criterion in Sec.
53.860(a)(2)(i) is satisfied, or the requirements in Sec. 73.56 of
this chapter if the criterion is not satisfied.
(d) Cybersecurity. Each holder of an OL or COL under this part must
develop, implement, and maintain a cybersecurity program under
Sec. Sec. 73.54 or 73.110 of this chapter.
(e) Information Security. Each holder of an OL or COL under this
part must develop, implement, and maintain an information protection
system under Sec. Sec. 73.21, 73.22, and 73.23 of this chapter, as
applicable.
Sec. 53.865 Quality assurance.
Each holder of an OL or COL under this part must develop,
implement, and maintain a quality assurance program in accordance with
appendix B of part 50 of this chapter. A written quality assurance
program manual must be developed and used to guide the conduct of the
program in accordance with generally accepted consensus codes and
standards that have been endorsed or otherwise found acceptable by the
NRC.
Sec. 53.870 Integrity assessment programs.
Each holder of an OL or COL under this part must develop,
implement, and maintain an integrity assessment program to monitor,
evaluate, and manage--
(a) The effects of plant aging on SR and NSRSS SSCs. The program
may refer to surveillances, tests, and inspections conducted for
specific SSCs in accordance with other requirements in this part or
conducted in accordance with applicable consensus codes and standards
endorsed or otherwise found acceptable by the NRC;
(b) Cyclic or transient load limits to ensure that SR and NSRSS
SSCs are maintained within the applicable design limits; and
(c) Degradation mechanisms related to chemical interactions,
operating temperatures, effects of irradiation, and other environmental
factors to ensure that the capabilities, availability, and reliability
of SR and NSRSS SSCs demonstrate compliance with the
[[Page 87071]]
functional design criteria of Sec. Sec. 53.410 and 53.420.
Sec. 53.875 Fire protection.
(a)(1) Each holder of an OL or COL under this part must have a fire
protection plan that describes the overall fire protection program for
the facility; identifies the various positions within the licensee's
organization that are responsible for the program; states the
authorities that are delegated to each of these positions to implement
those responsibilities; and outlines the plans for fire protection,
fire detection and suppression capability; and limitation of fire
damage.
(2) The fire protection plan must also describe specific features
necessary to implement the program described in paragraph (a)(1) of
this section such as the following: administrative controls and
personnel requirements for fire prevention and manual fire suppression
activities; automatic and manually operated fire detection and
suppression systems; and the means to limit fire damage to SSCs so that
the capability to demonstrate compliance with the requirements of Sec.
53.210 is ensured.
(b)(1) Each holder of an OL or COL under this part must develop a
performance-based or deterministic fire protection program that
demonstrates compliance with the safety criteria outlined in Sec. Sec.
53.210 and 53.220, related safety functions outlined in Sec. 53.230,
and defense in depth as outlined in Sec. 53.250 with specific fire
protection measures related to fire prevention, fire detection, and
fire suppression.
(2) The fire protection program must comply with the following:
(i) Safety-related and NSRSS SSCs must be designed, located, and
maintained to minimize, consistent with other safety requirements, the
probability and effect of fires and explosions.
(ii) Noncombustible and fire-resistant materials must be used
wherever practical throughout the facility, particularly in locations
with SR and NSRSS SSCs.
(iii) Fire detection and fire suppression systems of appropriate
capacity and capability must be provided and designed and maintained to
minimize the adverse effects of fires on SR and NSRSS SSCs.
(iv) Fire suppression systems must be designed and maintained to
ensure that their rupture or inadvertent operation does not
significantly impair the ability of SR and NSRSS SSCs to perform their
safety functions to satisfy Sec. 53.230.
Sec. 53.880 Inservice inspection and inservice testing.
(a) Each holder of an OL or COL under this part must develop,
implement, and maintain a program for inservice inspection (ISI) and
inservice testing (IST) prior to receiving an OL or COL. The ISI/IST
programs must, wherever applicable, be in accordance with generally
accepted consensus codes and standards that have been endorsed or
otherwise found acceptable by the NRC. The ISI/IST program must include
all inspections and tests required by the codes and standards used in
the design and be supplemented by risk insights that identify the most
important SSCs to plant safety. The types of testing and inspections
and their frequency should be informed by risk insights to maintain the
reliability and performance of SSCs consistent with the associated
design and analyses activities involving those SSCs. Risk insights must
also be used to determine when to conduct the inspections and tests
(e.g., full power, shutdown, refueling) to minimize risk to the plant
workers and the public. The ISI/IST program must be documented in a
written manual and managed by qualified personnel reporting to the
Plant Manager.
(b) Prior to plant operation, baseline inspections and testing must
be performed using the same techniques as will be used for future
inspections and testing. The results of these inspections and testing
must be used as benchmarks for evaluating the results of future
inspections and testing. Sufficient room and support must be provided
to accommodate the personnel, ISI/IST equipment, and shielding
necessary to perform the inspections and testing. Acceptance criteria
for determining whether corrective action is needed must be developed
(or taken from the codes and standards used in the design) for
evaluating the results of the inspections and testing. The results of
the inspections and testing must be provided to the Plant Manager who
is responsible for determining what, if any, corrective action is
needed and when it should be taken. The ISI/IST results and corrective
actions must be documented and the documentation retained for the life
of the plant.
Sec. 53.910 Procedures and guidelines.
(a) Each holder of an OL or COL under this part must have a program
for developing, implementing, and maintaining an integrated set of
procedures, guidelines, and related supporting activities to support
normal operations and respond to possible unplanned events.
(b) The program required by paragraph (a) of this section must
include but is not limited to development, implementation, maintenance,
and supporting activities of procedures and guidelines for the
following:
(1) Plant operations;
(2) Maintenance activities under Sec. 53.715;
(3) Program requirements under this subpart F of this part;
(4) Emergency operating procedures, if developed to address the
role of human actions in responding to LBEs;
(5) Accident management guidelines, if developed to address the
role of human actions in responding to LBEs;
(6) Procedures for each area in which licensed special nuclear
material is handled, used, or stored to protect personnel upon the
sounding of a criticality alarm required by Sec. 53.440(m); and
(7) Procedures that describe how the licensee will address the
following areas if the licensee is notified of a potential aircraft
threat:
(i) Verification of the authenticity of threat notifications;
(ii) Maintenance of continuous communication with threat
notification sources;
(iii) Contacting all onsite personnel and applicable offsite
response organizations;
(iv) Onsite actions necessary to enhance the capability of the
facility to mitigate the consequences of an aircraft impact;
(v) Measures to reduce visual discrimination of the site relative
to its surroundings or individual buildings within the protected area;
(vi) Dispersal of equipment and personnel, as well as rapid entry
into site protected areas for essential onsite personnel and offsite
responders who are necessary to mitigate the event; and
(vii) Recall of site personnel.
Subpart G--Decommissioning Requirements
Sec. 53.1000 Scope and purpose.
This subpart defines the requirements related to decommissioning
for applicants for, or holders of, an operating license (OL) or
combined license (COL). The requirements related to maintaining
financial assurance for decommissioning are in Sec. Sec. 53.1010
through 53.1060. The requirements for transitioning from operations to
decommissioning and for the release of property and termination of the
license are in Sec. Sec. 53.1070 through 53.1080.
[[Page 87072]]
Sec. 53.1010 Financial assurance for decommissioning.
(a) This section establishes requirements for indicating to the
U.S. Nuclear Regulatory Commission (NRC) how an applicant for or holder
of an OL or COL under this part will provide reasonable assurance that
funds will be available for the decommissioning process. Reasonable
assurance consists of a series of steps as provided in paragraph (b) of
this section and Sec. Sec. 53.1020, 53.1030 and 53.1040. Funding for
the decommissioning of commercial nuclear plants may also be subject to
the regulation of Federal or State government agencies (e.g., Federal
Energy Regulatory Commission (FERC) and State Public Utility
Commissions) that have jurisdiction over rate regulation. The
requirements of this subpart, in particular Sec. 53.1020, are in
addition to, and not a substitution for, other requirements, and are
not intended to be used by themselves or by other agencies to establish
rates.
(b) Each applicant for an OL or COL under this part must prepare a
plan and an associated decommissioning report that ensures and
documents that adequate funding will be available to decommission the
facility. Each holder of an OL or COL must implement and maintain the
plan.
(1)(i) Before the Commission issues an OL under this part, the
applicant must update the decommissioning report to certify that it has
provided financial assurance for decommissioning in the amount proposed
in the application and approved by the NRC under Sec. 53.1020.
(ii) No later than 30 days after the Commission issues the notice
of intended operation under Sec. 53.1452 for a COL under this part,
the licensee must update the decommissioning report to certify that it
has provided financial assurance for decommissioning in the amount
proposed in the application and approved by the NRC under Sec.
53.1020.
(2) The amount of financial assurance for decommissioning to be
provided must be based on a site-specific cost estimate for
decommissioning the facility under Sec. 53.1020.
(3) The amount of financial assurance for decommissioning to be
provided must be adjusted annually using a rate at least equal to that
stated in Sec. 53.1030.
(4) The amount of financial assurance for decommissioning to be
provided must be covered by one or more of the methods described in
Sec. 53.1040 as acceptable to the NRC. A copy of the financial
instrument obtained to satisfy the requirements of Sec. 53.1040 must
be submitted to the NRC as part of the application for an OL under this
part; however, an applicant for or holder of a COL need not obtain such
financial instrument or submit a copy to the Commission except as
provided in Sec. 53.1060(b).
Sec. 53.1020 Cost estimates for decommissioning.
Cost estimates for decommissioning must be site-specific. Site-
specific decommissioning cost estimates (DCEs) must account for the
engineering, labor, equipment, transportation, disposal, and related
charges needed to support termination of the license. They must include
the costs for decontaminating structures, systems, and components and
the site environs; removal of contaminated components and materials
from the plant and the site environs; disposal of removed components
and materials in appropriate facilities; and any other activities
supporting the release of the property and termination of the license.
They must also address the approach to annual adjustments required by
Sec. 53.1030. Finally, site-specific DCEs must include plans for
adjusting levels of funds assured for decommissioning to demonstrate
that a reasonable level of assurance will be provided that funds will
be available when needed to cover the cost of decommissioning.
Sec. 53.1030 Annual adjustments to cost estimates for
decommissioning.
Each holder of an OL or COL under this part must annually adjust
the cost estimate for decommissioning to account for escalation in
labor, energy, and waste burial costs. Licensees may elect to use
either a site-specific adjustment factor, approved as part of the plan
and associated decommissioning report required by Sec. 53.1010, in
paragraph (a) of this section or the generic adjustment factor in
paragraph (b) of this section.
(a) A site-specific adjustment factor must address the estimated
contributions and escalation of costs for the following aspects of
decommissioning:
(1) Labor, materials, and services;
(2) Energy and waste transportation; and
(3) Radioactive waste burial or other disposition.
(b) A generic adjustment factor must be at least equal to 0.65 L +
0.13 E + 0.22 B, where L and E are escalation factors for labor and
energy, respectively, and are to be taken from regional data of U.S.
Department of Labor Bureau of Labor Statistics and B is an escalation
factor for waste burial and is to be taken from NRC report NUREG-1307,
``Report on Waste Burial Charges.''
Sec. 53.1040 Methods for providing financial assurance for
decommissioning.
Financial assurance for decommissioning is to be provided by the
following methods.
(a) Prepayment. Prepayment is the deposit made preceding the start
of operation or the transfer of a license under Sec. 53.1570 into an
account segregated from licensee assets and outside the administrative
control of the licensee and its subsidiaries or affiliates of cash or
liquid assets such that the amount of funds would be sufficient to pay
decommissioning costs. Prepayment may be in the form of a trust, escrow
account, or Government fund with payment by certificate of deposit,
deposit of government or other securities, or other method acceptable
to the NRC. This trust, escrow account, Government fund, or other type
of agreement must be established in writing and maintained at all times
in the United States with an entity that is an appropriate State or
Federal Government agency, or an entity whose operations in which the
prepayment deposit is managed are regulated and examined by a Federal
or State agency. A licensee that has prepaid funds based on a site-
specific cost estimate under Sec. 53.1020 may take credit for
projected earnings on the prepaid decommissioning trust funds, using up
to a 2 percent annual real rate of return through the time of
termination of the license. A licensee may use a credit of greater than
2 percent if the licensee's rate-setting authority has specifically
authorized a higher rate. Actual earnings on existing funds may be used
to calculate future fund needs.
(b) External sinking fund. An external sinking fund is a fund
established and maintained by setting funds aside periodically in an
account segregated from licensee assets and outside the administrative
control of the licensee and its subsidiaries or affiliates in which the
total amount of funds would be sufficient to pay decommissioning costs.
An external sinking fund may be in the form of a trust, escrow account,
or Government fund, with payment by certificate of deposit, deposit of
Government or other securities, or other method acceptable to the NRC.
This trust, escrow account, Government fund, or other type of agreement
must be established in writing and maintained at all times in the
United States with an entity that is an appropriate State or Federal
Government agency, or an entity whose operations in which the external
sinking fund is managed are regulated and examined by a Federal or
State agency. A licensee that has collected funds based on a site-
specific cost
[[Page 87073]]
estimate under Sec. 53.1020 may take credit for projected earnings on
the external sinking funds using up to a 2 percent annual real rate of
return from the time of future funds' collection through the time of
termination of the license. A licensee may use a credit of greater than
2 percent if the licensee's rate-setting authority has specifically
authorized a higher rate. Actual earnings on existing funds may be used
to calculate future fund needs. A licensee whose rates for
decommissioning costs cover only a portion of these costs may make use
of this method only for the portion of these costs that are collected
in one of the manners described in this paragraph. This method may be
used as the exclusive mechanism relied upon for providing financial
assurance for decommissioning in the following circumstances:
(1) By a licensee that recovers, either directly or indirectly, the
estimated total cost of decommissioning through rates established by
``cost of service'' or similar ratemaking regulation. Public utility
districts, municipalities, rural electric cooperatives, and State and
Federal agencies, including associations of any of the foregoing, that
establish their own rates and are able to recover their cost of service
allocable to decommissioning, are deemed to satisfy this condition.
(2) By a licensee whose source of revenues for its external sinking
fund is a ``non-bypassable charge,'' the total amount of which will
provide funds estimated to be needed for decommissioning pursuant to
Sec. Sec. 53.1020, 53.1060, or 53.1575.
(c) A surety method, insurance, or other guarantee method. (1)
These methods guarantee that decommissioning costs will be paid. A
surety method may be in the form of a surety bond, or letter of credit.
Any surety method or insurance used to provide financial assurance for
decommissioning must contain the following conditions:
(i) The surety method or insurance must be open-ended, or, if
written for a specified term, such as 5 years, must be renewed
automatically, unless 90 days or more prior to the renewal day the
issuer notifies the NRC, the beneficiary, and the licensee of its
intention not to renew. The surety or insurance must also provide that
the full-face amount be paid to the beneficiary automatically prior to
the expiration without proof of forfeiture if the licensee fails to
provide a replacement acceptable to the NRC within 30 days after
receipt of notification of cancellation.
(ii) The surety or insurance must be payable to a trust established
for decommissioning costs. The trustee and trust must be acceptable to
the NRC. An acceptable trustee includes an appropriate State or Federal
Government agency or an entity that has the authority to act as a
trustee and whose trust operations are regulated and examined by a
Federal or State agency.
(2) A parent company guarantee of funds for decommissioning costs
based on a financial test may be used if the guarantee and test are as
contained in appendix A to 10 CFR part 30.
(3) For commercial companies that issue bonds, a guarantee of funds
by the applicant or licensee for decommissioning costs based on a
financial test may be used if the guarantee and test are as contained
in appendix C to 10 CFR part 30. For commercial companies that do not
issue bonds, a guarantee of funds by the applicant or licensee for
decommissioning costs may be used if the guarantee and test are as
contained in appendix D to 10 CFR part 30. A guarantee by the applicant
or licensee may not be used in any situation in which the applicant or
licensee has a parent company holding majority control of voting stock
of the company.
(d) Funding method for Federal licensees. For a Federal licensee, a
statement of intent containing a cost estimate for decommissioning and
indicating that funds for decommissioning will be obtained when
necessary.
(e) Contractual funding method. Contractual obligation(s) on the
part of a licensee's customer(s), the total amount of which over the
duration of the contract(s) will provide the licensee's total share of
uncollected funds estimated to be needed for decommissioning pursuant
to Sec. Sec. 53.1020, 53.1060, or 53.1575. To be acceptable to the NRC
as a method of decommissioning funding assurance, the terms of the
contract(s) must include provisions that the buyer(s) of electricity or
other products will pay for the decommissioning obligations specified
in the contract(s), notwithstanding the operational status either of
the licensed plant to which the contract(s) pertains or force majeure
provisions. All proceeds from the contract(s) for decommissioning
funding will be deposited to the external sinking fund. The NRC
reserves the right to evaluate the terms of any contract(s) and the
financial qualifications of the contracting entity or entities offered
as assurance for decommissioning funding.
(f) Other funding mechanisms. Any other mechanism, or combination
of mechanisms, that provides, as determined by the NRC upon its
evaluation of the specific circumstances of each licensee submittal,
assurance of decommissioning funding equivalent to that provided by the
mechanisms specified in paragraphs (a) through (e) of this section.
Licensees who do not have sources of funding described in paragraph (b)
of this section may use an external sinking fund in combination with a
guarantee mechanism, as specified in paragraph (c) of this section,
provided that the total amount of funds estimated to be necessary for
decommissioning is assured.
Sec. 53.1045 Limitations on the use of decommissioning trust funds.
(a)(1) Decommissioning trust funds may be used by licensees if--
(i) The withdrawals are for expenses for decommissioning activities
consistent with the definition of decommission or decommissioning in
Sec. 53.020;
(ii) The expenditure would not reduce the value of the
decommissioning trust below an amount necessary to place and maintain
the reactor in a safe storage condition if unforeseen conditions or
expenses arise; and
(iii) The withdrawals would not inhibit the ability of the licensee
to complete funding of any shortfalls in the decommissioning trust
needed to ensure the availability of funds to ultimately release the
site and terminate the license.
(2) Initially, 3 percent of the amount determined in accordance
with Sec. 53.1020 may be used for decommissioning planning. For
licensees that have submitted the certifications required under Sec.
53.1070 and commencing 90 days after the NRC has received the post-
shutdown decommissioning activities report (PSDAR) required by Sec.
53.1060, an additional 20 percent may be used. An updated site-specific
DCE must be submitted to the NRC prior to the licensee using any
funding in excess of these amounts.
(b) Licensees that are not ``electric utilities'' as defined in
Sec. 53.020 that use prepayment or an external sinking fund to provide
financial assurance must provide in the terms of the arrangements
governing the trust, escrow account, or Government fund, used to
segregate and manage the funds that--
(1) The trustee, manager, investment advisor, or other person
directing investment of the funds--
(i) Is prohibited from investing the funds in securities or other
obligations of the licensee or any other owner or operator of any
commercial nuclear
[[Page 87074]]
plant or their affiliates, subsidiaries, successors or assigns, or in a
mutual fund in which at least 50 percent of the fund is invested in the
securities of a licensee or parent company whose subsidiary is an owner
or operator of a foreign or domestic commercial nuclear plant. However,
the funds may be invested in securities tied to market indices or other
non-nuclear sector collective, commingled, or mutual funds, provided
that no more than 10 percent of trust assets may be indirectly invested
in securities of any entity owning or operating one or more commercial
nuclear plants.
(ii) Is obligated at all times to adhere to a standard of care set
forth in the trust, which either shall be the standard of care, whether
in investing or otherwise, required by State or Federal law or one or
more State or Federal regulatory agencies with jurisdiction over the
trust funds, or, in the absence of any such standard of care, whether
in investing or otherwise, that a prudent investor would use in the
same circumstances. The term ``prudent investor,'' shall have the same
meaning as set forth in FERC's ``Regulations Governing Nuclear Plant
Decommissioning Trust Funds'' at 18 CFR 35.32(a)(3), or any successor
regulation.
(2) The licensee, its affiliates, and its subsidiaries are
prohibited from being engaged as investment manager for the funds or
from giving day-to-day management direction of the funds' investments
or direction on individual investments by the funds, except in the case
of passive fund management of trust funds where management is limited
to investments tracking market indices.
(3) The trust, escrow account, Government fund, or other account
used to segregate and manage the funds may not be amended in any
material respect without written notification to the Director, Office
of Nuclear Reactor Regulation, or Director, Office of Nuclear Material
Safety and Safeguards, as applicable, at least 30 working days before
the proposed effective date of the amendment. The licensee must provide
the text of the proposed amendment and a statement of the reason for
the proposed amendment. The trust, escrow account, Government fund, or
other account may not be amended if the person responsible for managing
the trust, escrow account, Government fund, or other account receives
written notice of objection from the Director, Office of Nuclear
Reactor Regulation, or Director, Office of Nuclear Material Safety and
Safeguards, as applicable, within the notice period.
(4) Except for withdrawals being made under paragraph (a) of this
section or for payments of ordinary administrative costs (including
taxes) and other incidental expenses of the fund (including legal,
accounting, actuarial, and trustee expenses) in connection with the
operation of the fund, no disbursement or payment may be made from the
trust, escrow account, Government fund, or other account used to
segregate and manage the funds until written notice of the intention to
make a disbursement or payment has been given to the Director, Office
of Nuclear Reactor Regulation, or Director, Office of Nuclear Material
Safety and Safeguards, as applicable, at least 30 working days before
the date of the intended disbursement or payment. The disbursement or
payment from the trust, escrow account, Government fund or other
account may be made following the 30 working day notice period if the
person responsible for managing the trust, escrow account, Government
fund, or other account does not receive written notice of objection
from the Director, Office of Nuclear Reactor Regulation, or Director,
Office of Nuclear Material Safety and Safeguards, as applicable, within
the notice period. Disbursements or payments from the trust, escrow
account, Government fund, or other account used to segregate and manage
the funds, other than for payment of ordinary administrative costs
(including taxes) and other incidental expenses of the fund (including
legal, accounting, actuarial, and trustee expenses) in connection with
the operation of the fund, are restricted to decommissioning expenses
or transfer to another financial assurance method acceptable under
Sec. 53.1040 until final decommissioning has been completed. After
decommissioning has begun and withdrawals from the decommissioning fund
are made under paragraph (a) of this section, no further notification
need be made to the NRC.
(c) Licensees that are ``electric utilities'' under Sec. 53.020
that use prepayment or an external sinking fund to provide financial
assurance must include a provision in the terms of the trust, escrow
account, Government fund, or other account used to segregate and manage
funds that except for withdrawals being made under paragraph (a) of
this section or for payments of ordinary administrative costs
(including taxes) and other incidental expenses of the fund (including
legal, accounting, actuarial, and trustee expenses) in connection with
the operation of the fund, no disbursement or payment may be made from
the trust, escrow account, Government fund, or other account used to
segregate and manage the funds until written notice of the intention to
make a disbursement or payment has been given the Director, Office of
Nuclear Reactor Regulation, or Director, Office of Nuclear Material
Safety and Safeguards, as applicable, at least 30 working days before
the date of the intended disbursement or payment. The disbursement or
payment from the trust, escrow account, Government fund or other
account may be made following the 30 working day notice period if the
person responsible for managing the trust, escrow account, Government
fund, or other account does not receive written notice of objection
from the Director, Office of Nuclear Reactor Regulation, or Director,
Office of Nuclear Material Safety and Safeguards, as applicable, within
the notice period. Disbursements or payments from the trust, escrow
account, Government fund, or other account used to segregate and manage
the funds, other than for payment of ordinary administrative costs
(including taxes) and other incidental expenses of the fund (including
legal, accounting, actuarial, and trustee expenses) in connection with
the operation of the fund, are restricted to decommissioning expenses
or transfer to another financial assurance method acceptable under
Sec. 53.1040 until final decommissioning has been completed. After
decommissioning has begun and withdrawals from the decommissioning fund
are made under paragraph (a) of this section, no further notification
need be made to the NRC.
(d) A licensee that is not an ``electric utility'' under Sec.
53.020 and using a surety method, insurance, or other guarantee method
to provide financial assurance must provide that the trust established
for decommissioning costs to which the surety or insurance is payable
contains in its terms the requirements in Sec. 53.1045(b)(1) through
(4).
Sec. 53.1050 NRC oversight.
The NRC reserves the right to take the following steps in order to
ensure a licensee's adequate accumulation of decommissioning funds:
review, as needed, the rate of accumulation of decommissioning funds
and, either independently or in cooperation with FERC and the
licensee's State Public Utility Commission, take additional actions as
appropriate on a case-by-case basis, including modification of a
licensee's schedule for the accumulation of decommissioning funds.
[[Page 87075]]
Sec. 53.1060 Reporting and recordkeeping requirements.
(a) Each holder of an OL under this part or holder of a COL under
this part after the date that the Commission has made the finding under
Sec. 53.1452(g) must report, at least once every 2 years, by March 31,
on the status of its certification of decommissioning funding for each
commercial nuclear reactor or part of a commercial nuclear reactor that
it owns. The information in this report must include, at a minimum, the
amount of decommissioning funds estimated to be required under
Sec. Sec. 53.1020 and 53.1030; the amount of decommissioning funds
accumulated to the end of the calendar year preceding the date of the
report; a schedule of the annual amounts remaining to be collected; the
assumptions used regarding rates of escalation in decommissioning
costs, rates of earnings on decommissioning funds, and rates of other
factors used in funding projections; any contracts upon which the
licensee is relying under Sec. 53.1040(e); any modifications occurring
to a licensee's method of providing financial assurance since the last
submitted report; and any material changes to trust agreements. If any
of the preceding items is not applicable, the licensee should so state
in its report. Any licensee for a plant that is within 5 years of the
projected end of its operation, or where conditions have changed such
that it will close within 5 years (before the end of its licensed
life), or that has already closed (before the end of its licensed
life), or that is involved in a merger or an acquisition must submit
this report annually.
(b) Each holder of a COL under this part must, 2 years before and 1
year before the scheduled date for initial loading of fuel (or, for a
fueled manufactured reactor, 2 years before and 1 year before the
scheduled date for initiating the physical removal of any one of the
independent physical mechanisms to prevent criticality required under
Sec. 53.620(d)(1)), submit a report to the NRC containing a
certification updating the DCEs and a copy of the financial instrument
to be used to satisfy Sec. 53.1040. No later than 30 days after the
Commission publishes notice in the Federal Register under Sec.
53.1452(a), the licensee must submit an updated decommissioning report
required under Sec. 53.1010(b)(1)(ii), including a copy of the
financial instrument obtained to satisfy Sec. 53.1040.
(c) Each licensee must keep records of information important to the
safe and effective decommissioning of the facility in an identified
location until the license is terminated by the Commission. If records
of relevant information are kept for other purposes, reference to these
records and their locations may be used. Information the Commission
considers important to decommissioning consists of--
(1) Records of spills or other unusual occurrences involving the
spread of contamination in and around the facility, equipment, or site.
These records may be limited to instances when significant
contamination remains after any cleanup procedures or when there is
reasonable likelihood that contaminants may have spread to inaccessible
areas as in the case of possible seepage into porous materials such as
concrete. These records must include any known information on
identification of involved nuclides, quantities, forms, and
concentrations.
(2) As-built drawings and modifications of structures and equipment
in restricted areas where radioactive materials are used and/or stored
and of locations of possible inaccessible contamination such as buried
pipes that may be subject to contamination. If required drawings are
referenced, each relevant document need not be indexed individually. If
drawings are not available, the licensee must substitute appropriate
records of available information concerning these areas and locations.
(3) Records of the cost estimate performed for the decommissioning
funding plan or of the amount certified for decommissioning, and
records of the funding method used for assuring funds if either a
funding plan or certification is used.
(4) Records of--
(i) The licensed site area, as originally licensed and any
revisions, which must include a site map and any acquisition or use of
property outside the originally licensed site area for the purpose of
receiving, possessing, or using licensed materials;
(ii) The licensed activities carried out on the acquired or used
property; and
(iii) The release and final disposition of any property recorded in
paragraph (c)(4)(i) of this section, the historical site assessment
performed for the release, radiation surveys performed to support
release of the property, submittals to the NRC made under Sec.
53.1070, and the methods employed to ensure that the property met the
radiological criteria of subpart E of 10 CFR part 20 at the time the
property was released.
(d) Each holder of an OL or COL under this part must at or about 5
years prior to the projected end of operations submit a preliminary DCE
which includes an up-to-date assessment of the major factors that could
affect the cost to decommission.
(e) Prior to or within 2 years following permanent cessation of
operations, the licensee must submit a PSDAR to the NRC, and a copy to
the affected State(s). The PSDAR must contain a description of the
planned decommissioning activities along with a schedule for their
accomplishment, a discussion that provides the reasons for concluding
that the environmental impacts associated with site-specific
decommissioning activities will be bounded by appropriate previously
issued environmental impact statements, and a site-specific DCE,
including the projected cost of managing irradiated fuel.
(f) For decommissioning activities that delay completion of
decommissioning by including a period of storage or surveillance, the
licensee must provide a means of adjusting cost estimates and
associated funding levels over the storage or surveillance period.
(g) After submitting its site-specific DCE required by paragraph
(e) of this section, and until the licensee has completed its final
radiation survey and demonstrated that residual radioactivity has been
reduced to a level that permits termination of its license, the
licensee must annually submit to the NRC, by March 31, a financial
assurance status report. The report must include the following
information, current through the end of the previous calendar year:
(1) The amount spent on decommissioning, both cumulative and over
the previous calendar year, the remaining balance of any
decommissioning funds, and the amount provided by other financial
assurance methods being relied upon;
(2) An estimate of the costs to complete decommissioning,
reflecting any difference between actual and estimated costs for work
performed during the year, and the decommissioning criteria upon which
the estimate is based;
(3) Any modifications occurring to a licensee's current method of
providing financial assurance since the last submitted report; and
(4) Any material changes to trust agreements or financial assurance
contracts.
(5) If the sum of the balance of any remaining decommissioning
funds, plus earnings on such funds calculated at not greater than a 2
percent real rate of return, together with the amount provided by other
financial assurance methods being relied upon, does not cover the
estimated cost to complete the decommissioning, the financial assurance
status report must include
[[Page 87076]]
additional financial assurance to cover the estimated cost of
completion.
(h) After submitting its site-specific DCE required by paragraph
(e) of this section, the licensee must annually submit to the NRC, by
March 31, a report on the status of its funding for managing irradiated
fuel. The report must include the following information, current
through the end of the previous calendar year:
(1) The amount of funds accumulated to cover the cost of managing
the irradiated fuel;
(2) The projected cost of managing irradiated fuel until title to
the fuel and possession of the fuel is transferred to the Secretary of
Energy; and
(3) If the funds accumulated do not cover the projected cost, a
plan to obtain additional funds to cover the cost.
Sec. 53.1070 Termination of license.
For each holder of an OL or COL under this part--
(a)(1) When the licensee has determined to permanently cease
operations the licensee must, within 30 days, submit a written
certification to the NRC, consistent with the requirements of Sec.
53.040(b)(8);
(2) When appropriate to support decommissioning activities and the
eventual permanent removal of fuel from the reactor vessel, the
licensee must develop defueled technical specifications by reviewing
the operational technical specifications and determining which
specifications no longer apply during decommissioning and which ones
should remain applicable. The licensee must make the appropriate
submittals to the NRC in accordance with Sec. 53.1510 to request
changes to the technical specifications; and
(3)(i) Once fuel has been permanently removed from the reactor
vessel, the licensee must submit a written certification to the NRC
that meets the requirements of Sec. 53.040(b)(9); and
(ii) The licensee must establish and maintain staffing consisting
of certified fuel handlers, as defined under Sec. 53.020, and other
non-licensed personnel with appropriate qualifications, and in
sufficient numbers, to ensure support for facility operations and
radiological control activities, as required by the facility defueled
technical specifications. These personnel must be subject to the
training requirements of Sec. 53.830.
(b) Upon docketing of the certifications for permanent cessation of
operations and permanent removal of fuel from the reactor vessel, or
when a final legally effective order to permanently cease operations
has come into effect, the license issued under this part no longer
authorizes operation of the reactor or emplacement or retention of fuel
into the reactor vessel.
(c) Decommissioning will be completed within 60 years of permanent
cessation of operations. Completion of decommissioning beyond 60 years
will be approved by the Commission only when necessary to protect
public health and safety. Factors that will be considered by the
Commission in evaluating an alternative that provides for completion of
decommissioning beyond 60 years of permanent cessation of operations
include unavailability of waste disposal capacity and other site-
specific factors affecting the licensee's capability to carry out
decommissioning, including presence of other nuclear facilities at the
site.
(d)(1) Prior to or within 2 years following permanent cessation of
operations, the licensee must submit a PSDAR and site-specific DCE in
accordance with Sec. 53.1060(e).
(2) The NRC must notice receipt of the PSDAR and make the PSDAR
publicly available and publish notice of its availability for public
comment in the Federal Register. The NRC must also schedule a public
meeting readily accessible to individuals in the vicinity of the
licensee's facility. The NRC must publish a notice in the Federal
Register and in a forum, such as local newspapers, that is readily
accessible to individuals in the vicinity of the site, announcing the
date, time, and location of the meeting, along with a brief description
of the purpose of the meeting.
(e) Licensees must not perform any major decommissioning
activities, as defined in Sec. 53.020, until 90 days after the NRC has
received the licensee's PSDAR submittal and until certifications of
permanent cessation of operations and permanent removal of fuel from
the reactor vessel, as required under paragraph (a) of this section,
have been submitted.
(f) Licensees must not perform any decommissioning activities, as
defined in Sec. 53.020, that--
(1) Foreclose release of the site for possible unrestricted use;
(2) Result in significant environmental impacts not previously
reviewed; or
(3) Result in there no longer being reasonable assurance that
adequate funds will be available for decommissioning.
(g) In taking actions permitted under Sec. 53.1540 following
submittal of the PSDAR, the licensee must notify the NRC in writing,
and send a copy to the affected State(s), before performing any
decommissioning activity inconsistent with, or making any significant
schedule change from, those actions and schedules described in the
PSDAR, including changes that increase the decommissioning cost by more
than 20 percent from the previously provided DCE.
(h) Licensees may use decommissioning trust funds consistent with
the limitations of Sec. 53.1045(a). Licensees must report on the
status of decommissioning trust funds consistent with the requirements
of Sec. 53.1060.
(i) Licensees must submit an application for termination of license
in accordance with Sec. 53.1070. The application for termination of
license must be accompanied or preceded by a license termination plan
to be submitted for NRC approval.
(1) The license termination plan must be a supplement to the Final
Safety Analysis Report or equivalent and must be submitted at least 2
years before termination of the license date.
(2) The license termination plan must include--
(i) A site characterization;
(ii) Identification of remaining dismantlement activities;
(iii) Plans for site remediation;
(iv) Detailed plans for the final radiation survey;
(v) A description of the end use of the site, if restricted;
(vi) An updated site-specific estimate of remaining decommissioning
costs;
(vii) A supplement to the environmental report, pursuant to Sec.
51.53 of this chapter, describing any new information or significant
environmental change associated with the licensee's proposed
termination activities; and
(viii) Identification of parts, if any, of the facility or site
that were released for use before approval of the license termination
plan.
(3) Following receipt of the license termination plan, the NRC must
make the license termination plan publicly available and publish notice
of its availability for public comment in the Federal Register. The NRC
must also schedule a public meeting readily accessible to individuals
in the vicinity of the licensee's facility upon receipt of the license
termination plan. The NRC must publish a notice in the Federal Register
and in a forum, such as local newspapers, that is readily accessible to
individuals in the vicinity of the site, announcing the date, time, and
location of the meeting, along with a brief description of the purpose
of the meeting.
(j) If the license termination plan demonstrates that the remainder
of
[[Page 87077]]
decommissioning activities will be performed in accordance with the
regulations in this chapter, will not be inimical to the common defense
and security or to the health and safety of the public, and will not
have a significant effect on the quality of the environment and after
notice to interested persons, the Commission will approve the plan, by
license amendment, subject to such conditions and limitations as it
deems appropriate and necessary and authorize implementation of the
license termination plan.
(k) The Commission will terminate the license if it determines
that--
(1) The remaining dismantlement has been performed in accordance
with the approved license termination plan, and
(2) The final radiation survey and associated documentation,
including an assessment of dose contributions associated with parts
released for use before approval of the license termination plan,
demonstrate that the facility and site have met the criteria for
decommissioning in subpart E of 10 CFR part 20.
Sec. 53.1075 Program requirements during decommissioning.
(a) Licensees that have submitted the certifications required under
Sec. 53.1070 must maintain a decommissioning fire protection program
to address the potential for fires that could cause the release or
spread of radioactive materials.
(1) The objectives of the decommissioning fire protection program
are to
(i) Reasonably prevent these fires from occurring;
(ii) Rapidly detect, control, and extinguish those fires that do
occur and that could result in a radiological hazard; and
(iii) Ensure that the risk of fire-induced radiological hazards to
the public, environment, and plant personnel is minimized.
(2) The licensee must assess the decommissioning fire protection
program on a regular basis. The licensee must revise the
decommissioning fire protection program documentation as appropriate
throughout the various stages of facility decommissioning.
(3) The licensee may make changes to the decommissioning fire
protection program without NRC approval if these changes do not reduce
the effectiveness of fire protection for structures, systems, and
components that could result in a radiological hazard, taking into
account the decommissioning plant conditions and activities.
(b) [Reserved]
Sec. 53.1080 Release of part of a commercial nuclear plant or site
for unrestricted use.
(a) Prior written NRC approval is required to release part of a
commercial nuclear plant or site for unrestricted use at any time
before receiving approval of a license termination plan. Section
53.1060 specifies recordkeeping requirements associated with partial
release. Holders of an OL or COL under this part seeking NRC review and
approval must--
(1) Evaluate the effect of releasing the property to ensure that--
(i) The dose to individual members of the public does not exceed
the limits and standards of subpart D of 10 CFR part 20;
(ii) There is no reduction in the effectiveness of emergency
planning or physical security;
(iii) Effluent releases remain within license conditions;
(iv) The environmental monitoring program and offsite dose
calculation manual are revised to account for the changes;
(v) The siting criteria of subpart D of this part continue to be
met; and
(vi) All other applicable statutory and regulatory requirements
continue to be met.
(2) Perform a historical site assessment of the part of the
commercial nuclear plant or site to be released; and
(3) Perform surveys adequate to demonstrate compliance with the
radiological criteria for unrestricted use specified in Sec. 20.1402
of this chapter for impacted areas.
(b) For release of non-impacted areas, the licensee may submit a
written request for NRC review and approval of the release if a license
amendment is not otherwise required. The request submittal must
include--
(1) The results of the evaluations performed in accordance with
paragraphs (a)(1) and (a)(2) of this section;
(2) A description of the part of the commercial nuclear plant or
site to be released;
(3) The schedule for release of the property;
(4) The results of the evaluations performed in accordance with
Sec. 53.1540; and
(5) A discussion that provides the reasons for concluding that the
environmental impacts associated with the licensee's proposed release
of the property will be bounded by appropriate previously issued
environmental impact statements.
(c) After receiving a request from the licensee for NRC approval of
the release of a non-impacted area, the NRC must--
(1) Determine whether the licensee has adequately evaluated the
effect of releasing the property as required by paragraph (a)(1) of
this section;
(2) Determine whether the licensee's classification of any release
areas as non- impacted is adequately justified; and
(3) If determining that the licensee's submittal is adequate,
inform the licensee in writing that the release is approved.
(d) For release of impacted areas, the licensee must submit an
application for amendment of its license for the release of the
property. The application must include--
(1) The information specified in paragraphs (b)(1) through (b)(3)
of this section;
(2) The methods used for and results obtained from the radiation
surveys required to demonstrate compliance with the radiological
criteria for unrestricted use specified in Sec. 20.1402; and
(3) A supplement to the environmental report, under Sec. 51.53 of
this chapter, describing any new information or significant
environmental change associated with the licensee's proposed release of
the property.
(e) After receiving a license amendment application from the
licensee for the release of an impacted area, the NRC must--
(1) Determine whether the licensee has adequately evaluated the
effect of releasing the property as required by paragraph (a)(1) of
this section;
(2) Determine whether the licensee's classification of any release
areas as non-impacted is adequately justified;
(3) Determine whether the licensee's radiation survey for an
impacted area is adequate; and
(4) If determining that the licensee's submittal is adequate,
approve the licensee's amendment application.
(f) The NRC must publish notice receipt of the release approval
request or license amendment application in the Federal Register and
make the approval request or license amendment application available
for public comment. Before acting on an approval request or license
amendment application submitted in accordance with this section, the
NRC must conduct a public meeting readily accessible to individuals in
the vicinity of the licensee's facility for the purpose of obtaining
public comments on the proposed release of part of the commercial
nuclear plant or site. The NRC must publish a document in the Federal
Register and in a forum, such as local newspapers, which is readily
accessible to individuals in the vicinity of the site, announcing the
date, time,
[[Page 87078]]
and location of the meeting, along with a brief description of the
purpose of the meeting.
Subpart H--Licenses, Certifications, and Approvals
Sec. 53.1100 Filing of application for licenses, certifications, or
approvals; oath or affirmation.
(a) Serving of applications. (1) Each filing of an application for
a standard design approval, standard design certification, or license
under this part, and any amendments to the applications, must be
submitted to the U.S. Nuclear Regulatory Commission (NRC) under Sec.
53.040, as applicable.
(2) Each applicant for a construction permit (CP), early site
permit, combined license (COL), or manufacturing license (ML) under
this part must, upon notification by the presiding officer designated
to conduct the public hearing required by the Atomic Energy Act of
1954, as amended, (the Act) update the application and serve the
updated copies of the application or parts of it, eliminating all
superseded information, together with an index of the updated
application, as directed by presiding officer. Any subsequent amendment
to the application must be served on those served copies of the
application and must be submitted to the NRC as specified in Sec.
53.040, as applicable.
(3) The applicant must make a copy of the updated application
available at the public hearing for the use of any other parties to the
proceeding and must certify that the updated copies of the application
contain the current contents of the application submitted in accordance
with the requirements under this part.
(4) At the time of filing an application, the Commission will make
available at the NRC website, https://www.nrc.gov, a copy of the
application, subsequent amendments, and other records pertinent to the
matter that is the subject of the application for public inspection and
copying.
(5) The serving of copies required by this section must not occur
until the application has been docketed under Sec. 2.101(a) of this
chapter. Copies must be submitted to the Commission, as specified in
Sec. 53.040, as applicable, to enable the Director, Office of Nuclear
Reactor Regulation to determine whether the application is sufficiently
complete to permit docketing.
(b) Oath or affirmation. Each application for a standard design
approval, standard design certification, or license, including,
whenever appropriate, a CP or early site permit, or amendment of it,
and each amendment of each application must be executed in a signed
original by the applicant or duly authorized officer thereof under oath
or affirmation.
(c) [Reserved]
(d) [Reserved]
(e) Filing fees. Each application for a standard design approval,
standard design certification, or commercial nuclear plant license
under this part, including, whenever appropriate, a CP, COL, operating
license (OL), ML, or early site permit, other than a license exempted
from 10 CFR part 170, must be accompanied by the fee prescribed in 10
CFR part 170. No fee will be required to accompany an application for
renewal, amendment, or termination of a CP, OL, COL, or ML, except as
provided in Sec. 170.21 of this chapter.
(f) Environmental report. An application for a CP, OL, early site
permit, design certification, COL, or ML for a commercial nuclear plant
must be accompanied by an environmental report required under subpart A
of 10 CFR part 51.
Sec. 53.1101 Requirement for license.
Except as provided in Sec. 53.1120, no person within the United
States may transfer or receive in interstate commerce, manufacture,
produce, transfer, acquire, possess, or use any utilization facility
except as authorized by a license issued by the Commission.
Sec. 53.1103 Combining applications and licenses.
(a) An applicant may combine several applications in one
application for different kinds of licenses under the regulations in
this chapter.
(b) The Commission may combine in a single license the activities
of an applicant which would otherwise be licensed separately.
Sec. 53.1106 Elimination of repetition.
An applicant may incorporate by reference in its application
information contained in previous applications, statements, or reports
filed with the Commission, provided, however, that such references are
clear and specific.
Sec. 53.1109 Contents of applications; general information.
Each application must include, unless otherwise indicated in this
subpart--
(a) Name of applicant;
(b) Address of applicant;
(c) Description of business or occupation of applicant;
(d)(1) If applicant is an individual, the citizenship of applicant;
(2) If applicant is a partnership, the name, citizenship and
address of each partner and the principal location where the
partnership does business;
(3) If applicant is a corporation or an unincorporated association,
the following information:
(i) The State where it is incorporated or organized and the
principal location where it does business;
(ii) The names, addresses and citizenship of its directors and of
its principal officers; and
(iii) Whether it is owned, controlled, or dominated by an alien, a
foreign corporation, or foreign government, and if so, give details; or
(4) If the applicant is acting as agent or representative of
another person in filing the application, identify the principal and
furnish information required under this paragraph with respect to such
principal;
(e) The class and type of license applied for, the use to which the
facility will be put, the period of time for which the license is
sought, and a list of other licenses, except operator's licenses,
issued or applied for in connection with the proposed facility;
(f) [Reserved]
(g)(1) Except as provided in paragraph (g)(2) of this section, if
the application is for an OL or COL for a commercial nuclear plant, or
if the application is for an early site permit for a commercial nuclear
plant and contains plans for coping with emergencies under Sec.
53.1146(b)(2)(ii), the applicant must submit the radiological emergency
response plans of State, local, and participating Tribal governmental
entities in the United States that are wholly or partially within the
plume exposure pathway emergency planning zone (EPZ),\1\ and the plans
of State governments wholly or partially within the ingestion pathway
EPZ.\2\ If the application is for an early site permit that, under
Sec. 53.1146(b)(2)(i), proposes major features of the emergency plans
describing the EPZs, then the descriptions of the EPZs must meet the
requirements of this paragraph. Generally, the plume exposure pathway
EPZ for a commercial nuclear plant must consist of an area about 10
miles (16 km) in radius and the ingestion pathway EPZ must consist of
an area about 50 miles (80 km) in radius. The exact size and
configuration of the EPZs surrounding a particular commercial nuclear
plant must be determined in relation to the local emergency response
needs and capabilities as they are affected by such conditions as
demography, topography, land characteristics, access routes, and
jurisdictional boundaries. The size of the EPZs also may be determined
on a case-by-case basis for gas-cooled reactors and for reactors with
an authorized power level less than 250
[[Page 87079]]
megawatt thermal. The plans for the ingestion pathway must focus on
such actions as are appropriate to protect the food ingestion pathway.
(2) Applicants for commercial nuclear plants consisting of either
small modular reactors or non-light-water reactors complying with Sec.
50.160 of this chapter who apply for a CP, an OL, a COL, or an early
site permit under this part must submit as part of the application the
analysis used to determine whether the criteria in Sec.
53.1109(g)(2)(i)(A) and (B) are met and, if they are met, the size of
the plume exposure pathway EPZ.
(i) The plume exposure pathway EPZ is the area within which:
(A) Public dose, as defined in Sec. 20.1003 of this chapter, is
projected to exceed 10 millisieverts (1 rem) total effective dose
equivalent over 96 hours from the release of radioactive materials from
the facility considering accident likelihood and source term, timing of
the accident sequence, and meteorology; and
(B) Pre-determined, prompt protective measures are necessary.
(ii) If the application is for an OL or COL or if the application
is for an early site permit and contains plans for coping with
emergencies under Sec. 53.1146(b)(2)(ii), and if the plume exposure
pathway EPZ extends beyond the site boundary:
(A) The applicant must submit radiological emergency response plans
of State, local, and participating Tribal governmental entities in the
United States that are wholly or partially within the plume exposure
pathway EPZ.
(B) The exact configuration of the plume exposure pathway EPZ
surrounding the facility shall be determined in relation to the local
emergency response needs and capabilities as they are affected by such
conditions as demography, topography, land characteristics, access
routes, and jurisdictional boundaries.
(iii) If the application is for an early site permit that, under
Sec. 53.1146(b)(2)(i), proposes major features of the emergency plans
and describes the EPZ, and if the EPZ extends beyond the site boundary,
then the exact configuration of the plume exposure pathway EPZ
surrounding the facility must be determined in relation to the local
emergency response needs and capabilities as they are affected by such
conditions as demography, topography, land characteristics, access
routes, and jurisdictional boundaries.
(h) [Reserved]
(i) A list of the names and addresses of such regulatory agencies
as may have jurisdiction over the rates and services incident to the
proposed activity, and a list of trade and news publications which
circulate in the area where the proposed activity will be conducted and
which are considered appropriate to give reasonable notice of the
application to those municipalities, private utilities, public bodies,
and cooperatives, which might have a potential interest in the
facility; and
(j) If the application contains Restricted Data or classified
National Security information, confirmation that all Restricted Data
and classified National Security information are separated from the
unclassified information.
\1\ EPZs are discussed in NUREG-0396, U.S. Environmental
Protection Agency 520/1-78-016, ``Planning Basis for the Development
of State and Local Government Radiological Emergency Response Plans
in Support of Light-Water Nuclear Power Plants,'' December 1978.
\2\ If the State, local, and participating Tribal emergency
response plans have been previously provided to the NRC for
inclusion in the facility docket, the applicant need only provide
the appropriate reference to meet this requirement.
Sec. 53.1112 Environmental conditions.
(a) Each CP, early site permit, and COL under this part may include
conditions to address environmental issues during construction. These
conditions are to be set out in an attachment to the license, which is
incorporated in and made a part of the license. These conditions will
be derived from information contained in the environmental report
submitted pursuant to Sec. 51.50 of this chapter, as analyzed and
evaluated in the NRC record of decision and will identify the
obligations of the licensee in the environmental area, including, as
appropriate, requirements for reporting and keeping records of
environmental data, and any conditions and monitoring requirement for
the protection of the nonaquatic environment.
(b) Each license authorizing operation of a commercial nuclear
plant under this part, and each license for a commercial nuclear plant
for which the certification of permanent cessation of operations
required under Sec. 53.1070 has been submitted may include conditions
to address environmental issues during operation and decommissioning.
These conditions are to be set out in an attachment to the license,
which is incorporated in and made a part of the license. These
conditions will be derived from information contained in the
environmental report or the supplement to the environmental report
submitted under Sec. Sec. 51.50 and 51.53 of this chapter as analyzed
and evaluated in the NRC record of decision, and will identify the
obligations of the licensee in the environmental area, including, as
appropriate, requirements for reporting and keeping records of
environmental data and any conditions and monitoring requirement for
the protection of the nonaquatic environment.
Sec. 53.1115 Agreement limiting access to classified information.
As part of its application and in any event before the receipt of
Restricted Data or classified National Security Information or the
issuance of a license or standard design approval under this part, or
before the Commission has adopted a final standard design certification
rule under this part, the applicant must agree in writing that it will
not permit any individual to have access to or any facility to possess
Restricted Data or classified National Security Information until the
individual and/or facility has been approved for access under the
provisions of 10 CFR parts 25 and/or 95. The agreement of the applicant
becomes part of the license or standard design approval.
Sec. 53.1118 Ineligibility of certain applicants.
Any person who is a citizen, national, or agent of a foreign
country, or any corporation, or other entity which the Commission knows
or has reason to believe is owned, controlled, or dominated by an
alien, a foreign corporation, or a foreign government, will be
ineligible to apply for and obtain a license.
Sec. 53.1120 Exceptions and exemptions from licensing requirements.
Nothing in this part must be deemed to require a license for--
(a) The manufacture, production, or acquisition by the Department
of Defense of any utilization facility authorized pursuant to section
91 of the Act or the use of such facility by the Department of Defense
or by a person under contract with and for the account of the
Department of Defense;
(b) Except to the extent that the Department of Energy facilities
of the types subject to licensing pursuant to section 202 of the Energy
Reorganization Act of 1974, as amended, are involved--
(1)(i) The processing, fabrication or refining of special nuclear
material (SNM) or the separation of SNM, or the separation of SNM from
other substances by a prime contractor of the Department of Energy
under a prime contract for--
[[Page 87080]]
(A) The performance of work for the Department of Energy at a
United States government-owned or controlled site;
(B) Research in, or development, manufacture, storage, testing or
transportation of, atomic weapons or components thereof; or
(C) The use or operation of a utilization facility in a United
States owned vehicle or vessel; or
(ii) The processing, fabrication or refining of SNM of the
separation of SNM, or the separation of SNM from other substances by a
prime contractor or subcontractor of the Commission or the Department
of Energy under a prime contract or subcontract when the Commission
determines that the exemption of the prime contractor or subcontractor
is authorized by law; and that, under the terms of the contract or
subcontract, there is adequate assurance that the work thereunder can
be accomplished without undue risk to the public health and safety; or
(2)(i) The construction or operation of a utilization facility for
the Department of Energy at a United States government-owned or
controlled site, including the transportation of the utilization
facility to or from such site and the performance of contract services
during temporary interruptions of such transportation; or the
construction or operation of a utilization facility for the Department
of Energy in the performance of research in, or development,
manufacture, storage, testing, or transportation of, atomic weapons or
components thereof; or the use or operation of a utilization facility
for the Department of Energy in a United States government-owned
vehicle or vessel; provided that such activities are conducted by a
prime contractor of the Department of Energy under a prime contract
with the Department of Energy; or
(ii) The construction or operation of a utilization facility by a
prime contractor or subcontractor of the Commission or the Department
of Energy under his or her prime contract or subcontract when the
Commission determines that the exemption of the prime contractor or
subcontractor is authorized by law; and that, under the terms of the
contract or subcontract, there is adequate assurance that the work
thereunder can be accomplished without undue risk to the public health
and safety; or
(c) The transportation or possession of any utilization facility by
a common or contract carrier or warehouse employee in the regular
course of carriage for another or storage incident thereto.
Sec. 53.1121 Public inspection of applications.
Applications and documents submitted to the Commission in
connection with applications may be made available for public
inspection under the provisions of part 2 of this chapter.
Sec. 53.1124 Relationship between sections.
(a) Limited work authorization. An application for a limited work
authorization (LWA) under this part may be submitted as part of an
application for an early site permit, CP, or COL under this part as
required in Sec. 53.1130(a)(2).
(b) Early site permit. (1) A holder of an early site permit may
request an LWA.
(2) An application for a CP or COL under this part may, but need
not, reference an early site permit.
(c) Standard design approval. An application for a standard design
approval under this part may, but need not, reference an OL or custom
COL under this part that is essentially the same as the information
supporting the standard design for which approval is being requested.
(d) Standard design certification. An application for a standard
design certification under this part may, but need not, reference an OL
or custom COL under this part that is essentially the same as the
standard design for which certification is being requested.
(e) Manufacturing license. (1) A manufactured reactor manufactured
under an ML issued under this part may only be transported to and
installed at a site for which a COL under this part has been issued.
(2) An ML applicant under this part may reference a standard design
certification or a standard design approval under this part in its
application.
(f) Construction permit. An application for a CP may, but need not,
reference a standard design certification or standard design approval
issued under this part, respectively, and may also reference an early
site permit issued under this part. In the absence of a demonstration
that an entity other than the one originally sponsoring a standard
design certification is qualified to supply a design, the Commission
will entertain an application for a CP that references a standard
design certification issued under this part only if the entity that
sponsored the certification supplies the design for the applicant's
use.
(g) Operating license. (1) An application for an OL under this part
may, but need not, reference an early site permit, standard design
certification, or standard design approval issued under this part. In
the absence of a demonstration that an entity other than the one
originally sponsoring a standard design certification is qualified to
supply a design, the Commission will entertain an application for an OL
that references a standard design certification issued under this part
only if the entity that sponsored the certification supplies the design
for the applicant's use.
(2) The holder of a CP must, at the time of submission of the Final
Safety Analysis Report (FSAR), file an application for an OL.
(h) Combined licenses. An application for a COL under this part
may, but need not, reference an early site permit, standard design
certification, standard design approval, or ML issued under this part.
In the absence of a demonstration that an entity other than the one
originally sponsoring and obtaining a standard design certification is
qualified to supply a design, the Commission will entertain an
application for a COL that references a standard design certification
issued under this part only if the entity that sponsored the
certification supplies the design for the applicant's use.
Sec. 53.1130 Limited work authorizations.
(a) Request for limited work authorization. (1) Any person to whom
the Commission may otherwise issue either a license or permit related
to a commercial nuclear plant may request an LWA allowing that person
to perform the driving of piles, subsurface preparation, placement of
backfill, concrete, or permanent retaining walls within an excavation,
and installation of the foundation, including placement of concrete,
any of which are for a structure, system, or component (SSC) of the
facility for which either a CP or COL is otherwise required under Sec.
53.610.
(2) An application for an LWA may be submitted as part of a
complete application for a CP or COL in accordance with Sec.
2.101(a)(1) through (a)(5) of this chapter, or as a partial application
in accordance with Sec. 2.101(a)(9) of this chapter. An application
for an LWA by the holder of an early site permit must be submitted as a
complete application in accordance with Sec. 2.101(a)(1) through
(a)(4) of this chapter.
(3) The application must include--
(i) A Safety Analysis Report required by Sec. Sec. 53.1146,
53.1309 or 53.1416, as applicable, a description of the activities
requested to be performed, and the design and construction information
[[Page 87081]]
otherwise required by the Commission's rules and regulations to be
submitted for a CP or COL under this part but limited to those portions
of the facility that are within the scope of the LWA. The Safety
Analysis Report must demonstrate that activities conducted under the
LWA will be conducted in compliance with the technically relevant
Commission requirements in 10 CFR chapter I applicable to the design of
those portions of the facility within the scope of the LWA;
(ii) An environmental report in accordance with Sec. 51.49 of this
chapter; and
(iii) A plan for redress of activities performed under the LWA,
should limited work activities be terminated by the holder, or the LWA
be revoked by the NRC or upon effectiveness of the Commission's final
decision denying the associated CP or COL application, as applicable.
(b) Issuance of limited work authorization. (1) The Director,
Office of Nuclear Reactor Regulation may issue an LWA only after--
(i) The NRC staff issues the final environmental impact statement
for the LWA under subpart A of part 51 of this chapter;
(ii) The presiding officer makes the finding in Sec. Sec.
51.105(c) or 51.107(d) of this chapter, as applicable;
(iii) The Director determines that the applicable standards and
requirements of the Act, and the Commission's regulations applicable to
the activities to be conducted under the LWA, have been met, the
applicant is technically qualified to engage in the activities
authorized, and that issuance of the LWA will provide reasonable
assurance of adequate protection to public health and safety and will
not be inimical to the common defense and security; and
(iv) The presiding officer finds that there are no unresolved
safety issues relating to the activities to be conducted under the LWA
that would constitute good cause for withholding the authorization.
(2) Each LWA will specify the activities that the holder is
authorized to perform.
(c) Effect of limited work authorization. Any activities undertaken
under an LWA are entirely at the risk of the applicant and, except as
to the matters determined under paragraph (b)(1) of this section, the
issuance of the LWA has no bearing on the issuance of a CP or COL with
respect to the requirements of the Act and rules, regulations, or
orders issued under the Act. The environmental impact statement for a
CP or COL application for which an LWA was previously issued will not
address, and the presiding officer will not consider, the sunk costs of
the holder of the LWA in determining the proposed action (i.e.,
issuance of the CP or COL).
(d) Implementation of redress plan. If construction is terminated
by the holder, the underlying application is withdrawn by the applicant
or denied by the NRC, or the LWA is revoked by the NRC, then the holder
must begin implementation of the redress plan in a reasonable time. The
holder must also complete the redress of the site no later than 18
months after termination of construction, revocation of the LWA, or
upon effectiveness of the Commission's final decision denying the
associated CP application or the associated COL application, as
applicable.
Sec. 53.1140 Early site permits.
Sections 53.1140 through 53.1188 set out the requirements and
procedures applicable to Commission issuance of an early site permit
under this part for approval of a site for a commercial nuclear plant
separate from the filing of an application for a CP or COL for the
facility.
Sec. 53.1143 Filing of applications.
Any person who may apply for a CP or for a COL under this part, may
file an application for an early site permit with the Director, Office
of Nuclear Reactor Regulation. An application for an early site permit
may be filed notwithstanding the fact that an application for a CP or a
COL has not been filed in connection with the site for which a permit
is sought.
Sec. 53.1144 Contents of applications for early site permits; general
information.
The application must contain all of the information required by
Sec. 53.1109(a) through (d) and (j).
Sec. 53.1146 Contents of applications for early site permits;
technical information.
(a) The application must contain--
(1) A Site Safety Analysis Report that must include the following:
(i) The specific number, type, and thermal power level of the
facilities, or range of possible facilities, for which the site may be
used;
(ii) The anticipated maximum levels of radiological and thermal
effluents each facility will produce;
(iii) The type of cooling systems, including intakes and outflows,
where appropriate, that may be associated with each facility;
(iv) The boundaries of the site;
(v) The proposed general location of each facility on the site;
(vi) The external hazards and site characteristics required by this
part;
(vii) The location and description of any nearby industrial,
military, or transportation facilities and routes;
(viii) The existing and projected future population profile of the
area surrounding the site;
(ix) A description and assessment of the site on which a facility
is to be located. The assessment must address the requirements of
subpart D of this part;
(x) Information demonstrating that site characteristics are such
that adequate security plans and measures can be developed; and
(xi) A description of the quality assurance program (QAP) required
by appendix B to part 50 of this chapter applied to site-related
activities for the future design, fabrication, construction, and
testing of the SSCs of a facility or facilities that may be constructed
on the site.
(2) A complete environmental report as required by Sec. 51.50(b)
of this chapter.
(b)(1) The Site Safety Analysis Report must identify physical
characteristics of the proposed site, such as egress limitations from
the area surrounding the site, that could pose a significant impediment
to the development of emergency plans. If physical characteristics are
identified that could pose a significant impediment to the development
of emergency plans, the application must identify measures that would,
when implemented, mitigate or eliminate the significant impediment.
(2) The Site Safety Analysis Report may also--
(i) Propose major features of the emergency plans, under either
Sec. 50.160 or the requirements in appendix E to part 50 and Sec.
50.47(b) of this chapter, as applicable, such as the exact size and
configuration of the EPZs, for review and approval by the NRC, in
consultation with the Federal Emergency Management Agency (FEMA), as
applicable, in the absence of complete and integrated emergency plans;
or
(ii) Propose complete and integrated emergency plans for review and
approval by the NRC, in consultation with FEMA, as applicable, in
accordance with either Sec. 50.160 or the requirements in appendix E
to part 50 and Sec. 50.47(b) of this chapter. To the extent approval
of emergency plans is sought, the application must contain the
information required by Sec. 53.1109(g).
(3) Emergency plans submitted under paragraph (b)(2)(ii) of this
section must include the proposed inspections, tests, and analyses that
the holder of a COL referencing the early site permit must perform, and
the acceptance criteria that are necessary and sufficient to provide
[[Page 87082]]
reasonable assurance that, if the inspections, tests, and analyses are
performed and the acceptance criteria met, the facility has been
constructed and will be operated in conformity with the emergency
plans, the provisions of the Act, and the Commission's rules and
regulations. Major features of an emergency plan submitted under
paragraph (b)(2)(i) of this section may include proposed inspections,
tests, analyses, and acceptance criteria (ITAAC).
(4) Under paragraphs (b)(1) and (b)(2)(i) of this section, the Site
Safety Analysis Report must include, where appropriate, a description
of contacts and arrangements made with Federal, State, participating
Tribal, and local governmental agencies with emergency planning
responsibilities. The Site Safety Analysis Report must contain any
certifications that have been obtained. If these certifications, where
appropriate, cannot be obtained, the Site Safety Analysis Report must
contain information, including a utility plan, sufficient to show that
the proposed plans provide reasonable assurance that adequate
protective measures can and will be taken in the event of a
radiological emergency at the site. Under the option set forth in
paragraph (b)(2)(ii) of this section, the applicant must make good
faith efforts, where appropriate, to obtain from the same governmental
agencies certifications that--
(i) The proposed emergency plans are practicable;
(ii) These agencies are committed to participating in any further
development of the plans, including any required field demonstrations;
and
(iii) That these agencies are committed to executing their
responsibilities under the plans in the event of an emergency.
(c) An applicant may request that an LWA under Sec. 53.1130 be
issued in conjunction with the early site permit. The application must
include the information otherwise required by Sec. 53.1130.
(d) Each applicant for an early site permit under this part must
protect safeguards information against unauthorized disclosure in
accordance with the requirements in Sec. Sec. 73.21 and 73.22 of this
chapter, as applicable.
Sec. 53.1149 Review of applications.
(a) Standards for review of applications. Applications filed under
this part will be reviewed according to the applicable standards set
out in this part. In addition, the Commission must prepare an
environmental impact statement during review of the application, under
the applicable provisions of 10 CFR part 51. The Commission must
determine, after consultation with FEMA, as applicable, whether the
information required of the applicant by Sec. 53.1146(b)(1) shows that
there is no significant impediment to the development of emergency
plans that cannot be mitigated or eliminated by measures proposed by
the applicant, whether any major features of emergency plans submitted
by the applicant under Sec. 53.1146(b)(2)(i) are acceptable under
either Sec. 50.160 or appendix E to part 50 and Sec. 50.47(b) of this
chapter, and whether any emergency plans submitted by the applicant
under Sec. 53.1146(b)(2)(ii) provide reasonable assurance that
adequate protective measures can and will be taken in the event of a
radiological emergency.
(b) Administrative review of applications; hearings. An early site
permit application is subject to all procedural requirements in 10 CFR
part 2, including the requirements for docketing in Sec. 2.101(a)(1)
through (4) of this chapter, and the requirements for issuance of a
notice of hearing in Sec. 2.104(a) and (d) of this chapter, provided
that the designated sections may not be construed to require that the
environmental report, or draft or final environmental impact statement
includes an assessment of the benefits of construction and operation of
the reactor or reactors, or an analysis of alternative energy sources.
The presiding officer in an early site permit hearing must not admit
contentions proffered by any party concerning an assessment of the
benefits of construction and operation of the reactor or reactors, or
an analysis of alternative energy sources if those issues were not
addressed by the applicant in the early site permit application. All
hearings conducted on applications for early site permits filed under
this part are governed by the procedures contained in subparts C, G, L,
and N of 10 CFR part 2, as applicable.
Sec. 53.1155 Referral to the Advisory Committee on Reactor
Safeguards.
The Commission must refer a copy of the application for an early
site permit to the Advisory Committee on Reactor Safeguards (ACRS). The
ACRS must report on those portions of the application which concern
safety.
Sec. 53.1158 Issuance of early site permit.
(a) After conducting a hearing under Sec. 53.1149(b) and receiving
the report to be submitted by the ACRS under Sec. 53.1155, the
Commission may issue an early site permit, in the form the Commission
deems appropriate, if the Commission finds that--
(1) An application for an early site permit demonstrates compliance
with the applicable standards and requirements of the Act and the
Commission's regulations;
(2) Notifications, if any, to other agencies or bodies have been
duly made;
(3) There is reasonable assurance that the site is in conformity
with the provisions of the Act and the Commission's regulations;
(4) The applicant is technically qualified to engage in any
activities authorized;
(5) The proposed ITAAC, including any on emergency planning, are
necessary and sufficient, within the scope of the early site permit, to
provide reasonable assurance that the facility has been constructed and
will be operated in conformity with the license, the provisions of the
Act, and the Commission's regulations;
(6) Issuance of the permit will not be inimical to the common
defense and security or to the health and safety of the public;
(7) Any significant adverse environmental impact resulting from
activities requested under Sec. 53.1146(c) can be redressed; and
(8) The findings required by subpart A of 10 CFR part 51 have been
made.
(b) The early site permit must specify the site characteristics,
design parameters, and terms and conditions of the early site permit
the Commission deems appropriate. Before issuance of either a CP or COL
referencing an early site permit, the Commission must find that any
relevant terms and conditions of the early site permit have been met.
Any terms or conditions of the early site permit that could not be met
by the time of issuance of the CP or COL, must be set forth as terms or
conditions of the CP or COL.
(c) The early site permit must specify those Sec. 53.1130(b)
activities requested under Sec. 53.1146(c) that the permit holder is
authorized to perform.
Sec. 53.1161 Extent of activities permitted.
If the activities authorized by Sec. 53.1158(c) are performed and
the site is not referenced in an application for a CP or a COL issued
under this part while the permit remains valid, then the early site
permit remains in effect solely for the purpose of site redress, and
the holder of the permit must redress the site under the terms of the
site redress plan required by Sec. 53.1146(c). If, before redress is
complete, a use not envisaged in the redress plan is found for the site
or parts thereof, the holder of the permit
[[Page 87083]]
must carry out the redress plan to the greatest extent possible
consistent with the alternate use.
Sec. 53.1164 Duration of permit.
(a) Except as provided in paragraph (b) of this section, an early
site permit issued under this subpart may be valid for not less than
10, nor more than 20 years from the date of issuance.
(b) An early site permit continues to be valid beyond the date of
expiration in any proceeding on a CP application or a COL application
that references the early site permit and is docketed before the date
of expiration of the early site permit, or, if a timely application for
renewal of the permit has been docketed, before the Commission has
determined whether to renew the permit.
(c) An applicant for a CP or COL may, at its own risk, reference in
its application a site for which an early site permit application has
been docketed but not granted.
(d) Upon issuance of a CP or COL, a referenced early site permit is
subsumed, to the extent referenced, into the CP or COL.
Sec. 53.1167 Limited work authorization after issuance of early site
permit.
A holder of an early site permit may request an LWA under Sec.
53.1130.
Sec. 53.1170 Transfer of early site permit.
An application to transfer an early site permit will be processed
under Sec. 53.1570.
Sec. 53.1173 Application for renewal.
(a) Not less than 12, nor more than 36 months before the expiration
date stated in the early site permit, or any later renewal period, the
permit holder may apply for a renewal of the permit. An application for
renewal must contain all information necessary to bring up to date the
information and data contained in the previous application.
(b) Any person whose interests may be affected by renewal of the
permit may request a hearing on the application for renewal. The
request for a hearing must comply with Sec. 2.309 of this chapter. If
a hearing is granted, notice of the hearing will be published under
Sec. 2.309 of this chapter.
(c) An early site permit, either original or renewed, for which a
timely application for renewal has been filed, remains in effect until
the Commission has determined whether to renew the permit. If the
permit is not renewed, it continues to be valid in certain proceedings
in accordance with the provisions of Sec. 53.1164(b).
(d) The Commission must refer a copy of the application for renewal
to the ACRS. The ACRS must report on those portions of the application
which concern safety and must apply the criteria set forth in Sec.
53.1176.
Sec. 53.1176 Criteria for renewal.
(a) The Commission must grant the renewal if it determines that--
(1) The site complies with the Act, the Commission's regulations,
and orders applicable and in effect at the time the site permit was
originally issued; and
(2) Any new requirements the Commission may wish to impose--
(i) Are necessary for adequate protection to public health and
safety or common defense and security;
(ii) Are necessary for compliance with the Commission's
regulations, and orders applicable and in effect at the time the site
permit was originally issued; or
(iii) Would provide a substantial increase in overall protection of
the public health and safety or the common defense and security to be
derived from the new requirements, and the direct and indirect costs of
implementation of those requirements are justified in view of this
increased protection.
(b) A denial of renewal under the provisions of Sec. 53.1176(a)
does not bar the permit holder or another applicant from filing a new
application for the site which proposes changes to the site or the way
that it is used to correct the deficiencies cited in the denial of the
renewal.
Sec. 53.1179 Duration of renewal.
Each renewal of an early site permit may be for not less than 10,
nor more than 20 years, plus any remaining years on the early site
permit then in effect before renewal.
Sec. 53.1182 Use of site for other purposes.
A site for which an early site permit has been issued under this
part may be used for purposes other than those described in the permit,
including the location of other types of energy facilities. The permit
holder must inform the Director, Office of Nuclear Reactor Regulation
(Director), of any significant uses for the site which have not been
approved in the early site permit. The information about the activities
must be given to the Director at least 30 days in advance of any actual
construction or site modification for the activities. The information
provided could be the basis for imposing new requirements on the
permit, under the provisions of Sec. 53.1188. If the permit holder
informs the Director that the holder no longer intends to use the site
for a commercial nuclear plant, the Director may terminate the permit.
Sec. 53.1188 Finality of early site permit determinations.
(a) Commission finality. (1) While an early site permit is in
effect under Sec. 53.1164 or Sec. 53.1179, the Commission may not
change or impose new site characteristics, design parameters, or terms
and conditions, including emergency planning requirements, on the early
site permit unless the Commission--
(i) Determines that a modification is necessary to bring the permit
or the site into compliance with the Commission's regulations and
orders applicable and in effect at the time the permit was issued;
(ii) Determines the modification is necessary to assure adequate
protection of the public health and safety or the common defense and
security;
(iii) Determines that a modification is necessary based on an
update under paragraph (b) of this section; or
(iv) Issues a variance requested under paragraph (d) of this
section.
(2) In making the findings required for issuance of a CP, COL, or
OL, or the findings required by Sec. 53.1452(g), or in any enforcement
hearing other than one initiated by the Commission under paragraph
(a)(1) of this section, if the application for the CP, COL, or OL
references an early site permit, the Commission must treat as resolved
those matters resolved in the proceeding on the application for
issuance or renewal of the early site permit, except as provided for in
paragraphs (b), (c), and (d) of this section.
(i) If the Commission grants a CP application that references an
early site permit and an application for an OL references the CP, the
Commission must treat as resolved those matters resolved in the
proceeding for the issuance or renewal of the early site permit, except
as provided for in paragraphs (b), (c), and (d) of this section.
(ii) If the early site permit approved an emergency plan (or major
features thereof) that is in use by a licensee of a commercial nuclear
plant, the Commission must treat as resolved changes to the early site
permit emergency plan (or major features thereof) that are identical to
changes made to the licensee's emergency plans under Sec. 53.1565
occurring after issuance of the early site permit.
(iii) If the early site permit approved an emergency plan (or major
features thereof) that is not in use by a licensee of a commercial
nuclear plant, the Commission must treat as resolved changes that are
equivalent to those that could be made under Sec. 53.1565 without
prior NRC approval had the emergency plan been in use by a licensee.
[[Page 87084]]
(b) Updating of early site permit-emergency preparedness. An
applicant for a CP, OL, or COL who has filed an application referencing
an early site permit issued under this subpart must update the
emergency preparedness information that was provided under Sec.
53.1146(b) and discuss whether the updated information materially
changes the bases for compliance with applicable NRC requirements.
(c) Hearings and petitions. (1) In any proceeding for the issuance
of a CP, OL, or COL referencing an early site permit, contentions on
the following matters may be litigated in the same manner as other
issues material to the proceeding:
(i) The nuclear reactor proposed to be built does not fit within
one or more of the site characteristics or design parameters included
in the early site permit;
(ii) One or more of the terms and conditions of the early site
permit have not been met;
(iii) A variance requested under paragraph (d) of this section is
unwarranted or should be modified;
(iv) New or additional information is provided in the application
that substantially alters the bases for a previous NRC conclusion or
constitutes a sufficient basis for the Commission to modify or impose
new terms and conditions related to emergency preparedness; or
(v) Any significant environmental issue that was not resolved in
the early site permit proceeding, or any issue involving the impacts of
construction and operation of the facility that was resolved in the
early site permit proceeding for which significant new information has
been identified.
(2) Any person may file a petition requesting that the site
characteristics, design parameters, or terms and conditions of the
early site permit be modified, or that the permit be suspended or
revoked. The petition will be considered under Sec. 2.206 of this
chapter. Before construction commences, the Commission must consider
the petition and determine whether any immediate action is required. If
the petition is granted, an appropriate order will be issued.
Construction under the CP or COL will not be affected by the granting
of the petition unless the order is made immediately effective. Any
change required by the Commission in response to the petition must
demonstrate compliance with the requirements of paragraph (a)(1) of
this section.
(d) Variances. An applicant for a CP, OL, or COL referencing an
early site permit may include in its application a request for a
variance from one or more site characteristics, design parameters, or
terms and conditions of the early site permit, or from the Site Safety
Analysis Report. In determining whether to grant the variance, the
Commission must apply the same technically relevant criteria applicable
to the application for the original or renewed early site permit. Once
a CP or COL referencing an early site permit is issued, variances from
the early site permit will not be granted for that CP or COL.
(e) Early site permit amendment. The holder of an early site permit
may not make changes to the early site permit or the Site Safety
Analysis Report without prior Commission approval. The request for a
change to the early site permit must be in the form of an application
for a license amendment and must demonstrate compliance with the
requirements of Sec. Sec. 53.1510 and 53.1520.
Sec. 53.1200 Standard design approvals.
Sections 53.1200 through 53.1221 set out procedures for the filing,
NRC staff review, and referral to the ACRS of standard designs, or
major portions thereof, for a commercial nuclear plant under this part.
Sec. 53.1203 Filing of applications.
Any person may submit a proposed standard design for a commercial
nuclear plant to the NRC staff for its review. The submittal may
consist of either the final design for the entire facility or the final
design for major portions thereof.
Sec. 53.1206 Contents of applications for standard design approvals;
general information.
The application must contain all of the information required by
Sec. 53.1109(a) through (c) and (j).
Sec. 53.1209 Contents of applications for standard design approvals;
technical information.
(a) Major portion of a standard design. If the applicant seeks
review of a major portion of a standard design, the application need
only contain the information required by this section to the extent the
requirements are applicable to the major portion of the standard design
for which NRC staff approval is sought. If an applicant seeks approval
of a major portion of the design, the scope of the application for
which approval is sought must include all functional design criteria
necessary to demonstrate compliance with the safety criteria in
Sec. Sec. 53.210, 53.220 and 53.450(e), as applicable, for the major
portion of the standard design for which NRC staff approval is sought.
Such applicants must identify conditions related to interfaces with
systems outside the scope of the major portion of the standard design
for which NRC staff approval is sought, and functional or physical
boundary conditions between the major portion of the standard design
for which NRC staff approval is sought and the remainder of the
standard design. These conditions must be demonstrated when the
standard design approval is incorporated into a subsequent CP, design
certification, ML, or COL application.
(b) Final Safety Analysis Report. The application must contain an
FSAR that describes the facility and the limits on its operation,
presents a safety analysis of the SSCs and of the facility, or major
portions thereof, for which the applicant seeks design approval, and
must include the following information:
(1) Site Parameters. The site parameters postulated for the design
under this part, including the design-basis external hazard levels for
the relevant external hazards, and an analysis and evaluation of the
design in terms of those site parameters.
(2) Design information. Except as specified in this paragraph, an
application for a standard design approval for a commercial nuclear
plant must include the design information equivalent to that required
for a standard design certification under Sec. 53.1239(a)(2) through
(27) for those portions of a commercial nuclear plant included in the
standard design approval.
Sec. 53.1210 Contents of applications for standard design approvals;
other application content.
(a) In addition to the FSAR, the application must also include the
following:
(1) Availability Controls (if not included in the FSAR). A
description of the controls on plant operations, including availability
controls, to provide reasonable confidence that the configurations and
special treatments for safety-related (SR) SSCs and non-safety-related
but safety-significant (NSRSS) SSCs provide the capabilities and
reliabilities required to demonstrate compliance with the safety
criteria of Sec. 53.220.
(2) Safeguards Information. A description of the program to protect
Safeguards Information against unauthorized disclosure in accordance
with the requirements in Sec. Sec. 73.21 and 73.22 of this chapter, as
applicable.
(b) If there are SSCs of the plant which required research and
development to confirm the adequacy of their design, provide a report
in the application which documents the
[[Page 87085]]
resolution of any safety questions associated with such SSCs.
(c) A description of how the performance of each design feature has
been demonstrated capable of fulfilling functional design criteria
considering interdependent effects through either analysis, appropriate
test programs, prototype testing, operating experience, or a
combination thereof, in accordance with Sec. 53.440(a).
Sec. 53.1212 Standards for review of applications.
Applications filed under this part will be reviewed under the
standards set out in 10 CFR parts 20, 53, and 73.
Sec. 53.1215 Referral to the Advisory Committee on Reactor
Safeguards.
The Commission must refer a copy of the application to the ACRS.
The ACRS must report on those portions of the application which concern
safety.
Sec. 53.1218 Staff approval of design.
(a) Upon completion of its review of a submittal under Sec. Sec.
53.1200 through 53.1221 and receipt of a report by the ACRS under Sec.
53.1215, the NRC staff must publish a determination in the Federal
Register as to whether or not the design is acceptable, subject to
appropriate terms and conditions, and make an analysis of the design in
the form of a report available at the NRC website, https://www.nrc.gov.
(b) A standard design approval issued under this section is valid
for 15 years from the date of issuance and may not be renewed. A design
approval continues to be valid beyond the date of expiration in any
proceeding on an application for a CP, OL, COL, or ML under this part
that references the design approval and is docketed before the date of
expiration of the design approval.
Sec. 53.1221 Finality of standard design approvals; information
requests.
(a) An approved design must be used by and relied upon by the NRC
staff and the ACRS in their reviews of any standard design
certification or individual facility license application under this
part that incorporates by reference a standard design approved under
this part unless there exists significant new information that
substantially affects the earlier determination or other good cause.
(b) The determination and report by the NRC staff do not constitute
a commitment to issue a permit or license, or in any way affect the
authority of the Commission, Atomic Safety and Licensing Board Panel,
or presiding officers in any proceeding under part 2 of this chapter.
(c) Except for information requests seeking to verify compliance
with the current licensing basis of the standard design approval,
information requests to the holder of a standard design approval must
be evaluated before issuance to ensure that the burden to be imposed on
respondents is justified in view of the potential safety significance
of the issue to be addressed in the requested information. Each
evaluation performed by the NRC staff must be in accordance with Sec.
53.1580 and must be approved by the Executive Director for Operations
or authorized designee before issuance of the request.
(d) The Commission will require, before granting a CP, COL, OL, or
ML that references a standard design approval, that engineering
documents, such as analyses, drawings, procurement specifications, or
construction and installation specifications, be completed and
available for audit if the more detailed information is necessary for
the Commission to verify the information in the application and make
its safety determination, including the determination that the
application is consistent with the design approval information. This
information may be acquired by appropriate arrangements with the design
approval applicant.
Sec. 53.1230 Standard design certifications.
Sections 53.1230 through 53.1263 set forth the requirements and
procedures applicable to the Commission's issuance of rules granting
standard design certifications for commercial nuclear plants under this
part separate from the filing of an application for a CP or COL for
such a facility.
Sec. 53.1233 Filing of applications.
(a) An application for design certification may be filed
notwithstanding the fact that an application for a CP, COL, or ML for
such a facility has not been filed.
(b) The application must comply with the applicable filing
requirements of Sec. 53.040 and Sec. Sec. 2.811 through 2.819 of this
chapter.
Sec. 53.1236 Contents of applications for standard design
certifications; general information.
The application must contain all of the information required by
Sec. 53.1109(a) through (c) and (j).
Sec. 53.1239 Contents of applications for standard design
certifications; technical information.
The application must contain a level of design information
sufficient to enable the Commission to judge the applicant's proposed
means of assuring that construction conforms to the design and to reach
a final conclusion on all safety questions associated with the design
before the certification is granted. The information submitted for a
design certification must include performance requirements and design
information sufficiently detailed to permit the preparation of
acceptance and inspection requirements by the NRC. The Commission will
require, before design certification, that information normally
contained in engineering documents, such as analyses, drawings,
procurement specifications, or construction and installation
specifications, be completed and available for audit if the more
detailed information is necessary for the Commission to verify the
information in the application and make its safety determination.
(a) Final Safety Analysis Report. The application must contain an
FSAR that describes the facility and the limits on its operation, and
presents a safety analysis of the SSCs, and must include the following
information:
(1) Site Parameters. The site parameters postulated for the design
under this part, including the design-basis external hazard levels for
the relevant external hazards, and an analysis and evaluation of the
design in terms of those site parameters.
(2) Plant Description and Safety Functions--(i) General Plant
Description. A general description of the commercial nuclear plant
including reactor type, the intended use of the reactor, nuclear design
(e.g., neutron spectrum, reactor control, multi-unit reactor control),
overall layout of the plant including significant plant features and
SSCs, maximum power level and the nature and inventory of radioactive
materials.
(ii) Safety functions. A description of the primary and additional
safety functions required under Sec. 53.230 and a summary of how each
safety function is satisfied.
(3) Design Features and functional design criteria--licensing-basis
events. (i) A description of the design features required by Sec.
53.400 and the functional design criteria required by Sec. Sec. 53.410
and 53.420 that, when combined with corresponding human actions and
programmatic controls, demonstrate that the plant will demonstrate
compliance with the safety criteria defined in Sec. 53.210 and
established in accordance with Sec. 53.220, or more restrictive
alternative criteria adopted under Sec. 53.470, during licensing-basis
events (LBEs).
(ii) A description of how design features demonstrate compliance
with
[[Page 87086]]
the requirements of Sec. 53.440(a) through (i) and (k) through (m).
(4) Design Features Supporting Normal Operations. A description of
the design features required by Sec. 53.425 to support the holder of
an OL or COL complying with Sec. 53.260 during normal operations.
(5) Design Features and Functional Design Criteria--aircraft
impact. A description of the design features and functional design
criteria required to demonstrate compliance with the requirements of
Sec. 53.440(j) for addressing the impact of a large, commercial
aircraft.
(6) Earthquake engineering. The information necessary to
demonstrate that the commercial nuclear plant complies with the
earthquake engineering criteria in Sec. 53.480.
(7) Programmatic Controls and Interfaces. (i) A description of the
corresponding programmatic controls and interfaces necessary to achieve
and maintain the reliability and capability of SSCs relied upon to
demonstrate compliance with the functional design criteria required by
Sec. Sec. 53.410 and 53.420 and the safety criteria in Sec. Sec.
53.210 and 53.220, or more restrictive alternative criteria adopted
under Sec. 53.470, and necessary to maintain consistency with analyses
required by Sec. 53.450.
(ii) For an application for a multi-unit commercial nuclear plant,
the programmatic controls and interfaces must also be described for
different modular configurations, as required by Sec. 53.440(i),
including any restrictions that will be necessary during the
construction and startup of any given unit to ensure the safe operation
of the overall commercial nuclear plant to be licensed under this part.
(8) Programmatic Controls for Normal Operations. A description of
how programmatic controls, including monitoring programs, would provide
assurance that design features and procedures will enable the holder of
an OL or COL to comply with Sec. 53.260.
(9) Design Features Supporting the Protection of Plant Workers. A
description of the design features required by Sec. 53.430 to support
the holder of an OL or COL complying with Sec. 53.270.
(10) Programmatic Controls for Protection of Plant Workers. A
description of how programmatic controls, including monitoring
programs, would provide assurance that design features and procedures
will enable the holder of an OL or COL to comply with Sec. 53.270.
(11) Codes and Standards. A description of generally accepted
consensus codes and standards used to design the design features, as
required by Sec. 53.440(b).
(12) Materials. A description of the materials used for SR and
NSRSS SSCs and a description of the qualification of these materials
for their service conditions over the plant lifetime, as required by
Sec. 53.440(c).
(13) Integrity Assessment Program. A description of a design
integrity assessment program that addresses the elements described in
Sec. 53.440(d).
(14) Safety and Security. Confirmation that safety and security
were considered together in the design process, as required by Sec.
53.440(f).
(15) Criticality. Information demonstrating how the applicant will
comply with requirements for criticality accidents in Sec. 53.440(m).
(16) Multi-unit Plants. For an application for standard design
certification of a multi-unit commercial nuclear plant, the possible
operating configurations of the reactor units, including common
systems, interface requirements, and system interactions, as required
by Sec. 53.440(i).
(17) SSC Classification. (i) The classification of SSCs according
to their safety significance under Sec. 53.460(a).
(ii) For SR and NSRSS SSCs, the conditions under which they must
perform the safety functions required by Sec. 53.230, including
environmental conditions.
(18) Probabilistic Risk Assessment. A description of the
probabilistic risk assessment (PRA) required by Sec. 53.450(a) and its
results.
(19) Analyses. A description of the analyses performed under Sec.
53.450(b) through (g) that includes the following information:
(i) A description of the analysis of LBEs and its results, as
described in Sec. 53.240. This analysis description must--
(A) Address the elements in Sec. 53.450(e) and (f); and
(B) Under Sec. 53.460(c)--
(1) Describe any human actions that are necessary to prevent or
mitigate LBEs;
(2) Describe how those human actions are capable of being reliably
performed under the postulated environmental conditions present; and
(3) Describe how those human actions would be addressed by programs
established under subpart F of this part.
(ii)(A) A description of how SSCs relied on to meet the safety
criteria defined in Sec. 53.210 are protected against or designed to
withstand the effects of external hazards under Sec. 53.510.
(B) The information necessary to demonstrate that the commercial
nuclear plant complies with the earthquake engineering criteria in
Sec. 53.480.
(iii) A description of the defense-in-depth measures required by
Sec. 53.250.
(iv) A description of all plant operating states where there is the
potential for the uncontrolled release of radioactive material to the
environment, as required by Sec. 53.450(b)(4).
(v) A description of the events that challenge plant control and
safety systems whose failure could lead to an undesirable end state
and/or radioactive material release, as required by Sec. 53.450(b)(5).
(vi) A description of the analytical codes used in modeling plant
behavior in analyses of LBEs and how these codes are qualified for the
range of conditions for which they were used, as required by Sec.
53.450(d).
(vii) If not described in addressing paragraph (5) of this section,
the results of other analyses required by Sec. 53.450(g).
(20) Special Treatments. A description of special treatments
established as required by Sec. 53.460.
(21) Analytical Margins. A description of any alternative criteria
adopted to demonstrate analytical margins supporting operational
flexibilities, if applicable, as required by Sec. 53.470.
(22) Quality Assurance. A description of the QAP applied to the
design of the SSCs of the commercial nuclear plant, as required by
Sec. 53.460(b). The description of the QAP for a commercial nuclear
plant must include a discussion of how the applicable requirements of
appendix B to part 50 of this chapter were satisfied.
(23) Design Features and Controls to Address the Minimization of
Contamination. The information required by Sec. 20.1406 of this
chapter.
(24) Interface Requirements. (i) A description analysis, and
evaluation of the interfaces between the standard design and the
balance of the commercial nuclear plant that may impact the ability of
the plant to demonstrate compliance with the functional design criteria
or the safety criteria of subparts B and C of this part.
(ii) Confirmation that interface requirements are verifiable
through inspections, testing, or analysis. These requirements must be
sufficiently detailed to allow for completion of the final safety
analysis by license applicants that reference the certified design
under this subpart. The method to be used for verification of interface
requirements must be included as part of the proposed ITAAC required by
Sec. 53.1241(a)(3).
[[Page 87087]]
(iii) A representative conceptual design for those portions of the
plant for which the application does not seek certification to aid the
NRC in its review of the FSAR and to permit assessment of the adequacy
of the interface requirements under paragraph (a)(24)(i) of this
section.
(25) Technical Qualifications. A description of the technical
qualifications of the applicant to engage in the proposed activities in
accordance with the regulations in this chapter.
(26) Technical Specifications. Proposed technical specifications
prepared under Sec. 53.710(a) for those areas addressed by the design
certification.
(27) Role of personnel. Information to address the following for
the role of personnel in ensuring safe operations:
(i) A description of how the human factors engineering design
requirements of Sec. 53.440(n)(1) are addressed;
(ii) A description of how the human system interface design
requirements of Sec. 53.440(n)(2) are addressed;
(iii) A concept of operations that is of sufficient scope and
detail to address the requirements of Sec. 53.440(n)(3);
(iv) A functional requirements analysis and function allocation
that is of sufficient scope and detail to address the requirements of
Sec. 53.440(n)(4).
(b) [Reserved]
Sec. 53.1241 Contents of applications for standard design
certifications; other application content.
(a) In addition to the FSAR, the application must also include the
following:
(1) Environmental report. An environmental report as required by
Sec. 51.55 of this chapter.
(2) Availability Controls (if not included in the FSAR). A
description of the controls on plant operations, including availability
controls, to provide reasonable confidence that the configurations and
special treatments for SR and NSRSS SSCs provide the capabilities and
reliabilities required to demonstrate compliance with the safety
criteria of Sec. 53.220, or more restrictive alternative criteria
adopted under Sec. 53.470.
(3) Inspections, tests, analyses, and acceptance criteria. The
proposed ITAAC that are necessary and sufficient to provide reasonable
assurance that, if the inspections, tests, and analyses are performed
and the acceptance criteria met, a facility that incorporates the
design certification has been constructed and will be operated in
conformity with the design certification, the provisions of the Act,
and the Commission's rules and regulations.
(4) Safeguards information. A description of the program to protect
Safeguards Information against unauthorized disclosure in accordance
with the requirements in Sec. Sec. 73.21 and 73.22 of this chapter, as
applicable.
(b) If there are SSCs of the plant which required research and
development to confirm the adequacy of their design, provide a report
in the application which documents the resolution of any safety
questions associated with such SSCs.
(c) A description of how the performance of each design feature has
been demonstrated capable of fulfilling functional design criteria
considering interdependent effects through either analysis, appropriate
test programs, prototype testing, operating experience, or a
combination thereof, in accordance with Sec. 53.440(a).
Sec. 53.1242 Review of applications.
(a) Standards for review of applications. Applications filed under
this part will be reviewed for compliance with the standards set out in
this part and 10 CFR parts 20, 51, and 73.
(b) Administrative review of applications; hearings. (1) A standard
design certification is a rule that will be issued under the provisions
of subpart H of 10 CFR part 2, as supplemented by the provisions of
this section. The Commission must initiate the rulemaking after an
application has been filed under Sec. 53.1233 and must specify the
procedures to be used for the rulemaking. The notice of proposed
rulemaking published in the Federal Register must provide an
opportunity for the submission of comments on the proposed design
certification rule. If, at the time a proposed design certification
rule is published in the Federal Register under this paragraph, the
Commission decides that a legislative hearing should be held, the
information required by Sec. 2.1502(c) of this chapter must be
included in the Federal Register document for the proposed design
certification.
(2) Following the submission of comments on the proposed design
certification rule, the Commission may, at its discretion, hold a
legislative hearing under the procedures in subpart O of part 2 of this
chapter. The Commission must publish a document in the Federal Register
of its decision to hold a legislative hearing. The document must
contain the information specified in Sec. 2.1502(c) of this chapter
and specify whether the Commission or a presiding officer will conduct
the legislative hearing.
(3) Notwithstanding anything in Sec. 2.390 of this chapter to the
contrary, proprietary information will be protected in the same manner
and to the same extent as proprietary information submitted in
connection with applications for licenses, provided that the design
certification will be published in chapter I of this title.
(c) Reference to an issued operating license or combined license.
In those cases where a design certification application is preceded by
the issuance of an OL or custom COL for a commercial nuclear plant that
is essentially the same as the standard design for which certification
is being requested, the NRC review will follow the processes for
referencing a standard design approval in Sec. 53.1221, to the extent
practicable.
Sec. 53.1245 Referral to the Advisory Committee on Reactor
Safeguards.
The Commission must refer a copy of the application to the ACRS.
The ACRS must report on those portions of the application which concern
safety.
Sec. 53.1248 Issuance of standard design certification.
(a) After conducting a rulemaking proceeding under Sec. 53.1242 on
an application for a standard design certification and receiving the
report to be submitted by the ACRS under Sec. 53.1245, the Commission
may issue a standard design certification in the form of a rule for the
design that is the subject of the application, if the Commission
determines that--
(1) The application demonstrates compliance with the applicable
standards and requirements of the Act and the Commission's regulations;
(2) Notifications, if any, to other agencies or bodies have been
duly made;
(3) There is reasonable assurance that the standard design conforms
with the provisions of the Act and the Commission's regulations;
(4) The applicant is technically qualified;
(5) The proposed ITAAC are necessary and sufficient, within the
scope of the standard design, to provide reasonable assurance that, if
the inspections, tests, and analyses are performed and the acceptance
criteria met, the facility has been constructed and will be operated in
accordance with the design certification, the provisions of the Act,
and the Commission's regulations;
(6) Issuance of the standard design certification will not be
inimical to the common defense and security or to the health and safety
of the public;
[[Page 87088]]
(7) The findings required by subpart A of part 51 of this chapter
have been made; and
(8) The applicant has implemented the QAP described or referenced
in the Safety Analysis Report.
(b) The design certification rule must specify the site parameters,
design characteristics, and any additional requirements and
restrictions of the design certification rule.
(c) After the Commission has adopted a final design certification
rule, the applicant must not permit any individual to have access to or
any facility to possess restricted data or classified National Security
Information until the individual and/or facility has been approved for
access under the provisions of 10 CFR parts 25 and/or 95, as
applicable.
Sec. 53.1251 Duration of certification.
(a) Except as provided in paragraph (b) of this section, a standard
design certification issued under this subpart is valid for 15 years
from the effective date of the rule.
(b) A standard design certification continues to be valid beyond
the date of expiration in any proceeding on an application for a COL or
an OL under this part that references the standard design certification
and is docketed either before the date of expiration of the
certification, or, if a timely application for renewal of the
certification has been filed, before the Commission has determined
whether to renew the certification. A design certification also
continues to be valid beyond the date of expiration in any hearing held
under Sec. 53.1452 before operation begins under a COL that references
the design certification.
(c) An applicant for a CP, OL, COL, or ML under this part may, at
its own risk, reference in its application a design for which a design
certification application has been docketed but not granted.
Sec. 53.1254 Application for renewal.
(a) Not less than 12 nor more than 36 months before the expiration
of the initial 15-year period, or any later renewal period, any person
may apply for renewal of the certification. An application for renewal
must contain all information necessary to bring up to date the
information and data contained in the previous application. The
Commission will require, before renewal of certification, that
engineering documents, such as analyses, drawings, procurement
specifications, or construction and installation specifications, be
completed and available for audit if the more detailed information is
necessary for the Commission to verify the information in the
application and make its safety determination. Notice and comment
procedures must be used for a rulemaking proceeding on the application
for renewal. The Commission, in its discretion, may require the use of
additional procedures in individual renewal proceedings.
(b) A design certification, either original or renewed, for which a
timely application for renewal has been filed remains in effect until
the Commission has determined whether to renew the certification. If
the certification is not renewed, it continues to be valid in certain
proceedings under Sec. 53.1251.
(c) The Commission must refer a copy of the application for renewal
to the ACRS. The ACRS must report on those portions of the application
which concern safety and must apply the criteria set forth in Sec.
53.1257.
Sec. 53.1257 Criteria for renewal.
(a) The Commission must issue a rule granting the renewal if the
design, either as originally certified or as modified during the
rulemaking on the renewal, complies with the Act and the Commission's
regulations applicable and in effect at the time the certification was
issued.
(b) The Commission may impose other requirements if it determines
that--
(1) They are necessary for adequate protection to public health and
safety or common defense and security;
(2) They are necessary for compliance with the Commission's
regulations and orders applicable and in effect at the time the design
certification was issued; or
(3) There is a substantial increase in overall protection of the
public health and safety or the common defense and security to be
derived from the new requirements, and the direct and indirect costs of
implementing those requirements are justified in view of this increased
protection.
(c) In addition, the applicant for renewal may request an amendment
to the design certification. The Commission must grant the amendment
request if it determines that the amendment will comply with the Act
and the Commission's regulations in effect at the time of renewal. If
the amendment request entails such an extensive change to the design
certification that an essentially new standard design is being
proposed, an application for a design certification must be filed in
accordance with this subpart.
(d) Denial of renewal does not bar the applicant, or another
applicant, from filing a new application for certification of the
design, which proposes design changes that correct the deficiencies
cited in the denial of the renewal.
Sec. 53.1260 Duration of renewal.
Each renewal of certification for a standard design will be for not
less than 10, nor more than 15 years.
Sec. 53.1263 Finality of standard design certifications.
(a)(1) While a standard design certification rule is in effect
under Sec. Sec. 53.1251 or 53.1260, the Commission may not modify,
rescind, or impose new requirements on the certification information,
whether on its own motion, or in response to a petition from any
person, unless the Commission determines in a rulemaking that the
change--
(i) Is necessary either to bring the certification information or
the referencing plants into compliance with the Commission's
regulations applicable and in effect at the time the certification was
issued;
(ii) Is necessary to provide adequate protection of the public
health and safety or the common defense and security;
(iii) Reduces unnecessary regulatory burden and maintains
protection to public health and safety and the common defense and
security;
(iv) Provides the detailed design information to be verified under
those ITAAC that are directed at certification information (i.e.,
design acceptance criteria);
(v) Is necessary to correct material errors in the certification
information;
(vi) Substantially increases overall safety, reliability, or
security of facility design, construction, or operation, and the direct
and indirect costs of implementation of the rule change are justified
in view of this increased safety, reliability, or security; or
(vii) Contributes to increased standardization of the certification
information.
(2)(i) In a rulemaking under Sec. 53.1263(a)(1), except for Sec.
53.1263(a)(1)(ii), the Commission will give consideration to whether
the benefits justify the costs for plants that are already licensed or
for which an application for a permit or license is under
consideration.
(ii) The rulemaking procedures for changes under Sec.
53.1263(a)(1) must provide for notice and opportunity for public
comment.
(3) Any modification the NRC imposes on a design certification rule
[[Page 87089]]
under paragraph (a)(1) of this section will be applied to all plants
referencing the certified design, except those to which the
modification has been rendered technically irrelevant by action taken
under paragraphs (a)(4) or (b) of this section.
(4) The Commission may not impose new requirements by plant-
specific order on any part of the design of a specific plant
referencing the design certification rule if that part was approved in
the design certification while a design certification rule is in effect
under Sec. 53.1248, unless--
(i) A modification is necessary to secure compliance with the
Commission's regulations applicable and in effect at the time the
certification was issued, or to assure adequate protection of the
public health and safety or the common defense and security; and
(ii) Special circumstances as defined in Sec. 53.080 are present.
In addition to the factors listed in Sec. 53.080, the Commission must
consider whether the special circumstances which Sec. 53.080 requires
to be present outweigh any decrease in safety that may result from the
reduction in standardization caused by the plant-specific order.
(5) Except as provided in Sec. 2.335 of this chapter, in making
the findings required for issuance of a COL, CP, OL, or ML, or for any
hearing under Sec. 53.1452, the Commission must treat as resolved
those matters resolved in connection with the issuance or renewal of a
design certification rule.
(b) An applicant who references a design certification rule may
request an exemption from one or more elements of the certification
information. The Commission may grant such a request only if it
determines that the exemption will comply with the requirements of
Sec. 53.080. In addition to the factors listed in Sec. 53.080, the
Commission must consider whether the special circumstances that Sec.
53.080 requires to be present outweigh any decrease in safety that may
result from the reduction in standardization caused by the exemption.
The granting of an exemption on request of an applicant is subject to
litigation in the same manner as other issues in the OL or COL hearing.
(c) The Commission will require, before granting a CP, COL, OL, or
ML that references a design certification rule, that information
normally contained in engineering documents, such as analyses,
drawings, procurement specifications, or construction and installation
specifications, be completed and available for audit if the more
detailed information is necessary for the Commission to verify the
information in the application and make its safety determination,
including the determination that the application is consistent with the
certification information. This information may be acquired by
appropriate arrangements with the design certification applicant.
Sec. 53.1270 Manufacturing licenses.
Sections 53.1270 through 53.1295 set out the requirements and
procedures applicable to Commission issuance of a license under this
part authorizing manufacture of manufactured reactors to be installed
at sites not identified in the ML application.
Sec. 53.1273 Filing of applications.
Any person, except one excluded by Sec. 53.1118, may file an
application for an ML under this part with the Director, Office of
Nuclear Reactor Regulation.
Sec. 53.1276 Contents of applications for manufacturing licenses;
general information.
Each application for an ML must include the information contained
in Sec. 53.1109(a) through (e), and (j).
Sec. 53.1279 Contents of applications for manufacturing licenses;
technical information.
(a) Final Safety Analysis Report-siting and design. The application
must include an FSAR containing the information set forth below, with a
level of design information sufficient to enable the Commission to
judge the applicant's proposed means of ensuring that the manufacturing
conforms to the design and to reach a final conclusion on all safety
questions associated with the design, permit the preparation of
construction and installation specifications by an applicant who seeks
to use the manufactured reactor, and permit the preparation of
acceptance and inspection requirements by the NRC. The application must
include the following information:
(1) Site parameters. The site parameters postulated for the design
under this part, including the design-basis external hazard levels for
the relevant external hazards, and an analysis and evaluation of the
design in terms of those site parameters.
(2) Design information. Except as specified in this paragraph, the
design information equivalent to that required for a standard design
certification as defined in Sec. 53.1239(a)(2) through (27) for those
portions of a commercial nuclear plant included in the manufactured
reactor.
(3) Quality assurance program. A description of the QAP applied to
the design, and to be applied to the fabrication and testing of the
SSCs of the manufactured reactor under Sec. 53.620(a)(6), including a
discussion of how the applicable requirements of appendix B to part 50
of this chapter will be satisfied;
(4) Conceptual designs. Representative conceptual designs for one
or more commercial nuclear plants using the manufactured reactor;
(5) Operating configurations. If multiple manufactured reactors may
be installed at a commercial nuclear plant, a description of the
possible operating configurations, including common systems, interface
requirements, and system interactions. The final safety analysis must
also account for differences among the possible configurations,
including any restrictions that will be necessary during the
construction and startup of a given manufactured reactor to ensure the
safe operation of any commercial nuclear reactor already operating;
(6) Interface requirements. (i) The interface requirements between
the manufactured reactor and the remaining portions of the commercial
nuclear plant or connections to other facilities outside of the
commercial nuclear plant.
(ii) Confirmation that interface requirements are verifiable
through inspections, testing, or analysis. These requirements must be
sufficiently detailed to allow for completion of the final safety
analysis by license applicants that reference the manufactured reactor
manufactured under this subpart. Applicants for a COL under this part
will need to verify the interface requirements at the installation
site. The method to be used for verification of interface requirements
must be included as part of the proposed ITAAC required by Sec.
53.1282(a).
(iii) Information to support development of radiation monitoring
programs required under subpart F of this part by an applicant for a
COL, including potential pathways for radionuclides produced within the
manufactured reactor to enter interfacing systems.
(b) Final Safety Analysis Report--manufacturing information. The
FSAR must include the following information related to the
manufacturing processes, organization, controls, and inspections:
(1) A description, including references to generally accepted
consensus codes and standards, of the processes that will be used to
procure, fabricate, and assemble components that make up the
manufactured reactor. The description should clearly define which
activities are proposed to be within the
[[Page 87090]]
scope of the ML and those, such as the making of a component to be
procured from a separate company for installation in the manufactured
reactor, that are not considered to be within the scope of the ML;
(2) A description of the organizational and management structure
singularly responsible for direction of design and manufacture of the
manufactured reactor. The information should include a description of
the management plans, technical qualifications, and controls in place
to demonstrate compliance with the requirements of Sec. 53.620;
(3) A description of the inspections and tests to be performed as
part of the manufacturing process, including the inspection of procured
components, inspection and testing of fabrication processes such as the
molding, welding, or coating of components, and inspections and testing
of the assembled manufactured reactor or portions of the manufactured
reactor;
(4) A description of the fitness-for-duty program required by part
26 of this chapter and its implementation.
(c) Deployment of the completed manufactured reactor. The
application must include the following information related to the
deployment of a manufactured reactor:
(1) Procedures governing the preparation of the manufactured
reactor or portions of the manufactured reactor for shipping to the
site where it is to be operated; the conduct of shipping; and verifying
the condition of the shipped items upon receipt at the site;
(2) Details of the interaction of the design, manufacture, and
installation of a manufactured reactor within the applicant's
organization and the manner by which the applicant will ensure close
integration between the designer, contractors, and any facility in
which the manufactured reactor is to be installed;
(3) A description of the measures to be used for the control of
interfaces, including the consideration of key site parameters, between
the holder of the ML and the holder of the COL for the commercial
nuclear plant at which the manufactured reactor is to be installed.
(d) Special considerations for factory fueling. In addition to the
above paragraphs, an application for an ML for a manufactured reactor
that will be fueled at the factory under a 10 CFR part 70 license must
include the following information related to loading fuel and the
required independent physical mechanisms to prevent criticality and to
otherwise provide assurance that the fueled manufactured reactor can be
successfully transported, installed, and operated at a site for which
the Commission has issued a COL that authorizes construction and
operation of a commercial nuclear plant using the manufactured reactor:
(1) A description of the procedures used during the fueling of the
manufactured reactor that ensure that the configuration of fuel within
the fueled manufactured reactor is consistent with the design and
analyses supporting operation of the manufactured reactor under the COL
at the place of operation. The description may reference the applicable
10 CFR part 70 application and other sections of the Safety Analysis
Report supporting the ML license application.
(i) The application must describe the measures taken for in-factory
inspections and non-nuclear testing performed to ensure that the
configuration of fuel within the fueled manufactured reactor is
consistent with the design and analyses supporting operation of the
manufactured reactor under the COL at the place of operation.
(ii) The application must describe the design features included in
the manufactured reactor to prevent criticality, including at least two
independent mechanisms each of which is sufficient to prevent
criticality, the associated functional design criteria applied to those
design features, and the physical and programmatic controls implemented
during manufacturing, storage, and transport that are credited to
assure the features function as designed when subject to potential
hazards and human errors. The descriptions must include how those
measures will be controlled during installation under the ML and
removal under the COL at the place of operation.
(2) A description of the procedures governing the transfer of
responsibilities for the fueled manufactured reactor from the holder of
the ML to the holder of the COL for the installation site.
(3) If available at the time of filing the ML application or, if
not available at the time of filing the ML application, submitted as an
amendment to the ML or ML application at the time of filing the Part 70
application, a description of the programs needed to demonstrate
compliance with the requirements of Sec. 53.620(d) and 10 CFR parts
70, 71, and 73 for the receipt, storage, and loading of SNM into a
manufactured reactor and the transport of the fueled manufactured
reactor to a site for which the Commission has issued a COL that
authorizes construction and operation of a commercial nuclear plant
using the manufactured reactor, including the following.
(i) A physical security program in accordance with Sec.
53.620(d)(2)(i).
(ii) A cybersecurity program in accordance with Sec.
53.620(d)(2)(i).
Sec. 53.1282 Contents of applications for manufacturing licenses;
other application content.
(a) Inspections, tests, analyses, and acceptance criteria. (1) The
application must contain proposed inspections, tests, and analyses that
the COL holder must perform, and the acceptance criteria that are
necessary and sufficient to provide reasonable assurance that, if the
inspections, tests, and analyses are performed and the acceptance
criteria met:
(i) The reactor has been manufactured in conformity with the ML,
the provisions of the Act, and the Commission's rules and regulations;
and
(ii) The manufactured reactor will be operated in conformity with
the approved design and any license authorizing operation of the
manufactured reactor.
(2) If the application references a standard design certification,
the ITAAC contained in the certified design must apply to those
portions of the facility design that are covered by the design
certification.
(3) If the application references a standard design certification,
the application may include a notification that a required inspection,
test, or analysis in the design certification ITAAC has been
successfully completed and that the corresponding acceptance criterion
has been met. The Federal Register notification required by Sec.
53.1285 must indicate that the application includes this notification.
(b) Environmental report. (1) The application must contain an
environmental report as required by Sec. 51.54 of this chapter.
(2) If the ML application references a standard design
certification, the environmental report need not contain a discussion
of severe accident mitigation design alternatives for the manufactured
reactor as used in a commercial nuclear plant.
(c) Safeguards information. The application must contain a
description of the program to protect safeguards information against
unauthorized disclosure in accordance with the requirements in
Sec. Sec. 73.21 and 73.22 of this chapter, as applicable.
(d) Performance demonstration. A description of how the performance
of each design feature has been demonstrated capable of fulfilling
functional design criteria considering interdependent effects through
either analysis, appropriate test programs, prototype testing,
operating experience,
[[Page 87091]]
or a combination thereof, in accordance with Sec. 53.440(a).
Sec. 53.1285 Review of applications.
(a) Standards for review of applications. Applications for MLs
under this part will be reviewed according to the applicable standards
set out in this subpart as well as applicable standards in this part
and 10 CFR parts 20, 25, 26, 51, 70, 71, 73, and 75.
(b) Administrative review of applications, hearings. A proceeding
on an ML is subject to all applicable procedural requirements contained
in 10 CFR part 2, including the requirements for docketing in Sec.
2.101(a)(1) through (4) of this chapter, and the requirements for
issuance of a notice of proposed action in Sec. 2.105 of this chapter,
provided, however, that the designated sections may not be construed to
require that the environmental report or draft or final environmental
impact statement include an assessment of the benefits of constructing
and/or operating the manufactured reactor or an evaluation of
alternative energy sources. All hearings on MLs are governed by the
hearing procedures contained in 10 CFR part 2, subparts C, E, G, L, and
N.
Sec. 53.1286 Referral to the Advisory Committee on Reactor
Safeguards.
The Commission must refer a copy of the application to the ACRS.
The ACRS must report on those portions of the application which concern
safety.
Sec. 53.1287 Issuance of manufacturing licenses.
(a) After completing any hearing under Sec. 53.1285(b), and
receiving the report submitted by the ACRS, the Commission may issue an
ML if the Commission finds that--
(1) Applicable standards and requirements of the Act and the
Commission's regulations have been met;
(2) There is reasonable assurance that the manufactured reactor
will be manufactured, and can be transported, incorporated into a
commercial nuclear plant, and operated in conformity with the ML, the
provision of the Act, and the Commission's regulations;
(3) The proposed manufactured reactor can be incorporated into a
commercial nuclear plant and operated at sites having characteristics
that fall within the site parameters postulated for the design of the
manufactured reactors without undue risk to the health and safety of
the public;
(4) The applicant is technically qualified to design and
manufacture the proposed manufactured reactor;
(5) The proposed ITAAC are necessary and sufficient, within the
scope of the ML, to provide reasonable assurance that the manufactured
reactor has been manufactured and will be operated in conformity with
the license, the provisions of the Act, and the Commission's
regulations;
(6) The issuance of a license to the applicant will not be inimical
to the common defense and security or to the health and safety of the
public; and
(7) The findings required by subpart A of 10 CFR part 51 have been
made.
(b) Each ML issued under this subpart must specify--
(1) Terms and conditions as the Commission deems necessary and
appropriate;
(2) Technical specifications for operation of the manufactured
reactor, as the Commission deems necessary and appropriate;
(3) Site parameters and design characteristics for the manufactured
reactor;
(4) The interface requirements to be met by the site-specific
elements of the facility, such as the energy conversions systems and
ultimate heat sink, not within the scope of the manufactured reactor;
and
(5) The entity with design authority for the manufactured reactor
covered by the license.
Sec. 53.1288 Finality of manufacturing licenses.
(a)(1) Notwithstanding any provision in Sec. 53.1590, during the
term of an ML issued under this part the Commission may not modify,
rescind, or impose new requirements on the design of the manufactured
reactor, or the requirements for the manufacture of the manufactured
reactor, unless the Commission determines that a modification is
necessary to bring the design of the reactor or its manufacture into
compliance with the Commission's requirements applicable and in effect
at the time the ML was issued, or to provide reasonable assurance of
adequate protection to public health and safety or common defense and
security.
(2) Any modification to the design of a manufactured reactor that
is imposed by the Commission under paragraph (a)(1) of this section
will be applied to all manufactured reactors manufactured under the
license, including those that have already been transported and sited,
except those manufactured reactors to which the modification has been
rendered technically irrelevant by action taken under Sec. 53.1530 or
paragraph (b) of this section.
(3) In making the findings required under this part for issuance of
a COL, in any hearing under Sec. 53.1452, or in any enforcement
hearing other than one initiated by the Commission under paragraph
(a)(1) of this section, for which a manufactured reactor manufactured
under this subpart is referenced or used, the Commission must treat as
resolved those matters resolved in the proceeding on the application
for issuance or renewal of the ML, including the adequacy of design of
the manufactured reactor, the costs and benefits of severe accident
mitigation design alternatives, and the bases for not incorporating
severe accident mitigation design alternatives into the design of the
manufactured reactor to be manufactured.
(b) An applicant who references or uses a manufactured reactor
manufactured under an ML under this part may include in the application
a request for a departure from the design characteristics, site
parameters, terms and conditions, or approved design of the
manufactured reactor. The Commission may grant a request only if it
determines that the departure will comply with the requirements of
Sec. 53.080, and that the special circumstances outweigh any decrease
in safety that may result from the reduction in standardization caused
by the departure. The granting of a departure on request of an
applicant is subject to litigation in the same manner as other issues
in the COL hearing.
Sec. 53.1291 Duration of manufacturing licenses.
An ML issued under this part is valid for not less than 5, nor more
than 15 years from the date of issuance. Upon expiration of the ML, the
manufacture of any uncompleted manufactured reactors must cease unless
a timely application for renewal has been docketed with the NRC.
Sec. 53.1293 Transfer of manufacturing licenses.
An ML may be transferred under Sec. 53.1570.
Sec. 53.1295 Renewal of manufacturing licenses.
(a)(1) Not less than 12 months, nor more than 5 years before the
expiration of the ML, or any later renewal period, the holder of the ML
issued under this part may apply for a renewal of the license. An
application for renewal must contain all information necessary to bring
up to date the information and data contained in the previous
application.
(2) The filing of an application for a renewed license must be in
accordance with subpart A of 10 CFR part 2 and Sec. 53.1100.
[[Page 87092]]
(3) An ML issued under this part, either original or renewed, for
which a timely application for renewal has been filed, remains in
effect until the Commission has made a final determination on the
renewal application, provided, however, that the holder of an ML may
not begin manufacture of a manufactured reactor less than 6 months
before the expiration of the license.
(4) Any person whose interest may be affected by renewal of the
license may request a hearing on the application for renewal. The
request for a hearing must comply with Sec. 2.309 of this chapter. If
a hearing is granted, notice of the hearing will be published in
accordance with Sec. 2.104 of this chapter.
(5) The Commission must refer a copy of the application for renewal
to the ACRS. The ACRS must report on those portions of the application
which concern safety.
(b) The Commission may grant the renewal if the Commission
determines--
(1) The ML complies with the Act and the Commission's regulations
and orders applicable and in effect at the time the ML was originally
issued; and
(2) Any new requirements the Commission may wish to impose are--
(i) Necessary for adequate protection to public health and safety
or common defense and security;
(ii) Necessary for compliance with the Commission's regulations and
orders applicable and in effect at the time the ML was originally
issued; or
(iii) A substantial increase in overall protection of the public
health and safety or the common defense and security to be derived from
the new requirements, and the direct and indirect costs of
implementation of those requirements are justified in view of this
increased protection.
(c) A renewed ML may be issued for a term of not less than 5, nor
more than 15 years, plus any remaining years on the ML then in effect
before renewal. The renewed license must be subject to the requirements
of Sec. 53.1288.
Sec. 53.1300 Construction permits.
Sections 53.1300 through 53.1348 set out the requirements and
procedures applicable to Commission issuance of a CP for commercial
nuclear plants. A CP for the construction of a commercial nuclear plant
under this part will be issued before the issuance of an OL if the
application is otherwise acceptable and will be converted upon
completion of the facility and Commission action, into an OL as
provided under Sec. Sec. 53.1360 through 53.1405.
Sec. 53.1306 Contents of applications for construction permits;
general information.
An application for a CP must include the information required by
Sec. 53.1109 and the following information:
(a) Information sufficient to demonstrate to the Commission the
financial qualification of the applicant to carry out, under the
regulations in this chapter, the activities for which the permit is
sought. As applicable, the following should be provided:
(1) The information that demonstrates that the applicant possesses
or has reasonable assurance of obtaining the funds necessary to cover
estimated construction costs and related fuel cycle costs, including
estimates of the total construction costs and related fuel cycle costs
of the facility and must indicate the source(s) of funds to cover these
costs.
(2) Each application for a CP submitted by a newly formed entity
organized for the primary purpose of constructing and operating a
facility must also include information showing:
(i) The legal and financial relationships the entity has or
proposes to have with its stockholders or owners;
(ii) The stockholders' or owners' financial ability to meet any
contractual obligation to the entity that they have incurred or
proposed to incur; and
(iii) Any other information considered necessary by the Commission
to enable it to determine the applicant's financial qualification; and
(3) The Commission may request an established entity or newly-
formed entity to submit additional or more detailed information
respecting its financial arrangements and status of funds if the
Commission considers this information appropriate. This may include
information regarding an applicant's ability to continue the conduct of
the activities authorized by the CP and to decommission the facility.
(b) If the applicant proposes to construct or alter a facility, the
application must state the earliest and latest dates for completion of
the construction or alteration.
Sec. 53.1309 Contents of applications for construction permits;
technical information.
The application must contain a Preliminary Safety Analysis Report
(PSAR) that describes the facility and the limits on its operation and
presents a preliminary safety analysis of the SSCs of the facility as a
whole. The PSAR must include the following information, at a level of
detail sufficient to enable the Commission to reach a conclusion on
safety matters that must be resolved by the Commission before issuance
of a CP:
(a)(1) Site information. An application for a CP for a commercial
nuclear reactor must include the site information equivalent to that
required for an early site permit in Sec. 53.1146(a)(1)(iv) through
(x).
(2) Design information. Except as specified in this paragraph, an
application for a CP for a commercial nuclear plant must include the
design information equivalent to that required for a standard design
certification as defined in Sec. 53.1239(a)(2) through (27).
(i) Quality assurance program. A description of the QAP to be
applied to the design, fabrication, construction, and testing of the
SSCs of the facility under Sec. 53.610(a)(6), including a discussion
of how the requirements of appendix B to part 50 of this chapter will
be satisfied.
(ii) Preliminary design information. The information provided in
the application may include some aspects of the design that are not
fully developed, and the information is therefore preliminary. The
completed design, including any changes during construction, must be
described in the FSAR required in Sec. 53.1369 that supports an
application for an OL.
(iii) Planned research or testing. Descriptions of how design
features and related functional design criteria will fulfill the safety
criteria in subpart B, or more restrictive alternative criteria adopted
under Sec. 53.470, and how that has been or will be demonstrated
through either analysis, appropriate test programs, experience, or a
combination thereof. Where any design feature has not been fully
developed or demonstrated to fulfill the functional design criteria at
the time of an application for a CP, the applicant must provide a plan
for future analysis, research and development, test programs, gathering
of experience, or a combination thereof to provide reasonable
confidence that the required demonstration will be available for an
application for an OL.
(iv) Programmatic controls. Descriptions of the programmatic
controls may include those to be provided in the FSAR or other
licensing basis documents because they are necessary to achieve and
maintain the reliability and capability of SSCs relied upon to
demonstrate compliance with the established safety criteria and
functional design criteria required in subpart B, and to maintain
consistency with analyses required by Sec. 53.450.
(3) Technical qualifications. A description of the technical
qualifications of the applicant to engage in the proposed activities
under the regulations in this chapter.
[[Page 87093]]
(4) Emergency preparedness. A description of the applicant's
preliminary plans for coping with emergencies based on:
(i) Except as provided in paragraph (a)(4)(ii) of this section, the
requirements in appendix E to part 50.
(ii) For a commercial nuclear plant consisting of either small
modular reactors or non-light-water reactors, the requirements in
either Sec. 50.160 or appendix E to part 50.
(5) Physical security. A report that provides a preliminary
description of how the site characteristics support the development of
adequate security plans and measures consistent with the requirements
in Sec. 53.540.
(6) Fitness-for-duty program. A description of the fitness-for-duty
(FFD) program required by 10 CFR part 26 and its implementation.
(b) A description of the program to protect Safeguards Information
against unauthorized disclosure in accordance with the requirements in
Sec. Sec. 73.21 and 73.22 of this chapter, as applicable.
Sec. 53.1312 Contents of applications for construction permits; other
application content.
(a) In addition to the PSAR, the application must include the
following:
(1) An environmental report either under Sec. 51.50(a) of this
chapter if an LWA under Sec. 53.1130 is not requested in conjunction
with the CP application, or under Sec. Sec. 51.49 and 51.50(a) of this
chapter if an LWA is requested in conjunction with the CP application;
or
(2) If the applicant wishes to request that an LWA under Sec.
53.1130 be issued before issuance of the CP, the information otherwise
required by Sec. 53.1130, in accordance with either Sec. 2.101(a)(1)
through (a)(5), or Sec. 2.101(a)(9) of this chapter.
(b) If the CP application references an early site permit, standard
design approval, or standard design certification issued under this
part, then the following requirements apply:
(1) The PSAR need not contain information or analyses submitted to
the Commission in connection with the referenced NRC approval, permit,
or certification, provided, however, that the PSAR incorporates the
material by reference and confirms that the site and design of the
facility falls within parameter values postulated in the referenced NRC
approval, permit, or certification.
(2) The PSAR must provide a means to demonstrate that all terms and
conditions that have been included in the referenced NRC approval,
permit, or certification will be satisfied by the date of issuance of
the OL, as appropriate. If the PSAR does not demonstrate that each site
characteristic falls within the corresponding postulated site parameter
and each design characteristic of the facility falls within the
corresponding postulated design parameter, the application must justify
a departure, variance, or exemption from the referenced NRC approval,
license, or certification in regard to that particular site or design
characteristic in compliance with the requirements of this part.
(3) If a referenced early site permit approves complete and
integrated emergency plans, or major features of emergency plans, then
the PSAR must include any new or additional information that updates
and corrects the information that was provided under Sec.
53.1146(b)(2) and discuss whether the new or additional information
materially changes the bases for compliance with the applicable
requirements.
Sec. 53.1315 Review of applications.
(a) Standards for review of applications. Applications filed under
this part will be reviewed according to the standards set out in this
part and 10 CFR parts 20, 51, 73, and 140.
(b) Administrative review of applications; hearings. A proceeding
on a CP application is subject to all applicable procedural
requirements contained in 10 CFR part 2, including the requirements for
docketing (Sec. 2.101 of this chapter) and issuance of a notice of
hearing (Sec. 2.104 of this chapter). All hearings on CP applications
are governed by the procedures contained in 10 CFR part 2.
Sec. 53.1318 Finality of referenced NRC approvals, permits, and
certifications.
If the application for a CP under this part references an early
site permit, standard design approval, or standard design
certification, the scope and nature of matters resolved for the
application are governed by the relevant provisions addressing
finality, including Sec. Sec. 53.1188, 53.1221, and 53.1263.
Sec. 53.1324 Referral to the Advisory Committee on Reactor
Safeguards.
The Commission must refer a copy of the application to the ACRS.
The ACRS must report on those portions of the application that concern
safety and must apply the standards referenced in Sec. 53.1315, in
accordance with the finality provisions in Sec. 53.1318.
Sec. 53.1327 Authorization to conduct limited work authorization
activities.
(a) If the application does not reference an early site permit
which authorizes the holder to perform the activities under Sec.
53.1130, the applicant may not perform those activities without
obtaining the separate authorization required by Sec. 53.1130.
Authorization may be granted only after the presiding officer in the
proceeding on the application has made the findings and determination
required by Sec. 53.1130(b)(1)(ii) and (iv), and the Director, Office
of Nuclear Reactor Regulation makes the determination required by Sec.
53.1130(b)(1)(iii).
(b) If, after an applicant has performed the activities permitted
by paragraph (a) of this section, the application for the CP is
withdrawn or denied, then the applicant must implement an approved site
redress plan.
Sec. 53.1330 Exemptions, departures, and variances.
(a) Applicants for a CP under this part, or any amendment to a CP,
may include in the application a request for an exemption from one or
more of the Commission's regulations. The Commission may grant a
request if it determines that the exemption complies with Sec. 53.080.
(b) An applicant for a CP who has filed an application referencing
an NRC approval, permit, or certification issued under this part may
include in the application a request for exemptions, departures, or
variances related to the subject referenced NRC approval, permit, or
certification. In determining whether to grant the departure, variance,
or exemption, the Commission must apply the same technically relevant
criteria as were applicable to the application for the original or
renewed approval, license, or certification.
Sec. 53.1333 Issuance of construction permits.
(a) After conducting a hearing in accordance with Sec. 53.1315 and
receiving the report submitted by the ACRS, the Commission may issue a
CP only if the Commission finds that--
(1) The applicant has described the proposed design of the facility
and has identified the major features or components incorporated
therein for the protection of the health and safety of the public;
(2) Such further technical or design information as may be required
to complete the safety analysis, and which can reasonably be left for
later consideration, will be supplied in the FSAR;
(3) Safety features or components, if any, that require research
and development have been described by the applicant and the applicant
has identified, and there will be conducted,
[[Page 87094]]
a research and development program reasonably designed to resolve any
safety questions associated with such features or components; and
(4) On the basis of the foregoing, there is reasonable assurance of
the following--
(i) Such safety questions will be satisfactorily resolved at or
before the latest date stated in the application for completion of
construction of the proposed facility; and
(ii) Taking into consideration the site criteria contained subpart
D to this part, the proposed facility can be constructed and operated
at the proposed location without undue risk to the health and safety of
the public.
(b) A CP must contain the terms and conditions for the permit, as
the Commission deems necessary and appropriate. The Commission may, in
its discretion, incorporate in any CP provisions requiring the
applicant to furnish periodic reports of the progress and results of
research and development programs designed to resolve safety questions.
Sec. 53.1336 Finality of construction permits.
Notwithstanding any provision in Sec. 53.1590, a CP constitutes an
authorization to proceed with construction but does not constitute
Commission approval of the safety of any design feature or
specification unless the applicant specifically requests such approval
and such approval is incorporated in the permit. The applicant, at its
option, may request such approvals in the CP or by amendment to the CP.
If approved by the NRC and included in the permit, the NRC will
consider modifications to the approved design features or
specifications in accordance with Sec. 53.1590.
Sec. 53.1342 Duration of construction permits.
(a) A CP will state the earliest and latest dates for completion of
construction or alteration of the facility, not to exceed 40 years from
date of issuance.
(b) If the proposed construction or alteration of the facility is
not completed by the latest completion date, the CP shall expire, and
all rights are forfeited. However, upon good cause shown, the
Commission will extend the completion date for a reasonable period of
time. The Commission will recognize, among other things, developmental
problems attributed to the experimental nature of the facility or fire,
flood explosion, strike, sabotage, domestic violence, enemy action an
act of the elements, and other acts beyond the control of the permit
holder, as a basis for extending the completion date.
Sec. 53.1345 Transfer of construction permits.
A CP may be transferred under Sec. 53.1570.
Sec. 53.1348 Termination of construction permits.
When a permit holder has determined to permanently cease
construction, the holder must, within 30 days, submit a written
certification to the NRC.
Sec. 53.1360 Operating licenses.
Sections 53.1360 through 53.1405 set out the requirements and
procedures applicable to Commission issuance of an OL for a nuclear
power facility.
Sec. 53.1366 Contents of applications for operating licenses; general
information.
An application for an OL must include the information required by
Sec. 53.1109 and the following information:
(a) Except for an electric utility applicant, information
sufficient to demonstrate to the Commission the financial qualification
of the applicant to carry out, in accordance with the regulations in
this chapter, the activities for which the license is sought. As
applicable, the following should be provided:
(1) The applicant must submit information that demonstrates the
applicant possesses or has reasonable assurance of obtaining the funds
necessary to cover estimated operation costs for the period of the
license. The applicant must submit estimates for total annual operating
costs for each of the first 5 years of operation of the facility. The
applicant must also indicate the source(s) of funds to cover these
costs.
(2) Each application for an OL submitted by a newly-formed entity
organized for the primary purpose of operating the facility must also
include information showing--
(i) The legal and financial relationships the entity has or
proposes to have with its stockholders or owners;
(ii) The stockholders' or owners' financial ability to meet any
contractual obligation to the entity which they have incurred or
proposed to incur; and
(iii) Any other information considered necessary by the Commission
to enable it to determine the applicant's financial qualification.
(3) The Commission may request an established entity or newly
formed entity to submit additional or more detailed information
respecting its financial arrangements and status of funds if the
Commission considers this information appropriate. This may include
information regarding a licensee's ability to continue the conduct of
the activities authorized by the license and to decommission the
facility.
(b) The application must include information in the form of a
report, as described in subpart G, indicating how reasonable assurance
will be provided that funds will be available to decommission the
facility, including a copy of the financial instrument obtained to
satisfy the requirements of Sec. 53.1040.
Sec. 53.1369 Contents of applications for operating licenses;
technical information.
Final Safety Analysis Report. The application must contain an FSAR
that describes the facility and the limits on its operation and
presents a safety analysis of the SSCs of the facility as a whole. The
FSAR must include the following information, at a level of detail
sufficient to enable the Commission to reach a final conclusion on all
safety matters that must be resolved by the Commission before issuance
of an OL. The FSAR must include the following information:
(a) Site information. An application for an OL for a commercial
nuclear reactor must include the site information equivalent to that
required for an early site permit in Sec. 53.1146(a)(1)(iv) through
(x), including all current information, such as the results of
environmental and meteorological monitoring programs, which has been
developed since issuance of the CP, relating to site evaluation factors
identified in this part.
(b) Design information. Except as specified in this paragraph, an
FSAR for an OL for a commercial nuclear plant must include the final
design information equivalent to that required for a standard design
certification as defined in Sec. 53.1239(a)(2) through (7), (a)(9),
and (a)(11) through (a)(27).
(1) The completed design, including any changes during
construction, must be described.
(2) Where any design feature had not been fully developed or
demonstrated at the time of application for the CP, the applicant must
provide the analysis, research and development, test programs,
gathering of experience, or a combination thereof to provide the
required demonstration to fulfill the functional design criteria.
(c) Technical qualifications. A description of the technical
qualifications of the applicant to engage in the proposed activities in
accordance with the regulations in this chapter.
[[Page 87095]]
(d) Integrity assessment program. A description of an Integrity
Assessment Program that addresses the elements described in Sec.
53.870.
(e) Safeguards information. A description of the program to protect
Safeguards Information against unauthorized disclosure in accordance
with the requirements in Sec. Sec. 73.21 and 73.22 of this chapter, as
applicable.
(f) Emergency response facility or facilities. Description of
location and capabilities to be established for command and control,
support, and coordination of onsite and offsite, as applicable,
functions during reactor accident conditions.
(g) Role of personnel. (1) A description of the completed
assessments related to the role of personnel in ensuring safe
operations considering the analyses required by Sec. 53.730. These
assessments must include the following:
(i) Human factors engineering design requirements of Sec.
53.730(a);
(ii) Human system interface design requirements of Sec. 53.730(b);
(iii) Concept of operations of Sec. 53.730(c);
(iv) Functional requirements analysis and function allocation of
Sec. 53.730(d);
(2) A description of the program to be used for evaluating and
applying operating experience as required by Sec. 53.730(e);
(3) A staffing plan and supporting analyses as required by Sec.
53.730(f).
(h) Training, examination, and proficiency programs. (1) A
description of the training, examination, and proficiency programs
required by Sec. 53.730(g);
(2) A description of the training programs required by Sec.
53.830.
(i) Emergency plan. Emergency plans complying with the requirements
of Sec. 53.855.
(1) Include all emergency plan certifications, as applicable, that
have been obtained from the State, local, and participating Tribal
governmental agencies with emergency planning responsibilities that are
wholly or partially within the EPZ plume exposure pathway. These
certifications must state that--
(i) The proposed emergency plans are practicable;
(ii) These agencies are committed to participating in any further
development of the plans, including any required field demonstrations;
and
(iii) These agencies are committed to executing their
responsibilities under the plans in the event of an emergency.
(2) If certifications cannot be obtained after sustained, good
faith efforts by the applicant, then the application must contain
information, including a utility plan, sufficient to show that the
proposed plans provide reasonable assurance that adequate protective
measures can and will be taken in the event of a radiological emergency
at the site.
(3) If complete and integrated emergency plans were approved as
part of an early site permit, or submitted, reviewed, and approved as
part of the CP application, new certifications that demonstrate
compliance with the requirements of paragraph (i)(1) of this section
are not required.
(j) Organization. A description of the applicant's organizational
structure, allocations of responsibilities and authorities, and
personnel qualifications requirements for operation.
(k) Maintenance program. A description of a maintenance program
under Sec. 53.715.
(l) Quality assurance. A description of the QAP that demonstrates
compliance with the requirements under Sec. 53.865.
(m) Radiation protection program. A radiation protection program
description under Sec. 53.850.
(n) Security program. A physical security plan that describes how
the applicant will comply with Sec. 53.860 (and 10 CFR part 11, if
applicable, including the identification and description of jobs as
required by Sec. 11.11(a) of this chapter, at the proposed facility).
The plan must list tests, inspections, audits, and other means to be
used to demonstrate compliance with the requirements of 10 CFR parts 11
and 73, if applicable.
(o) Safeguards contingency plan. A safeguards contingency plan in
accordance with the criteria set forth in appendix C to 10 CFR part 73.
The safeguards contingency plan must include plans for dealing with
threats, thefts, and radiological sabotage, as defined in 10 CFR part
73, relating to the SNM and nuclear facilities licensed under this
chapter and in the applicant's possession and control. Each application
for this type of license must include the information contained in the
applicant's safeguards contingency plan. (Implementing procedures
required for this plan need not be submitted for approval.) \1\
(p) Security training and qualification. A training and
qualification plan that describes how the applicant will demonstrate
compliance with the criteria set forth in Sec. 73.100 of this chapter
or appendix B to 10 CFR part 73.
(q) Cybersecurity plan. A cybersecurity plan in accordance with the
criteria set forth in Sec. 73.54 or Sec. 73.110 of this chapter.
(r) Security, safeguards and cybersecurity plan implementation. A
description of the implementation of the physical security plan,
safeguards contingency plan, training and qualification plan, and
cybersecurity plan. Each applicant who prepares a physical security
plan, a safeguards contingency plan, a training and qualification plan,
or a cybersecurity plan must protect the plans and other related
Safeguards Information against unauthorized disclosure in accordance
with the requirements of Sec. Sec. 73.21 and 73.22 of this chapter.
(s) Fire protection program. A description of the fire protection
program under Sec. 53.875.
(t) Inservice inspection/inservice testing program. A description
of the inservice inspection and inservice testing programs under Sec.
53.880.
(u) [Reserved]
(v) [Reserved]
(w) General employee training. A description of the training
program required to demonstrate compliance with Sec. 53.830 and its
implementation.
(x) Fitness-for-duty program. A description of the FFD program
required by 10 CFR part 26 and its implementation.
(y) Other programs. A description and evaluation of the results of
the applicant's programs, including research and development, if any,
to demonstrate that any safety questions identified at the CP stage
have been resolved.
(z) Safety design feature performance. A description of how the
performance of each safety design feature has been demonstrated capable
of fulfilling functional design criteria considering interdependent
effects through either analysis, appropriate test programs, prototype
testing, operating experience, or a combination thereof, in accordance
with Sec. 53.440(a).
(aa) Technical specifications. Proposed technical specifications
prepared in accordance with the requirements of Sec. 53.710(a).
\1\ A physical security plan that contains all the information
required in both Sec. 73.55 or Sec. 73.100 of this chapter and
appendix C to 10 CFR part 73 satisfies the requirement for a
contingency plan.
Sec. 53.1372 Contents of applications for operating licenses; other
application content.
In addition to the FSAR, the application must also include the
following:
(a) Environmental report. An environmental report in accordance
with Sec. 51.53(b) of this chapter.
(b) Availability controls (if not included in the FSAR). A
description of
[[Page 87096]]
the controls on plant operations, including availability controls, to
provide reasonable confidence of safe operation and that the
configurations and special treatments for SR and NSRSS SSCs provide the
capabilities and reliabilities required to satisfy the safety criteria
of Sec. 53.220, or more restrictive alternative criteria adopted under
Sec. 53.470, if not addressed by Technical Specifications under Sec.
53.1369(aa).
Sec. 53.1375 Review of applications.
(a) Standards for review of applications. Applications filed under
this part will be reviewed according to the standards set out in 10 CFR
parts 20, 26, 51, 53, 73, and 140.
(b) Administrative review of applications; hearings. A proceeding
on an OL is subject to all applicable procedural requirements contained
in 10 CFR part 2, including the requirements for docketing (Sec. 2.101
of this chapter) and issuance of a notice of hearing (Sec. 2.104 of
this chapter). All hearings on OLs are governed by the procedures
contained in 10 CFR part 2.
Sec. 53.1381 Referral to the Advisory Committee on Reactor
Safeguards.
The Commission must refer a copy of the application to the ACRS.
The ACRS must report on those portions of the application that concern
safety and must apply the standards referenced in Sec. 53.1375.
Sec. 53.1384 Exemptions, departures, and variances.
(a) Applicants for an OL under this part, or any amendment to an
OL, may include in the application a request for an exemption from one
or more of the Commission's regulations. The Commission may grant an
exemption request if it determines that the exemption complies with
Sec. 53.080.
(b) An applicant for an OL who has filed an application referencing
an NRC approval, permit, license, or certification issued under this
part may include in the application a request for departures,
variances, or exemptions related to the subject referenced NRC
approval, permit, license, or certification. In determining whether to
grant the departure, variance, or exemption, the Commission must apply
the same technically relevant criteria as were applicable to the
application for the original or renewed approval, license, or
certification.
Sec. 53.1387 Issuance of operating licenses.
Upon completion of the construction or alteration of a facility, in
compliance with the terms and conditions of the construction permit and
subject to any necessary testing of the facility for health or safety
purposes, the Commission will, in the absence of good cause shown to
the contrary, issue an OL or an appropriate amendment of the license,
as the case may be.
(a)(1) After receiving the report submitted by the ACRS, the
Commission may issue an OL if the Commission finds that--
(i) Construction of the facility has been substantially completed
in conformity with the CP and the application as amended, the
provisions of the Act, and the rules and regulations of the Commission;
(ii) Any required notifications to other agencies or bodies have
been duly made;
(iii) The facility will operate in conformity with the application
as amended, the provisions of the Act, and the rules and regulations of
the Commission;
(iv) There is reasonable assurance that--
(A) the activities authorized by the OL can be conducted without
endangering the health and safety of the public; and
(B) such activities will be conducted in compliance with the
regulations in this chapter.
(v) The applicant is technically and financially qualified to
engage in the activities authorized, however, no finding of financial
qualification is necessary for an electric utility applicant for an OL;
(vi) Issuance of the license will not be inimical to the common
defense and security or to the health and safety of the public;
(vii) The applicable provisions of 10 CFR part 140 have been
satisfied; and
(viii) The findings required by subpart A of 10 CFR part 51 have
been made.
(2) [Reserved]
(b) [Reserved]
(c) The OL will include appropriate provisions with respect to any
uncompleted items of construction and such limitations or conditions as
are required to assure that operation during the period of the
completion of such items will not endanger public health and safety.
(d) The Commission will issue an OL in such form and containing
such conditions and limitations, including technical specifications, as
it deems necessary and appropriate.
Sec. 53.1390 Backfitting of operating licenses.
After issuance of an OL, the Commission may not modify, add, or
delete any term or condition of the OL, except in accordance with the
provisions of Sec. 53.1590.
Sec. 53.1396 Duration of operating licenses.
The Commission will issue an OL under this part for the term
requested by the applicant, not to exceed 40 years from the date of
issuance, or for the estimated useful life of the facility if the
Commission determines that the estimated useful life is less than the
term requested.
Sec. 53.1399 Transfer of an operating license.
An OL may be transferred under Sec. 53.1570.
Sec. 53.1402 Application for renewal.
The filing of an application for a renewed license must be in
accordance with Sec. 53.1595.
Sec. 53.1405 Continuation of an operating license.
Each OL for a facility that has permanently ceased operations
continues in effect beyond the expiration date to authorize ownership
and possession of the facility until the Commission notifies the
licensee in writing that the license is terminated. During this period
of continued effectiveness, the licensee must--
(a) Take actions necessary to decommission and decontaminate the
facility and continue to maintain the facility, including, where
applicable, the storage, control, and maintenance of the spent fuel in
a safe condition; and
(b) Conduct activities in accordance with all other restrictions
applicable to the facility in accordance with the NRC's regulations and
the provisions of the OL for the facility.
Sec. 53.1410 Combined licenses.
Sections 53.1410 through 53.1461 set out the requirements and
procedures applicable to Commission issuance of COLs for commercial
nuclear plants under this part.
Sec. 53.1413 Contents of applications for combined licenses; general
information.
An application for a COL must include the information required by
Sec. 53.1109 and the following information:
(a) Except for an electric utility applicant in regard to financial
assurance required after a Commission finding under Sec. 53.1452, the
application must include information sufficient to demonstrate to the
Commission the financial qualification of the applicant to carry out,
in accordance with the regulations in this chapter, the activities for
which the permit or license is sought. As applicable, the following
should be provided:
(1) The applicant must submit information that demonstrates that
the applicant possesses or has reasonable
[[Page 87097]]
assurance of obtaining the funds necessary to cover estimated
construction costs and related fuel cycle costs. The applicant must
submit estimates of the total construction costs of the facility and
related fuel cycle costs and must indicate the source(s) of funds to
cover these costs.
(2) The applicant must submit information that demonstrates the
applicant possesses or has reasonable assurance of obtaining the funds
necessary to cover estimated operation costs for the period of the
license. The applicant must submit estimates for total annual operating
costs for each of the first 5 years of operation of the facility. The
applicant must also indicate the source(s) of funds to cover these
costs.
(3) Each application for a COL submitted by a newly-formed entity
organized for the primary purpose of constructing and operating a
facility must also include information showing--
(i) The legal and financial relationships the entity has or
proposes to have with its stockholders or owners; and
(ii) The stockholders' or owners' financial ability to meet any
contractual obligation to the entity which they have incurred or
proposed to incur; and
(iii) Any other information considered necessary by the Commission
to enable it to determine the applicant's financial qualification.
(4) The Commission may request an established entity or newly
formed entity to submit additional or more detailed information
respecting its financial arrangements and status of funds if the
Commission considers this information appropriate. This may include
information regarding a licensee's ability to continue the conduct of
the activities authorized by the license and to decommission the
facility.
(b) The application must include information in the form of a
report, as described in subpart G of this part, indicating how
reasonable assurance will be provided that funds will be available to
decommission the facility.
Sec. 53.1416 Contents of applications for combined licenses;
technical information.
(a) Final Safety Analysis Report. The application must contain an
FSAR that describes the facility and the limits on its operation and
presents a safety analysis of the SSCs of the facility as a whole. The
Commission will require, before issuance of a COL, that engineering
documents, such as analyses, drawings, procurement specifications, or
construction and installation specifications, be completed and
available for audit if the more detailed information is necessary for
the Commission to verify the information in the application and make
its safety determination. The FSAR must include the following
information, at a level of detail sufficient to enable the Commission
to reach a final conclusion on all safety matters that must be resolved
by the Commission before issuance of a COL:
(1) Site information. An application for a COL for a commercial
nuclear reactor must include the site information required for an early
site permit in Sec. 53.1146(a)(1)(iv) through (x).
(2) Design information. An application for a COL for a commercial
nuclear plant must include the design information equivalent to that
required for a standard design certification as defined in Sec.
53.1239(a)(2) through (7), (a)(9), and (a)(11) through (27).
(3) Technical qualifications. A description of the technical
qualifications of the applicant to engage in the proposed activities in
accordance with the regulations in this chapter.
(4) Integrity assessment program. A description of an Integrity
Assessment Program that addresses the elements described in Sec.
53.870.
(5) Safeguards information. A description of the program to protect
Safeguards Information against unauthorized disclosure in accordance
with the requirements in Sec. Sec. 73.21 and 73.22 of this chapter, as
applicable.
(6) Emergency response facility or facilities. Description of the
locations and capabilities to be established for command and control,
support, and coordination of onsite and offsite, as applicable,
functions during reactor accident conditions.
(7) Role of personnel. (i) A description of the completed
assessments related to the role of personnel in ensuring safe
operations considering the analyses required by Sec. 53.730. These
assessments must include the following:
(A) Human factors engineering design requirements of Sec.
53.730(a);
(B) Human system interface design requirements of Sec. 53.730(b);
(C) Concept of operations of Sec. 53.730(c); and
(D) Functional requirements analysis and function allocation of
Sec. 53.730(d);
(ii) A description of the program to be used for evaluating and
applying operating experience as required by Sec. 53.730(e);
(iii) A staffing plan and supporting analyses as required by Sec.
53.730(f).
(8) Training, examination, and proficiency programs. (i) A
description of the training, examination, and proficiency programs
required by Sec. 53.730(g); and
(ii) A description of the training programs required by Sec.
53.830.
(9) Emergency plan. Emergency plans complying with the requirements
of Sec. 53.855.
(i) The emergency plan must include, as applicable, all emergency
plan certifications that have been obtained from the State, local, and
participating Tribal governmental agencies with emergency planning
responsibilities. The certifications must state that--
(A) The proposed emergency plans are practicable;
(B) These agencies are committed to participating in any further
development of the plans, including any required field demonstrations;
and
(C) These agencies are committed to executing their
responsibilities under the plans in the event of an emergency.
(ii) If certifications cannot be obtained after sustained, good
faith efforts by the applicant, then the application must contain
information, including a utility plan, sufficient to show that the
proposed plans provide reasonable assurance that adequate protective
measures can and will be taken in the event of a radiological emergency
at the site.
(10) Organization. A description of the applicant's organizational
structure, allocations of responsibilities and authorities, and
personnel qualifications requirements for operation.
(11) Maintenance program. A description of a maintenance program
under Sec. 53.715.
(12) Quality assurance. A description of the QAP under Sec.
53.865.
(13) Radiation protection program. A radiation protection program
description under Sec. 53.850.
(14) Security program. A physical security plan that describes how
the applicant will comply with Sec. 53.860 (and 10 CFR part 11, if
applicable, including the identification and description of jobs as
required by Sec. 11.11(a) of this chapter, at the proposed facility).
The plan must list tests, inspections, audits, and other means to be
used to demonstrate compliance with the requirements of 10 CFR parts 11
and 73, if applicable.
(15) Safeguards contingency plan. A safeguards contingency plan in
accordance with the criteria set forth in appendix C to 10 CFR part 73.
The safeguards contingency plan must include plans for dealing with
threats, thefts, and radiological sabotage, as defined in 10 CFR part
73, relating to the SNM and nuclear facilities licensed under this
chapter and in the applicant's possession and control. Each
[[Page 87098]]
application for this type of license must include the information
contained in the applicant's safeguards contingency plan.\1\
(Implementing procedures required for this plan need not be submitted
for approval.)
(16) Security training and qualification. A training and
qualification plan that describes how the applicant will demonstrate
compliance with the criteria set forth in Sec. 73.100 of this chapter
or appendix B to 10 CFR part 73.
(17) Cybersecurity plan. A cybersecurity plan in accordance with
the criteria set forth in Sec. 73.54 or Sec. 73.110 of this chapter.
(18) Security, safeguards and cybersecurity plan implementation. A
description of the implementation of the physical security plan,
safeguards contingency plan, training and qualification plan, and
cybersecurity plan. Each applicant who prepares a physical security
plan, a safeguards contingency plan, a training and qualification plan,
or a cybersecurity plan must protect the plans and other related
Safeguards Information against unauthorized disclosure in accordance
with the requirements of Sec. Sec. 73.21 and 73.22 of this chapter.
(19) Fire protection program. A description of the fire protection
program under Sec. 53.875.
(20) Inservice inspection/inservice testing program. Descriptions
of inservice inspection and inservice testing programs under Sec.
53.880.
(21) [Reserved]
(22) [Reserved]
(23) General employee training. A description of the training
program required to demonstrate compliance with Sec. 53.830 and its
implementation.
(24) Fitness-for-duty program. A description of the FFD program
under part 26 of this chapter and its implementation.
(25) Technical specifications. Proposed technical specifications
prepared in accordance with the requirements of Sec. 53.710(a).
(b) If there are SSCs of the plant for which research and
development is necessary to confirm the adequacy of their design, a
report which documents the resolution of any safety questions
associated with such SSCs.
(c) A description of how the performance of each safety design
feature has been demonstrated capable of fulfilling functional design
criteria considering interdependent effects through either analysis,
appropriate test programs, prototype testing, operating experience, or
a combination thereof, in accordance with Sec. 53.440(a).
(d) If the COL application references an early site permit, then
the following requirements apply:
(1) The FSAR need not contain information or analyses submitted to
the Commission in connection with the early site permit provided that
the FSAR must either include or incorporate by reference the early site
permit Site Safety Analysis Report and contain, in addition to the
information and analyses otherwise required, information sufficient to
demonstrate that the design of the facility falls within the site
characteristics and design parameters specified in the early site
permit.
(2) If the FSAR does not demonstrate that design of the facility
falls within the site characteristics and design parameters, the
application must include a request for a variance that complies with
the requirements of Sec. Sec. 53.1188(d) and 53.1437.
(3) The FSAR must demonstrate that all terms and conditions that
have been included in the early site permit will be satisfied by the
date of issuance of the COL. Any terms or conditions of the early site
permit that could not be met by the time of issuance of the COL must be
set forth as terms or conditions of the COL.
(4) If the early site permit approves complete and integrated
emergency plans, or major features of emergency plans, then the FSAR
must include any new or additional information that updates and
corrects the information that was provided under Sec. 53.1146(b)(2)
and discuss whether the new or additional information materially
changes the bases for compliance with the applicable requirements. The
application must identify changes to the emergency plans or major
features of emergency plans that have been incorporated into the
proposed facility emergency plans and that constitute or would
constitute a change in an emergency plan that results in reducing the
licensee's capability to perform an emergency planning function in the
event of a radiological emergency.
(5) If complete and integrated emergency plans are approved as part
of the early site permit, new certifications meeting the requirements
of paragraph (a)(9)(i) of this section are not required.
(e) If the COL application references a standard design approval,
then the following requirements apply:
(1) The FSAR need not contain information or analyses submitted to
the Commission in connection with the design approval, provided,
however, that the FSAR must either include or incorporate by reference
the standard design approval FSAR and must contain, in addition to the
information and analyses otherwise required, information sufficient to
demonstrate that the characteristics of the site fall within the site
parameters specified in the design approval. In addition, the plant-
specific PRA information must use the PRA information for the design
approval and must be updated to account for site specific design
information and any design changes or departures.
(2) The FSAR must demonstrate that all terms and conditions that
have been included in the design approval will be satisfied by the date
of issuance of the COL.
(f) If the COL application references a standard design
certification, then the following requirements apply:
(1) The FSAR need not contain information or analyses submitted to
the Commission in connection with the standard design certification,
provided, however, that the FSAR must either include or incorporate by
reference the standard design certification FSAR and must contain, in
addition to the information and analyses otherwise required,
information sufficient to demonstrate that the site characteristics
fall within the site parameters specified in the standard design
certification. In addition, the plant-specific PRA information must use
the PRA information for the standard design certification and must be
updated to account for site-specific design information and any design
changes or departures.
(2) The FSAR must demonstrate that the interface requirements
established for the design under Sec. 53.1239(a)(24) have been met.
(3) The FSAR must demonstrate that all requirements and
restrictions set forth in the referenced standard design certification
rule must be satisfied by the date of issuance of the COL. Any
requirements and restrictions set forth in the referenced standard
design certification rule that could not be satisfied by the time of
issuance of the COL, must be set forth as terms or conditions of the
COL.
(g) If the COL application references the use of one or more
manufactured reactors licensed under Sec. 53.1270, then the following
requirements apply:
(1) The FSAR need not contain information or analyses submitted to
the Commission in connection with the ML, provided, however, that the
FSAR must either include or incorporate by reference the ML FSAR and
must contain, in addition to the information and analyses otherwise
required, information sufficient to demonstrate that the site
characteristics fall within the site parameters specified in the ML.
[[Page 87099]]
In addition, the plant-specific PRA information must use the PRA
information for the manufactured reactor and must be updated to account
for site-specific design information and any design changes or
departures.
(2) The FSAR must demonstrate that the interface requirements
established for the design have been met.
(3) The FSAR must demonstrate that all terms and conditions that
have been included in the ML will be satisfied by the date of issuance
of the COL. Any terms or conditions of the ML that could not be met by
the time of issuance of the COL, must be set forth as terms or
conditions of the COL.
(h) Each applicant for a COL under this part must protect
Safeguards Information against unauthorized disclosure in accordance
with the requirements in Sec. Sec. 73.21 and 73.22 of this chapter, as
applicable.
\1\ A physical security plan that contains all the information
required in both Sec. 73.55 or Sec. 73.100 of this chapter and
appendix C to 10 CFR part 73 demonstrates compliance with the
requirement for a contingency plan.
Sec. 53.1419 Contents of applications for combined licenses; other
application content.
(a) In addition to the FSAR, the application must also include the
following:
(1) Environmental report. (i) An environmental report either in
accordance with Sec. 51.50(c) of this chapter if an LWA under Sec.
53.1130 is not requested in conjunction with the COL application, or in
accordance with Sec. Sec. 51.49 and 51.50(c) of this chapter if an LWA
is requested in conjunction with the COL application; or
(ii) If the applicant wishes to request that an LWA under Sec.
53.1130 be issued before issuance of the COL, the information otherwise
required by Sec. 53.1130, in accordance with either Sec. 2.101(a)(1)
through (a)(4), or Sec. 2.101(a)(9) of this chapter;
(2) Availability controls (if not included in the FSAR). A
description of the controls on plant operations, including availability
controls, to provide reasonable confidence of safe operation and that
the configurations and special treatments for SR SSCs and NSRSS SSCs
provide the capabilities and reliabilities required to satisfy the
safety criteria of Sec. 53.220, or more restrictive alternative
criteria adopted under Sec. 53.470, if not addressed by Technical
Specifications under Sec. 53.1416(a)(25); and
(3) Inspections, tests, analyses, and acceptance criteria. The
proposed inspections, tests, and analyses, including those applicable
to emergency planning, that the licensee must perform, and the
acceptance criteria that are necessary and sufficient to provide
reasonable assurance that, if the inspections, tests, and analyses are
performed and the acceptance criteria met, the facility has been
constructed and will be operated in conformity with the COL, the
provisions of the Act, and the Commission's rules and regulations.
(i) If the application references an early site permit with ITAAC,
the early site permit ITAAC must apply to those aspects of the COL
which are approved in the early site permit.
(ii) If the application references a standard design certification,
the ITAAC contained in the certified design must apply to those
portions of the facility design which are approved in the standard
design certification.
(iii) If the application references an ML, the ITAAC contained in
the ML must apply to those portions of the facility design which are
approved in the ML.
(iv) If the application references an early site permit with ITAAC,
a standard design certification, an ML, or combination thereof, the
application may include a notification that a required inspection,
test, or analysis in the ITAAC has been successfully completed and that
the corresponding acceptance criterion has been met. The Federal
Register notification required by Sec. 53.1422 of this chapter must
indicate that the application includes this notification.
(b) [Reserved]
Sec. 53.1422 Review of applications.
(a) Standards for review of applications. Applications filed under
this part will be reviewed according to the standards set out in this
part and 10 CFR parts 20, 51, 73, and 140.
(b) Administrative review of applications; hearings. A proceeding
on a COL is subject to all applicable procedural requirements contained
in 10 CFR part 2, including the requirements for docketing (Sec. 2.101
of this chapter) and issuance of a notice of hearing (Sec. 2.104 of
this chapter). If an applicant requests a Commission finding on certain
ITAAC with the issuance of the COL, then those ITAAC will be identified
in the notice of hearing. All hearings on COLs are governed by the
procedures contained in 10 CFR part 2.
Sec. 53.1425 Finality of referenced NRC approvals.
If the application for a COL under this part references an early
site permit, standard design certification rule, standard design
approval, or ML, issued under this part, the scope and nature of
matters resolved for the application and any COL issued are governed by
the relevant provisions addressing finality, including Sec. Sec.
53.1188, 53.1221, 53.1263, and 53.1288.
Sec. 53.1431 Referral to the Advisory Committee on Reactor
Safeguards.
The Commission must refer a copy of the application to the ACRS.
The ACRS must report on those portions of the application that concern
safety and must apply the standards referenced in Sec. 53.1422, in
accordance with the finality provisions in Sec. 53.1425.
Sec. 53.1434 Authorization to conduct limited work authorization
activities.
(a) If the application for a COL under this part does not reference
an early site permit which authorizes the holder to perform the
activities under Sec. 53.1130(b), the applicant may not perform those
activities without obtaining the separate authorization required by
Sec. 53.1130(a). Authorization may be granted only after the presiding
officer in the proceeding on the application has made the findings and
determination required by Sec. 53.1130(b)(1)(ii) and (b)(1)(iv), and
the Director, Office of Nuclear Reactor Regulation makes the
determination required by Sec. 53.1130(b)(1)(iii).
(b) If, after an applicant has performed the activities permitted
by a LWA issued under Sec. 53.1130, the application for the COL is
withdrawn or denied, then the applicant must implement the approved
site redress plan.
Sec. 53.1437 Exemptions, departures, and variances.
(a) An applicant for a COL, or any amendment to a COL, may include
in the application a request for an exemption from one or more of the
Commission's regulations.
(1) If the request is for an exemption from any part of a
referenced standard design certification rule, the Commission may grant
the request if it determines that the exemption complies with any
exemption provisions of the referenced standard design certification
rule, or with Sec. 53.1263 if there are no applicable exemption
provisions in the referenced standard design certification rule.
(2) For all other requests for exemptions, the Commission may grant
a request if it determines that the exemption complies with Sec.
53.080.
(b) An applicant for a COL who has filed an application referencing
an early site permit issued under Sec. 53.1158 may include in the
application a request for a variance from one or more site
characteristics, design parameters, or
[[Page 87100]]
terms and conditions of the permit, or from the Site Safety Analysis
Report. In determining whether to grant the variance, the Commission
must apply the same technically relevant criteria as were applicable to
the application for the original or renewed site permit. Once a COL
referencing an early site permit is issued, variances from the early
site permit will not be granted for that CP or COL.
(c) An applicant for a COL who has filed an application referencing
use of a manufactured reactor may include in the application a request
for a departure from one or more design characteristics, site
parameters, terms and conditions, or approved design of the
manufactured reactor under the ML issued under Sec. 53.1287. The
Commission may grant such a request only if it determines that the
departure will comply with the requirements of Sec. 53.080, and that
the special circumstances outweigh any decrease in safety that may
result from the reduction in standardization caused by the departure.
(d) Issuance of a variance under paragraph (b) of this section or a
departure under paragraph (c) of this section is subject to litigation
during the COL proceeding in the same manner as other issues material
to that proceeding.
Sec. 53.1440 Issuance of combined licenses.
(a)(1) After conducting a hearing under Sec. 53.1422(b) and
receiving the report submitted by the ACRS, the Commission may issue a
COL if the Commission finds that--
(i) The applicable standards and requirements of the Act and the
Commission's regulations have been met;
(ii) Any required notifications to other agencies or bodies have
been duly made;
(iii) There is reasonable assurance that the facility will be
constructed and will operate in conformity with the license, the
provisions of the Act, and the Commission's regulations;
(iv) The applicant is technically and financially qualified to
engage in the activities authorized; however, no finding of financial
qualification is necessary for an electric utility applicant for a COL;
(v) Issuance of the license will not be inimical to the common
defense and security or to the health and safety of the public; and
(vi) The findings required by subpart A of 10 CFR part 51 have been
made.
(2) The Commission may also find, at the time it issues the COL,
that certain acceptance criteria in one or more of the ITAAC in a
referenced early site permit, standard design certification, or ML have
been met. This finding will finally resolve that those acceptance
criteria have been met, those acceptance criteria will be deemed to be
excluded from the COL, and findings under Sec. 53.1452(g) with respect
to those acceptance criteria are unnecessary.
(b) The Commission must identify within the COL the inspections,
tests, and analyses, including those applicable to emergency planning,
that the licensee must perform, and the acceptance criteria that, if
met, are necessary and sufficient to provide reasonable assurance that
the facility has been constructed and will be operated in conformity
with the license, the provisions of the Act, and the Commission's rules
and regulations.
(c) A COL must contain the terms and conditions, including
technical specifications, as the Commission deems necessary and
appropriate.
Sec. 53.1443 Finality of combined licenses.
(a) After issuance of a COL, the Commission may not modify, add, or
delete any term or condition of the COL, the design of the facility,
the ITAAC contained in the license that are not derived from a
referenced standard design certification or ML, except under the
provisions of Sec. 53.1452 or Sec. 53.1590.
(b) If the COL does not reference a standard design certification
or use of a manufactured reactor under an ML issued under Sec.
53.1287, then a licensee may make changes in the facility as described
in the FSAR (as updated) and make changes in the procedures as
described in the FSAR (as updated) under the applicable change
processes in Sec. 53.1550.
(c) If the COL references a certified design, then--
(1) Changes to or departures from information within the scope of
the referenced standard design certification rule are subject to the
applicable change processes in that rule; and
(2) Changes that are not within the scope of the referenced
standard design certification rule are subject to the applicable change
processes in subpart I of this part, unless they also involve changes
to or noncompliance with information within the scope of the referenced
standard design certification rule. In these cases, the applicable
provisions of this section and the standard design certification rule
apply.
(d) If the COL references use of a manufactured reactor under an ML
issued under this part, then--
(1) Changes to or departures from information within the scope of
the manufactured reactor's design are subject to the change processes
in Sec. 53.1288; and
(2) Changes that are not within the scope of the manufactured
reactor's design are subject to the applicable change processes in
subpart I.
(e) The Commission may issue and make immediately effective any
amendment to a COL upon a determination by the Commission that the
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person. The amendment may be issued and made
immediately effective in advance of the holding and completion of any
required hearing. The amendment will be processed under the procedures
specified in Sec. 53.1515.
(f) Any modification to, addition to, or deletion from the terms
and conditions of a COL, including any modification to, addition to, or
deletion from the inspections, tests, and analyses, or related
acceptance criteria contained in the license is a proposed amendment to
the license. There must be an opportunity for a hearing on the
amendment.
Sec. 53.1449 Inspection during construction.
(a) Licensee schedule for inspections, tests, or analyses. The
licensee must submit to the NRC, no later than 1 year after issuance of
the COL or at the start of construction as defined at Sec. 53.020,
whichever is later, its schedule for completing the inspections, tests,
or analyses in the ITAAC. The licensee must submit updates to the ITAAC
schedules every 6 months thereafter and, within 1 year of its scheduled
date for initial loading of fuel (or, for a fueled manufactured
reactor, within 1 year of its scheduled date for initiating the
physical removal of any one of the independent physical mechanisms to
prevent criticality required under Sec. 53.620(d)(1)), the licensee
must submit updates to the ITAAC schedule every 30 days until the final
notification is provided to the NRC under paragraph (c)(1) of this
section.
(b) Licensee and applicant conduct of activities subject to ITAAC.
With respect to activities subject to an ITAAC, an applicant for a COL
may proceed at its own risk with design and procurement activities, and
a licensee may proceed at its own risk with design, procurement,
construction, and preoperational activities, even though the NRC may
not have found that any one of the prescribed acceptance criteria are
met.
(c) Licensee notifications. (1) ITAAC closure notification. The
licensee must notify the NRC that prescribed inspections, tests, and
analyses have been performed and that the prescribed acceptance
criteria are met. The
[[Page 87101]]
notification must contain sufficient information to demonstrate that
the prescribed inspections, test, and analyses have been performed and
that the prescribed acceptance criteria are met.
(2) ITAAC post-closure notifications. Following the licensee's
ITAAC closure notifications under paragraph (c)(1) of this section
until the Commission makes the finding under Sec. 53.1452(g), the
licensee must notify the NRC, in a timely manner, of new information
that materially alters the basis for determining that either
inspections, tests, or analyses were performed as required, or that
acceptance criteria are met. The notification must contain sufficient
information to demonstrate that, notwithstanding the new information,
the prescribed inspections, tests, and analyses have been performed as
required, and the prescribed acceptance criteria are met.
(3) Uncompleted ITAAC notification. If the licensee has not
provided, by the date 225 days before the scheduled date for initial
loading of fuel (or, for a fueled manufactured reactor, by the date 225
days before the scheduled date for initiating the physical removal of
any one of the independent physical mechanisms to prevent criticality
required under Sec. 53.620(d)(1)), the notification required by
paragraph (c)(1) of this section for all ITAAC, then the licensee must
notify the NRC that the prescribed inspections, tests, or analyses for
all uncompleted ITAAC will be performed and that the prescribed
acceptance criteria will be met prior to operation. The notification
must be provided no later than the date 225 days before the scheduled
date for initial loading of fuel (or, for a fueled manufactured
reactor, no later than the date 225 days before the scheduled date for
initiating the physical removal of any one of the independent physical
mechanisms to prevent criticality required under Sec. 53.620(d)(1)),
and must provide sufficient information to demonstrate that the
prescribed inspections, tests, or analyses will be performed and the
prescribed acceptance criteria for the uncompleted ITAAC will be met,
including, but not limited to, a description of the specific procedures
and analytical methods to be used for performing the prescribed
inspections, tests, and analyses and determining that the prescribed
acceptance criteria are met.
(4) All ITAAC complete notification. The licensee must notify the
NRC that all ITAAC are complete.
(d) Licensee determination of noncompliance with ITAAC. (1) In the
event that an activity is subject to an ITAAC derived from a referenced
standard design certification and the licensee has not demonstrated
that the prescribed acceptance criteria are met, the licensee may take
corrective actions to successfully complete that ITAAC or request an
exemption from the standard design certification ITAAC, as applicable.
A request for an exemption must also be accompanied by a request for a
license amendment under subpart I.
(2) In the event that an activity is subject to an ITAAC not
derived from a referenced standard design certification and the
licensee has not demonstrated that the prescribed acceptance criteria
are met, the licensee may take corrective actions to successfully
complete that ITAAC or request a license amendment under subpart I.
(e) NRC inspection, publication of notices, and availability of
licensee notifications. The NRC must ensure that the prescribed
inspections, tests, and analyses in the ITAAC are performed.
(1) At appropriate intervals until the last date for submission of
requests for hearing under Sec. 53.1452, the NRC must publish notices
in the Federal Register of the NRC staff's determination of the
successful completion of inspections, tests, and analyses.
(2) The NRC must make publicly available the licensee notifications
under paragraph (c) of this section. The NRC must, no later than the
date of publication of the notice of intended operation required by
Sec. 53.1452(a), make publicly available those licensee notifications
under paragraph (c) of this section that have been submitted to the NRC
at least 7 days before that notice.
Sec. 53.1452 Operation under a combined license.
(a) The licensee must notify the NRC of its scheduled date for
initial loading of fuel no later than 270 days before the scheduled
date and must notify the NRC of updates to its schedule every 30 days
thereafter.\1\ Not less than 180 days before the date scheduled for
initial loading of fuel into a plant by a licensee that has been issued
a COL under this part, the Commission must publish notice of intended
operation in the Federal Register.\2\ The notice must provide that any
person whose interest may be affected by operation of the plant may,
within 60 days, request that the Commission hold a hearing on whether
the facility as constructed complies, or on completion will comply,
with the acceptance criteria in the COL, except that a hearing must not
be granted for those ITAAC that the Commission found were met under
Sec. 53.1440(a)(2).
(b) A request for hearing under paragraph (a) of this section must
show, prima facie--
(1) That one or more of the acceptance criteria of the ITAAC in the
COL have not been, or will not be, met; and
(2) The specific operational consequences of nonconformance that
would be contrary to providing reasonable assurance of adequate
protection of the public health and safety.
(c) The Commission, acting as the presiding officer, must determine
whether to grant or deny the request for hearing under the applicable
requirements of Sec. 2.309 of this chapter. If the Commission grants
the request, the Commission, acting as the presiding officer, must
determine whether during a period of interim operation there will be
reasonable assurance of adequate protection to the public health and
safety. The Commission's determination must consider the petitioner's
prima facie showing and any answers thereto. If the Commission
determines there is such reasonable assurance, it must allow operation
during an interim period under the COL.
(d) The Commission, in its discretion, must determine appropriate
hearing procedures, whether informal or formal adjudicatory, for any
hearing under paragraph (a) of this section, and must state its reasons
therefore.
(e) The Commission must, to the maximum possible extent, render a
decision on issues raised by the hearing request within 180 days of the
publication of the notice provided by paragraph (a) of this section or
by the anticipated date for initial loading of fuel into the reactor,
whichever is later.
(f) A petition to modify the terms and conditions of the COL will
be processed as a request for action under Sec. 2.206 of this chapter.
The petitioner must file the petition with the Secretary of the
Commission. Before the licensed activity allegedly affected by the
petition (fuel loading, low power testing, etc.) commences, the
Commission must determine whether any immediate action is required. If
the petition is granted, then an appropriate order will be issued. Fuel
loading and operation under the COL will not be affected by the
granting of the petition unless the order is made immediately
effective.
(g) The licensee must not operate the facility until the Commission
makes a finding that the acceptance criteria in the COL are met, except
for those acceptance criteria that the Commission found were met under
Sec. 53.1440(a)(2). If the COL is for a modular design, each
[[Page 87102]]
reactor unit may require a separate finding as construction proceeds.
(h) After the Commission has made the finding in paragraph (g) of
this section, the ITAAC do not, by virtue of their inclusion in the
COL, constitute regulatory requirements either for licensees or for
renewal of the license; except for the specific ITAAC for which the
Commission has granted a hearing under paragraph (a) of this section,
all ITAAC expire upon final Commission action in the proceeding.
However, subsequent changes to the facility or procedures described in
the FSAR (as updated) must comply with the requirements in Sec.
53.1443(e) or (f), as applicable.
\1\ For licensees installing fueled manufactured reactors under
a COL, the COL holder must instead notify the NRC of its scheduled
date for initiating the physical removal of any one of the
independent physical mechanisms to prevent criticality required
under Sec. 53.620(d)(1) no later than 270 days before the scheduled
date and must notify the NRC of updates to its schedule every 30
days thereafter.
\2\ For licensees installing fueled manufactured reactors under
a COL, the Commission must instead publish notice of intended
operation in the Federal Register not less than 180 days before the
date scheduled for initiating the physical removal of any one of the
independent physical mechanisms to prevent criticality required
under Sec. 53.620(d)(1).
Sec. 53.1455 Duration of combined license.
A COL is issued for a specified period not to exceed 40 years from
the date on which the Commission makes a finding that acceptance
criteria are met under Sec. 53.1452(g) or allowing operation during an
interim period under the COL under Sec. 53.1452(c).
Sec. 53.1456 Transfer of a combined license.
A COL may be transferred under Sec. 53.1570.
Sec. 53.1458 Application for renewal.
The filing of an application for a renewed license must be in
accordance with Sec. 53.1595.
Sec. 53.1461 Continuation of combined license.
Each COL for a facility that has permanently ceased operations
continues in effect beyond the expiration date to authorize ownership
and possession of the facility until the Commission notifies the
licensee in writing that the license is terminated. During this period
of continued effectiveness, the licensee must--
(a) Take actions necessary to decommission and decontaminate the
facility and continue to maintain the facility, including, where
applicable, the storage, control and maintenance of the spent fuel, in
a safe condition; and
(b) Conduct activities in accordance with all other restrictions
applicable to the facility in accordance with the NRC's regulations and
the provisions of the COL for the facility.
Sec. 53.1470 Standardization of commercial nuclear plant designs:
licenses to construct and operate nuclear power reactors of identical
design at multiple sites.
(a) Except as otherwise specified in this section, the provisions
of this section apply to CP, OL, and COL applications for commercial
nuclear plants of identical design (the ``common design'') under this
part.
(b) Each application for a CP, OL, or COL submitted pursuant to
this section must be submitted as specified in Sec. Sec. 53.1300,
53.1360, or 53.1410, respectively, and Sec. 2.101 of this chapter.
Each application must state that the applicant wishes to construct a
facility identical to a facility proposed for one or more sites other
than the applicant's (the ``common design''), and the applicant wishes
to have the application considered under this section. Each application
must list each of the other applications to be treated together under
this section.
(c) Each application must include the information required by the
applicable sections of this subpart, provided however, that the
application must identify the common design, and, if applicable,
reference a standard design certification or standard design approval
under this part, or the use of a reactor manufactured under this part.
The FSAR for each application must either incorporate by reference or
include the final safety analysis of the common design, including, if
applicable, the FSAR for the referenced standard design certification,
standard design approval, or the manufactured reactor.
(d) Each application submitted pursuant to this section must
contain an environmental report under Sec. Sec. 53.1312(a)(1),
53.1372(a), or 53.1419(a)(1), as applicable, that complies with the
applicable provisions of 10 CFR part 51, provided, however, that the
application may incorporate by reference a single environmental report
on the environmental impacts of the common design that are applicable
to each site.
(e) Upon a determination that each application is acceptable for
docketing under Sec. 2.101 of this chapter, each application will be
docketed and a notice of docketing for each application will be
published in the Federal Register, under Sec. 2.104 of this chapter,
provided, however, that the notice must state that the application will
be processed under the provisions of this section and subpart D of 10
CFR part 2. At the discretion of the Commission, a single notice of
docketing for multiple applications may be published in the Federal
Register.
(f) The NRC must prepare an environmental assessment or draft and
final environmental impact statements for each of the applications
under 10 CFR part 51. Scoping under Sec. Sec. 51.28 and 51.29 of this
chapter for each of the license applications may be conducted
simultaneously and joint scoping may be conducted with respect to the
environmental issues relevant to the common design. If the applications
reference a standard design certification, then the environmental
assessment or environmental impact statement for each of the
applications must incorporate by reference the standard design
certification environmental assessment. If the applications do not
reference a standard design certification, then the NRC must prepare
environmental assessments or draft and final supplemental environmental
impact statements which address severe accident mitigation design
alternatives for the common design, which must be incorporated by
reference into the environmental assessment or environmental impact
statement prepared for each application. Scoping under Sec. Sec. 51.28
and 51.29 of this chapter for the supplemental environmental impact
statement may be conducted simultaneously and may be part of the
scoping for each of the applications.
(g) The ACRS must report on each of the applications as required by
the applicable sections of this subpart. Each report must be limited to
those safety matters for each application that are not relevant to the
common design. In addition, the ACRS must separately report on the
safety of the common design, provided, however, that the report need
not address the safety of a referenced standard design certification or
reactor manufactured under this part.
(h) The Commission must designate a presiding officer to conduct
the proceeding with respect to the health and safety, common defense
and security, and environmental matters relating to the common design
and affecting at least two applications. The hearing will be governed
by the applicable provisions of subparts A, C, G, L, N, and O of 10 CFR
part 2 relating to applications for CPs, OLs, and COLs. The presiding
officer must issue a partial initial decision on the common design.
(i) If the design for the power reactor(s) proposed in a particular
[[Page 87103]]
application is not identical to the others, that application may not be
processed under this section and subpart D of 10 CFR part 2.
(j) As used in this section, the design of a nuclear power reactor
included in a single referenced Safety Analysis Report means the design
of those SSCs important to radiological health and safety and the
common defense and security.
Subpart I--Maintaining and Revising Licensing-Basis Information
Sec. 53.1500 Licensing-basis information.
This subpart provides the requirements for each holder of a license
for a commercial nuclear plant licensed under this part to maintain
licensing-basis information as defined in Sec. 53.020; evaluate
changes to site characteristics, plant design features, and
programmatic controls to determine needed approvals and revisions; and
submit appropriate updates to the U.S. Nuclear Regulatory Commission
(NRC).
Sec. 53.1502 Specific terms and conditions of licenses.
(a) Each license issued under this part is subject to the
provisions of the Atomic Energy Act of 1954, as amended, (the Act) and
to all rules, regulations, and orders of the Commission. The terms and
conditions of the license will be subject to amendment, revision, or
modification, by reason of amendments of the Act or by reason of rules,
regulations, and orders issued in accordance with the terms of the Act.
(b) Each license issued under this part must be subject to all
conditions imposed as a matter of law by sections 401(a)(2) and 401(d)
of the Federal Water Pollution Control Act, as amended (33 U.S.C.A.
1341(a)(2) and (d)).
(c) A holder of an operating license (OL) or combined license (COL)
under this part may take reasonable action that departs from a license
condition or a technical specification included in a license issued
under this part in a national security emergency established by a law
enacted by the Congress or by an order or directive issued by the
President pursuant to statutes or the Constitution of the United
States. The authority under this paragraph must be exercised in
accordance with law, including section 57e of the Act, and is in
addition to the authority granted under Sec. 53.740(h), which remains
in effect unless otherwise directed by the Commission during a national
security emergency. The authority under this paragraph may be
exercised--
(1) When this action is immediately needed to implement national
security objectives as designated by the national command authority
through the Commission; and
(2) No action consistent with license conditions and technical
specifications that can satisfy national security objectives is
immediately apparent.
(d)(1) If the NRC finds that the state of emergency preparedness
does not provide reasonable assurance that adequate protective measures
can and will be taken in the event of a radiological emergency
(including findings based on requirements of 10 CFR part 50, appendix
E, section IV.D.3) and if the deficiencies (including deficiencies
based on requirements of 10 CFR part 50, appendix E, section IV.D.3)
are not corrected within 4 months of that finding, the Commission will
determine whether the facility must be shut down or cease operations
until such deficiencies are remedied or whether other enforcement
action is appropriate. In determining whether a shutdown or other
enforcement action is appropriate, the Commission will take into
account, among other factors, whether the licensee can demonstrate to
the Commission's satisfaction that the deficiencies in the plan are not
significant for the plant in question, or that adequate interim
compensating actions have been or will be taken promptly, or that that
there are other compelling reasons for continued operation.
(2) If the planning standards for radiological emergency
preparedness apply to offsite emergency response plans, or if the
planning activities in Sec. 50.160(b)(1)(iv)(B) apply, then the NRC
will base its finding on a review of the Federal Emergency Management
Agency findings and determinations as to whether State, participating
Tribal and local emergency plans are adequate and capable of being
implemented, and on the NRC assessment as to whether the licensee's
emergency plans are adequate and capable of being implemented. Nothing
in this paragraph must be construed as limiting the authority of the
Commission to take action under any other regulation or authority of
the Commission or at any time other than that specified in this
paragraph.
Sec. 53.1505 Changes to licensing-basis information requiring prior
NRC approval.
(a) Sections 53.1510 through 53.1520 provide the process for a
licensee to request and the NRC to issue amendments to licenses,
including any conditions contained therein, technical specifications or
other attachments to a license, and any orders issued by the NRC
modifying a license. Sections 53.1525 and 53.1530 govern proposed
changes to a commercial nuclear plant referencing a certified design or
manufacturing license (ML).
(b) A licensee may propose changing licensing-basis information
established by NRC regulations by requesting an exemption in accordance
with Sec. 53.080.
Sec. 53.1510 Application for amendment of license.
Whenever a holder of a license under this part desires to amend the
license, an application for an amendment must be filed with the
Commission, as specified in Sec. 53.040, that fully describes the
changes desired and, following as far as applicable, the form
prescribed for original applications. Applications for amendments
involving changes to plant structures, systems, and components (SSCs),
programmatic controls, or the role of plant personnel must include an
assessment of the changes in relation to the safety requirements in
subpart B of this part and the analyses requirements of Sec. 53.450 as
applicable, an analysis of whether the amendment involves no
significant hazards consideration using the standards in Sec. 53.1520,
and a consideration of environmental factors.
Sec. 53.1515 Public notices; State consultation.
The Commission will use the following procedures for an application
requesting an amendment to an OL or COL issued under this part.
(a) Public notices. (1)(i) The Commission may publish in the
Federal Register under Sec. 2.105 of this chapter an individual notice
of proposed action for an amendment for which it makes a proposed
determination that no significant hazards consideration is involved,
or, at least once every 30 days, publish a periodic Federal Register
notice of proposed actions, which identifies each amendment issued and
each amendment proposed to be issued since the last such periodic
notice, or it may publish both such notices.
(ii) For each amendment proposed to be issued, the notice will
(A) Contain the staff's proposed determination under the standards
in Sec. 53.1520;
(B) Provide a brief description of the amendment and of the
facility involved;
(C) Solicit public comments on the proposed determination; and
(D) Provide for a 30-day comment period.
(iii) The comment period will begin on the day after the date of
the publication of the first notice, and, normally, the amendment will
not be granted until after this comment period expires.
[[Page 87104]]
(2) The Commission may inform the public about the final
disposition of an amendment request for which it has made a proposed
determination of no significant hazards consideration either by issuing
an individual notice of issuance under Sec. 2.106 of this chapter or
by publishing such a notice in its periodic system of Federal Register
notices. In either event, it will not make and will not publish a final
determination of no significant hazards consideration unless it
receives a request for a hearing on that amendment request.
(3) Where the Commission makes a final determination that no
significant hazards consideration is involved and that the amendment
should be issued, the amendment will be effective on issuance, even if
adverse public comments have been received and even if an interested
person meeting the provisions for intervention called for in Sec.
2.309 of this chapter has filed a request for a hearing. The Commission
need hold any required hearing only after it issues an amendment,
unless it determines that a significant hazards consideration is
involved, in which case the Commission will provide an opportunity for
a prior hearing.
(4) Where the Commission finds that an emergency situation exists,
in that failure to act in a timely way would result in derating or
shutdown of a commercial nuclear reactor, or in prevention of either
resumption of operation or of increase in power output up to the
plant's licensed power level, it may issue a license amendment
involving no significant hazards consideration without prior notice and
opportunity for a hearing or for public comment. In such a situation,
the Commission will not publish a notice of proposed determination on
no significant hazards consideration but will publish a notice of
issuance under Sec. 2.106 of this chapter providing for opportunity
for a hearing and for public comment after issuance. The Commission
expects its licensees to apply for license amendments in a timely
fashion. It will decline to dispense with notice and comment on the
determination of no significant hazards consideration if it determines
that the licensee has abused the emergency provision by failing to make
timely application for the amendment and thus itself creating the
emergency. Whenever an emergency situation exists, a licensee
requesting an amendment must explain why this emergency situation
occurred and why it could not avoid this situation, and the Commission
will assess the licensee's reasons for failing to file an application
sufficiently in advance of that event.
(5) Where the Commission finds that exigent circumstances exist, in
that a licensee and the Commission must act quickly and that time does
not permit the Commission to publish a Federal Register notice allowing
30 days for prior public comment, and it also determines that the
amendment involves no significant hazards considerations, it--
(i)(A) Will either issue a Federal Register notice providing notice
of an opportunity for hearing and allowing at least 2 weeks from the
date of the notice for prior public comment; or
(B) Will use local media to provide reasonable notice to the public
in the area surrounding a licensee's facility of the licensee's
amendment and of its proposed determination as described in paragraph
(a)(1) of this section, consulting with the licensee on the proposed
media release and on the geographical area of its coverage;
(ii) Will provide for a reasonable opportunity for the public to
comment, using its best efforts to make available to the public
whatever means of communication it can for the public to respond
quickly, and, in the case of telephone comments, have these comments
recorded or transcribed, as necessary and appropriate;
(iii) When it has issued a local media release, may inform the
licensee of the public's comments, as necessary and appropriate;
(iv) Will publish a notice of issuance under Sec. 2.106 of this
chapter;
(v) Will provide a hearing after issuance, if one has been
requested by a person who satisfies the provisions for intervention
specified in Sec. 2.309 of this chapter; and
(vi) Will require the licensee to explain the exigency and why the
licensee cannot avoid it and use its normal public notice and comment
procedures in paragraph (a)(1) of this section if it determines that
the licensee has failed to use its best efforts to make a timely
application for the amendment in order to create the exigency and to
take advantage of this procedure.
(6) Where the Commission finds that significant hazards
considerations are involved, it will issue a Federal Register notice
providing an opportunity for a prior hearing even in an emergency
situation, unless it finds an imminent danger to the health or safety
of the public, in which case it will issue an appropriate order or rule
under 10 CFR part 2.
(b) State consultation. (1) At the time a licensee requests an
amendment, it must notify the State in which its facility is located of
its request by providing that State with a copy of its application and
its reasoned analysis about no significant hazards considerations and
indicate on the application that it has done so.
(2) The Commission will advise the State of its proposed
determination about no significant hazards consideration normally by
sending it a copy of the Federal Register notice.
(3) The Commission will make the names of the Project Manager or
other NRC personnel it designated to consult with the State available
to the State official designated to consult about its proposed
determination. The Commission will consider any comments of that State
official. If it does not hear from the State in a timely manner, it
will consider that the State has no interest in its determination;
nonetheless, to ensure that the State is aware of the application,
before it issues the amendment, it will make a good faith effort to
communicate directly with that official. (Inability to consult with a
responsible State official following good faith attempts will not
prevent the Commission from making effective a license amendment
involving no significant hazards consideration.)
(4) The Commission will make a good faith attempt to consult with
the State before it issues a license amendment involving no significant
hazards consideration. If, however, it does not have time to use its
normal consultation procedures because of an emergency situation, it
will attempt to communicate directly with the appropriate State
official. (Inability to consult with a responsible State official
following good faith attempts will not prevent the Commission from
making effective a license amendment involving no significant hazards
consideration, if the Commission deems it necessary in an emergency
situation.)
(5) After the Commission issues the requested amendment, it will
send a copy of its determination to the State.
(c) Caveats about State consultation. (1) The State consultation
procedures in paragraph (b) of this section do not give the State a
right--
(i) To veto the Commission's proposed or final determination;
(ii) To a hearing on the determination before the amendment becomes
effective; or
(iii) To insist upon a postponement of the determination or upon
issuance of the amendment.
(2) These procedures do not alter present provisions of law that
reserve to the Commission exclusive responsibility for setting and
enforcing radiological
[[Page 87105]]
health and safety requirements for commercial nuclear plants.
Sec. 53.1520 Issuance of amendment.
(a) In determining whether an amendment to a license will be issued
to the applicant, the Commission will be guided by the considerations
which govern the issuance of initial licenses to the extent applicable
and appropriate. If the application is for amendment of an OL or COL
and involves the material alteration of a commercial nuclear plant, a
construction permit (CP) will be issued before the issuance of the
amendment to the license, provided however, that if the application
involves a material alteration to a manufactured reactor under this
part before its installation at a site, or a COL before the date that
the Commission makes the finding under Sec. 53.1452(g), no application
for or issuance of a CP is required. If the amendment involves a
significant hazards consideration, the Commission will give notice of
its proposed action--
(1) Under Sec. 2.105 of this chapter before acting thereon; and
(2) As soon as practicable after the application has been docketed.
(b) The Commission will be particularly sensitive to a license
amendment request that involves irreversible consequences (such as one
that permits a significant increase in the amount of effluents or
radiation emitted by a commercial nuclear plant).
(c) The Commission may make a final determination, under the
procedures in Sec. 53.1515, that a proposed amendment to an OL or a
COL for a commercial nuclear plant under this part involves no
significant hazards consideration, if operation of the plant in
accordance with the proposed amendment would not--
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated; or
(2) Create the possibility of a new or different kind of an
accident from any accident previously evaluated; or
(3) Involve a significant reduction in a margin of safety.
Sec. 53.1525 Revising certification information within a design
certification rule.
(a) A holder of an OL or COL who references a design certification
rule issued under this part must request an exemption if proposing to
change one or more elements of the certification information. The
Commission may grant such a request only if it determines that the
exemption will comply with the requirements of Sec. 53.080 and that
the special circumstances outweigh any decrease in safety that may
result from the reduction in standardization caused by the departure.
(b) The request for an exemption must be included with any
associated license amendment request, which must be requested and
processed in accordance with Sec. Sec. 53.1510, 53.1515, and 53.1520.
(c) Licensees must evaluate changes to the design as described in
the Final Safety Analysis Report (FSAR) not involving changes to the
certification information using the criteria in Sec. 53.1550.
Sec. 53.1530 Revising design information within a manufacturing
license.
(a) The holder of an ML may not make changes to the design of the
manufactured reactor authorized to be manufactured without obtaining an
amendment pursuant to Sec. 53.1510 and, as applicable, Sec. 53.1520.
(b) The holder of a COL under this part who references or uses a
manufactured reactor under this part must request approval for any
proposed departure from the design characteristics, site parameters,
terms and conditions, or approved design of the manufactured reactor.
The application for such departures must be submitted and processed in
accordance with Sec. Sec. 53.1510, 53.1515, and 53.1520. In those
cases where an ML references a design certification rule, the amendment
application from the holder of the COL must also request an exemption
from the design certification rule under Sec. 53.1525 if one or more
elements of the certification information are adversely affected by the
proposed change. The holder of the COL must evaluate changes to the
commercial nuclear plant as described in the FSAR but outside of the
scope of the referenced ML using the criteria in Sec. 53.1550.
Sec. 53.1535 Amendments during construction.
(a) The holder of a CP or limited work authorization (LWA) under
this part may request an amendment to the CP or LWA in order to gain
Commission approval of the safety of selected design features or
specifications, including proposed departures from a design
certification rule or ML. Amendments to CPs or LWAs under this part
must be requested and processed under Sec. Sec. 53.1510 and 53.1520.
(b) The holder of a COL under this part for which the NRC has not
yet made a finding in accordance with Sec. 53.1452(g) must request
amendments required by Sec. 53.1525 or Sec. 53.1550 no later than 45
days from the date the licensee begins the construction of the SSCs to
implement the change or departure requiring NRC approval. The licensee
proceeds with such changes at its own risk recognizing that there is a
possibility that the amendment will not be granted.
Sec. 53.1540 Updating licensing-basis information and determining the
need for NRC approval.
(a) Sections 53.1545 through 53.1565 provide the process for a
holder of an OL or COL to modify licensing-basis information and to
evaluate potential changes to its facilities, procedures, programs, and
organizations to determine if NRC approval is required.
(b) Definitions for the purposes of Sec. Sec. 53.1545 through
53.1565--
Change means a modification or addition to, or removal from, the
commercial nuclear plant or procedures that affects a design feature or
related functional design criteria, method of performing or controlling
the functions of design features, or an evaluation that demonstrates
that intended functions will be accomplished.
Departure from a method of evaluation described in the Final Safety
Analysis Report (FSAR) (as updated) used in establishing the functional
design criteria for safety-related structures, systems, or components
or in the safety analyses means--
(1) Changing any of the elements of the method described in the
FSAR (as updated) unless the results of the analysis are conservative
or essentially the same; or
(2) Changing from a method described in the FSAR to another method
unless that method has been approved by NRC for the intended
application.
Facility as described in the FSAR (as updated) means--
(1) The SSCs that are described in the FSAR (as updated),
(2) The design and performance requirements for such SSCs described
in the FSAR (as updated), and
(3) The evaluations or methods of evaluation included in the FSAR
(as updated) for such SSCs which demonstrate that their intended
function(s) will be accomplished.
Final Safety Analysis Report (as updated) means the FSAR submitted
under Sec. 53.1369 or Sec. 53.1416, as amended and supplemented, and
as updated under Sec. 53.1545, as applicable.
Procedures as described in the Final Safety Analysis Report (as
updated) means those procedures that contain information described in
the FSAR (as updated) such as how SSCs are operated and controlled
(including assumed operator actions and response times).
[[Page 87106]]
Sec. 53.1545 Updating Final Safety Analysis Reports.
(a) Each holder of an OL or COL under this part for which the
Commission has made the finding under Sec. 53.1452(g) must update the
FSAR originally submitted as part of the application for the license
every 24 months or more frequently to assure that the information
included in the report contains the latest information developed. The
submittal must include the effects on the content of the FSAR of--
(1) Changes made to the facility or procedures as described in the
FSAR;
(2) Safety analyses and evaluations performed by the licensee
either in support of approved license amendments or in support of
conclusions that changes did not require a license amendment under
Sec. 53.1550;
(3) Updates to the probabilistic risk assessments required under
Sec. 53.450;
(4) The cumulative effects of the changes to the facility or
procedures on the margins to the safety criteria in Sec. Sec. 53.210,
53.220, 53.450(e), and 53.470 since the last FSAR update; and
(5) Analyses of new safety issues performed by or on behalf of the
licensee at Commission request.
(b)(1) The licensee must submit revisions containing updated
information to the Commission, under Sec. 53.040, identifying the
location of revised or new information.
(2) The submittal must include--
(i) A certification by a duly authorized officer of the licensee
that either the information accurately presents changes made since the
previous submittal, necessary to reflect information and analyses
submitted to the Commission or prepared pursuant to Commission
requirement, or that no such changes were made; and
(ii) An identification of changes made under the provisions of
Sec. 53.1550 but not previously submitted to the Commission.
(c) Each applicant for or holder of a COL under this part for which
the Commission has not made the finding under Sec. 53.1452(g) must
submit an update to the FSAR annually by providing the information
required in (a)(1) through (a)(5) of this section and meeting the
requirements of paragraph (b) of this section. Combined license
applicants who have requested the NRC to suspend its review of the COL
application and COL holders who have informed the NRC that they do not
plan to pursue construction need not submit an annual update of the
FSAR. If a COL applicant requests that the NRC resume its review, or a
COL holder notifies the NRC that the COL holder plans to commence or
resume construction, then the COL applicant or holder must submit to
NRC an update to its FSAR within 90 days of the request or
notification, as applicable, and annually thereafter.
(d) The FSAR (as updated) must be retained by the licensee until
the Commission terminates its license.
(e) Each holder of an ML under this part must submit an update of
the FSAR reflecting any modification to the design that is directed or
approved by the Commission under Sec. 53.1288 or Sec. 53.1530, and
any new analyses of the design requested by the Commission under Sec.
53.1580.
Sec. 53.1550 Evaluating changes to facility as described in Final
Safety Analysis Reports.
(a) The holder of an OL or COL may make changes in the facility as
described in the FSAR (as updated) and make changes in the procedures
as described in the FSAR (as updated) without obtaining a license
amendment pursuant to Sec. 53.1510 only if--
(1) A change to the technical specifications incorporated in the
license is not required; and
(2) The change meets all of the following criteria:
(i) Does not result in an increase to the frequency or consequences
of an event sequence such that an event sequence not previously
identified as risk significant becomes risk significant by the analyses
performed in accordance with Sec. 53.450(e).
(ii) Does not result in an increase to the frequency or
consequences of an event sequence such that an event sequence
identified as risk significant in accordance with Sec. 53.450(e)
exceeds the licensing-basis event evaluation criteria required to be
established in accordance with Sec. 53.450(e).
(iii) Does not involve either of the following: (A) a change to the
NRC-approved comprehensive risk metric(s) or associated risk
performance objective under Sec. 53.220(b), or (B) an increase to the
frequency or consequences of one or more event sequences such that
there is more than a minimal reduction in the margin between the
calculated comprehensive risks posed by the commercial nuclear plant
and the safety criteria of Sec. 53.220.
(iv) Does not involve a departure from a method of evaluation
described in the FSAR (as updated) used in assessing licensing-basis
events in accordance with Sec. 53.450 unless the results of the
analysis under Sec. 53.450 are conservative or essentially the same,
the revised method of evaluation has been previously approved by the
NRC for the intended application, or the revised method of evaluation
can be used under an NRC-endorsed consensus code or standard.
(v) Does not result in the escalation in the safety classification
of an SSC from non-safety-related to non-safety-related but safety-
significant or from non-safety-related but safety-significant to
safety-related.
(vi) Does not result in more than a minimal decrease in defense in
depth.
(vii) For commercial nuclear plants licensed under this part for
which alternative evaluation criteria are adopted in accordance with
Sec. 53.470, does not result in a change to the frequency or
consequences of event sequences such that the calculated margins
between the results for event sequences evaluated in accordance with
Sec. 53.450(e) and the alternative evaluation criteria decreases by 25
percent or more.
(viii) Does not result in the identification of a new design-basis
accident in accordance with Sec. 53.450(f).
(ix) Does not result in a decrease by 10 percent or more in the
margin between the consequence of any design-basis accident and the
safety criteria in Sec. 53.210.
(x) Does not prevent meeting the design requirements in Sec.
53.440(j) to limit the release of radionuclides from reactor systems,
waste stores, or other significant inventories of radioactive materials
assuming the impact of a large, commercial aircraft.
(3) In implementing this paragraph, the FSAR (as updated) is
considered to include FSAR changes since submittal of the last update
of the FSAR under Sec. 53.1545.
(4) The provisions in this section do not apply to changes to the
facility or procedures when the applicable regulations establish more
specific criteria for accomplishing such changes.
(b)(1) A licensee who references a design certification rule may
make departures from the standard design, without prior Commission
approval, unless the proposed departure involves a change to the design
as described in the rule certifying the design, in which case the
requirements of Sec. 53.1525 are applicable.
(2) The licensee must maintain records of all departures from the
certified design of the facility and these records must be maintained
and available for audit until the termination of the license. The
licensee must identify the location and nature of departures from
licensing-basis information within supporting documents for a certified
design within the updates to the Safety Analysis Report required by
Sec. 53.1545.
[[Page 87107]]
(3) Licensees for which the NRC has docketed the certifications
required under Sec. 53.1070 need not retain records of departures from
the design of the facility associated with SSCs that have been
permanently removed from service using an NRC-approved change process.
(c)(1) The licensee must maintain records of changes in the
facility and procedures made under paragraph (a) of this section. These
records must include a written evaluation which provides the bases for
the determination that the change does not require a license amendment
under paragraph (a)(2) of this section.
(2) The licensee must submit, as specified in Sec. 53.040, a
report containing a brief description of any departures and changes,
including a summary of the evaluation of each. A report must be
submitted at intervals not to exceed 24 months. For COLs, the report
must be submitted at intervals not to exceed 6 months during the period
from the date of application for a COL to the date the Commission makes
its findings under Sec. 53.1452(g).
(3) The records of changes in the facility must be maintained until
the termination of an OL or COL issued under this part, or the
termination of a renewed license issued under Sec. 53.1595--whichever
is later. Records of changes in procedures must be maintained for a
period of 5 years.
Sec. 53.1560 Updating program documents included in licensing-basis
information.
(a) Each holder under this part of an OL or COL for which the
Commission has made the finding under Sec. 53.1452(g) must biennially
or more frequently update the program documents submitted as part of an
application to obtain or maintain the license to assure that the
information included in the documents contains the latest information
developed. The submittals must include the effects on the content of
the program documents of--
(1) Changes made in the facility, procedures, licensee's
organization, or site environs;
(2) Safety analyses and evaluations performed by the applicant or
licensee either in support of approved license amendments or in support
of conclusions that changes did not require a license amendment in
accordance with Sec. 53.1550;
(3) Analyses of new safety issues performed by or on behalf of the
licensee at Commission request; and
(4) Changes to the programs as a result of operating experience,
corrective actions, or other reasons deemed appropriate to ensure the
programs serve their underlying purpose to support the requirements in
subpart B of this part or other NRC regulations.
(b)(1) The licensee must submit revisions containing updated
information to the Commission, as specified in Sec. 53.040,
identifying the location of revised or new information.
(2) The submittal must include--
(i) A certification by a duly authorized officer of the licensee
that either the information accurately presents changes made since the
previous submittals, necessary to reflect information and analyses
submitted to the Commission or prepared pursuant to Commission
requirement, or that no such changes were made; and
(ii) An identification of changes made under the provisions of
Sec. 53.1550 but not previously submitted to the Commission.
(c) The updated program documents must be retained by the licensee
until the Commission terminates their license.
Sec. 53.1565 Evaluating changes to programs included in licensing-
basis information.
(a) A licensee may make changes to the facility, procedures, or
organizations or address changes to site environs as described in the
program documents included in licensing-basis information without
obtaining prior NRC approval only if--
(1) A change to the technical specifications incorporated in the
license is not required;
(2) An exemption from an NRC regulation is not required; and
(3) The change conforms to program-specific requirements included
in regulations in this part, technical specifications, or the NRC-
approved program document included and reviewed as part of a license
application under subpart H or an amendment under this subpart.
(b) In implementing this section, the program documents (as
updated) include changes since submittal of the last updates of the
program documents pursuant to Sec. 53.1560.
(c) The provisions in this section do not apply to changes to the
program documents when the applicable regulations establish more
specific criteria for accomplishing such changes.
(d) To make changes to the facility, procedures, or organizations
or to address changes to site environs as described in the program
documents included in licensing-basis information for individual
programs, the following requirements must be satisfied:
(1) Quality assurance program--operation. (i) Each holder under
this part of an OL or COL, after the Commission makes the finding under
Sec. 53.1452(g), may make a change to a previously accepted quality
assurance program (QAP) description included or referenced in the
Safety Analysis Report without prior NRC approval, provided the change
does not reduce the commitments in the program description as accepted
by the NRC. Changes to the QAP description that do not reduce the
commitments must be submitted to the NRC in accordance with the
requirements of Sec. 53.1545. In addition to QAP changes involving
administrative improvements and clarifications, spelling corrections,
punctuation, or editorial items, the following changes are not
considered to be reductions in commitment:
(A) The use of a quality assurance (QA) standard approved by the
NRC which is more recent than the QA standard in the licensee's QAP at
the time of the change;
(B) The use of a QA alternative or exception approved by an NRC
safety evaluation, provided that the bases of the NRC approval are
applicable to the licensee's facility;
(C) The use of generic organizational position titles that clearly
denote the position function, supplemented as necessary by descriptive
text, rather than specific titles;
(D) The use of generic organizational charts to indicate functional
relationships, authorities, and responsibilities, or, alternately, the
use of descriptive text;
(E) The elimination of QAP information that duplicates language in
QA regulatory guides and QA standards to which the licensee is
committed; and
(F) Organizational revisions that ensure that persons and
organizations performing QA functions continue to have the requisite
authority and organizational freedom, including sufficient independence
from cost and schedule when opposed to safety considerations.
(ii) Changes to the QAP description that do reduce the commitments
must be submitted to the NRC and receive NRC approval prior to
implementation, as follows:
(A) Changes made to the QAP description as presented in the Safety
Analysis Report or in a topical report must be submitted as specified
in Sec. 53.040.
(B) The submittal of a change to the Safety Analysis Report QAP
description must include all pages affected by that change and must be
accompanied by a forwarding letter identifying the change, the reason
for the change, and the basis for concluding that the revised program
incorporating the change continues to satisfy the criteria of appendix
B to part
[[Page 87108]]
50 of this chapter and the Safety Analysis Report QAP description
commitments previously accepted by the NRC (the letter need not provide
the basis for changes that correct spelling, punctuation, or editorial
items).
(C) A copy of the forwarding letter identifying the change must be
maintained as a facility record for 3 years.
(D) Changes to the QAP description included or referenced in the
Safety Analysis Report shall be regarded as accepted by the Commission
upon receipt of a letter to this effect from the appropriate reviewing
office of the Commission or 60 days after submittal to the Commission,
whichever occurs first.
(2) Quality assurance program--siting, construction, and
manufacturing. Each holder of an LWA, early site permit, CP, ML, or
COL, before the Commission makes the finding under Sec. 53.1452(g) of
this chapter, under this part may make a change to a previously
accepted QAP description included or referenced in the Safety Analysis
Report without prior NRC approval, provided the change does not reduce
the commitments in the program description previously accepted by the
NRC. Changes to the QAP description that do not reduce the commitments
must be submitted to NRC within 90 days. Changes to the QAP description
that reduce the commitments must be submitted to NRC and receive NRC
approval before implementation, as follows:
(i) Changes to the Safety Analysis Report must be submitted for
review as specified in Sec. 53.040. Changes made to NRC-accepted QA
topical report descriptions must be submitted as specified in Sec.
53.040.
(ii) The submittal of a change to the Safety Analysis Report QAP
description must include all pages affected by that change and must be
accompanied by a forwarding letter identifying the change, the reason
for the change, and the basis for concluding that the revised program
incorporating the change continues to satisfy the criteria of appendix
B of part 50 of this chapter and the Safety Analysis Report QAP
description commitments previously accepted by the NRC (the letter need
not provide the basis for changes that correct spelling, punctuation,
or editorial items).
(iii) A copy of the forwarding letter identifying the changes must
be maintained as a facility record for 3 years.
(iv) Changes to the QAP description included or referenced in the
Safety Analysis Report shall be regarded as accepted by the Commission
upon receipt of a letter to this effect from the appropriate reviewing
office of the Commission or 60 days after submittal to the Commission,
whichever occurs first.
(3) Emergency preparedness program. (i) Definitions for the purpose
of paragraph (d)(3) of this section:
(A) Change means an action that results in modification or addition
to, or removal from, the licensee's emergency plan. All such changes
are subject to the provisions of this section except where the
applicable regulations establish specific criteria for accomplishing a
particular change.
(B) Emergency plan means the document(s), prepared and maintained
by the licensee, that identify and describe the licensee's methods for
maintaining emergency preparedness and responding to emergencies. An
emergency plan includes the plan as originally approved by the NRC and
all subsequent changes made by the licensee with, and without, prior
NRC review and approval under paragraph (d)(3) of this section.
(C) Emergency planning function means a capability or resource
necessary to prepare for and respond to a radiological emergency.
(D) Reduction in effectiveness means a change in an emergency plan
that results in reducing the licensee's capability to perform an
emergency planning function in the event of a radiological emergency.
(ii)(A) Except as provided in paragraph (d)(3)(ii)(B) of this
section, a holder of an OL under this part, or a COL under this part
after the Commission makes the finding under Sec. 53.1452(g), must
follow and maintain the effectiveness of an emergency plan that meets
the requirements in appendix E to part 50 of this chapter and the
planning standards of Sec. 50.47(b).
(B) A holder of an OL under this part for a commercial nuclear
plant consisting of small modular reactors (SMRs) or non-light-water
reactors, or a holder of a COL under this part after the Commission
makes the finding under Sec. 53.1452(g) for a commercial nuclear plant
consisting of either SMRs or non-light-water reactors, must follow and
maintain the effectiveness of either an emergency plan that meets the
requirements in Sec. 50.160 or an emergency plan that meets the
requirements in appendix E to part 50 of this chapter and the planning
standards of Sec. 50.47(b).
(iii)(A) Except as provided in paragraph (d)(3)(iii)(B) of this
section, the licensee may make changes to its emergency plan without
NRC approval only if the licensee performs and retains an analysis
demonstrating that the changes do not reduce the effectiveness of the
plan and the plan, as changed, continues to meet the requirements in
appendix E to part 50 of this chapter and the planning standards of
Sec. 50.47(b).
(B) A license under this part for a commercial nuclear plant
consisting of either SMRs or non-light-water reactors may make changes
to its emergency plan without NRC approval only if the licensee
performs and retains an analysis demonstrating that the changes do not
reduce the effectiveness of the plan and the plan, as changed,
continues to meet either the requirements in Sec. 50.160 or the
requirements in appendix E to part 50 and the planning standards of
Sec. 50.47(b).
(iv) The changes to a licensee's emergency plan that reduce the
effectiveness of the plan as defined in paragraph (d)(3)(i)(D) of this
section may not be implemented without prior approval by the NRC. A
licensee desiring to make such a change must submit an application for
an amendment to its license. In addition to the filing requirements of
Sec. Sec. 53.1510 and 53.1515, the request must include all emergency
plan pages affected by that change and must be accompanied by a
forwarding letter identifying the change, the reason for the change,
and the basis for concluding that the licensee's emergency plan, as
revised, will continue to meet either the requirements in Sec. 50.160
to this chapter or the requirements in appendix E to part 50 of this
chapter and the planning standards of Sec. 50.47(b) of this chapter.
(v) The licensee must retain a record of each change to the
emergency plan made without prior NRC approval for a period of three
years from the date of the change and shall submit, as specified in
Sec. 53.040, a report of each such change, including a summary of its
analysis, within 30 days after the change is put in effect.
(vi) The licensee must retain the emergency plan and each change
for which prior NRC approval was obtained pursuant to paragraph
(d)(3)(iv) of this section as a record until the Commission terminates
the license for the nuclear power reactor.
(vii)(A) The licensee must provide for the development, revision,
implementation, and maintenance of its emergency preparedness program.
The licensee must ensure that all program elements are reviewed by
persons who have no direct responsibility for the implementation of the
emergency preparedness program either--
[[Page 87109]]
(1) At intervals not to exceed 12 months; or
(2) As necessary, based on an assessment by the licensee against
performance indicators, and as soon as reasonably practicable after a
change occurs in personnel, procedures, equipment, or facilities that
potentially could adversely affect emergency preparedness, but no
longer than 12 months after the change. In any case, all elements of
the emergency preparedness program must be reviewed at least once every
24 months.
(B) The review must include an evaluation for adequacy of
interfaces with State participating Tribal and local governments and of
licensee drills, exercises, capabilities, and procedures. The results
of the review, along with recommendations for improvements, must be
documented, reported to the licensee's corporate and plant management,
and retained for a period of 5 years. The part of the review involving
the evaluation for adequacy of interface with State, participating
Tribal and local governments must be available to the appropriate
State, participating Tribal and local governments.
(4) Security programs. (i) The licensee must prepare and maintain
safeguards contingency plan procedures in accordance with appendix C of
part 73 of this chapter for affecting the actions and decisions
contained in the Responsibility Matrix of the safeguards contingency
plan. The licensee may not make a change that would decrease the
safeguard effectiveness of a physical security plan, or guard training
and qualification plan, or cybersecurity plan submitted under subpart H
or part 73 of this chapter, or of the first four categories of
information (Background, Generic Planning Base, Licensee Planning Base,
Responsibility Matrix) contained in a licensee safeguards contingency
plan submitted under subpart H or part 73 of this chapter, as
applicable, without prior approval of the Commission. A licensee
desiring to make such a change must submit an application for amendment
to the licensee's license under Sec. Sec. 53.1510, 53.1515, and
53.1520.
(ii) The licensee may make changes to the plans referenced in
paragraph (4)(i) of this section without prior Commission approval if
the changes do not decrease the safeguards effectiveness of the plan.
The licensee must maintain records of changes to the plans made without
prior Commission approval for a period of 3 years from the date of the
change, and must submit, as specified in Sec. 53.040, a report
containing a description of each change within 2 months after the
change is made. Prior to the safeguards contingency plan being put into
effect, the licensee must have--
(A) All safeguards capabilities specified in the safeguards
contingency plan available and functional;
(B) Detailed procedures developed according to appendix C to part
73 of this chapter available at the licensee's site; and
(C) All appropriate personnel trained to respond to safeguards
incidents as outlined in the plan and specified in the detailed
procedures.
(iii) The licensee must provide for the development, revision,
implementation, and maintenance of its safeguards contingency plan. The
licensee must ensure that all program elements are reviewed by
individuals independent of both security program management and
personnel who have direct responsibility for implementation of the
security program either--
(A) At intervals not to exceed 12 months; or
(B) As necessary, based on an assessment by the licensee against
performance indicators, and as soon as reasonably practicable after a
change occurs in personnel, procedures, equipment, or facilities that
potentially could adversely affect security, but no longer than 12
months after the change. In any case, all elements of the safeguards
contingency plan must be reviewed at least once every 24 months.
(iv) The review must include a review and audit of safeguards
contingency procedures and practices, an audit of the security system
testing and maintenance program, and a test of the safeguards systems
along with commitments established for response by local law
enforcement authorities. The results of the review and audit, along
with recommendations for improvements, must be documented, reported to
the licensee's corporate and plant management, and kept available at
the plant for inspection for a period of 3 years.
Sec. 53.1570 Transfer of licenses.
(a) No commercial nuclear plant license issued under this part, or
any right thereunder, shall be transferred, assigned, or in any manner
disposed of, either voluntarily or involuntarily, directly or
indirectly, through transfer of control of the license to any person,
unless the Commission gives its consent in writing.
(b)(1) An application for transfer of a license must include--
(i) As much of the information described in Sec. Sec. 53.1109,
53.1306, 53.1366, and 53.1413 with respect to the identity and
technical and financial qualifications of the proposed transferee as
would be required by those sections if the application were for an
initial license. The Commission may require additional information such
as data respecting proposed safeguards against hazards from radioactive
materials and the applicant's qualifications to protect against such
hazards.
(ii) A statement of the purposes for which the transfer of the
license is requested, the nature of the transaction necessitating or
making desirable the transfer of the license, and an agreement to limit
access to Restricted Data or Classified National Security Information
pursuant to Sec. 53.1115. The Commission may require any person who
submits an application for license pursuant to the provisions of this
section to file a written consent from the existing licensee or a
certified copy of an order or judgment of a court of competent
jurisdiction attesting to the person's right (subject to the licensing
requirements of the Act and these regulations) to possession of the
facility or site involved.
(2) [Reserved]
(c) After appropriate notice to interested persons, including the
existing licensee, and observance of such procedures as may be required
by the Act or regulations or orders of the Commission, the Commission
will approve an application for the transfer of a license, if the
Commission determines--
(1) That the proposed transferee is qualified to be the holder of
the license; and
(2) That transfer of the license is otherwise consistent with
applicable provisions of law, regulations, and orders issued by the
Commission pursuant thereto.
Sec. 53.1575 Termination of licenses.
(a) When the holder of an OL or COL under this part has determined
to permanently cease operations the licensee must, within 30 days,
submit a written certification to the NRC, consistent with the
requirements of Sec. 53.1070.
(b) Once fuel has been permanently removed from the reactor system,
the licensee must submit a written certification to the NRC that meets
the requirements of Sec. 53.1070.
(c)(1) Upon docketing of the certifications for permanent cessation
of operations and permanent removal of fuel from the reactor system, or
when a final legally effective order to permanently cease operations
has come into effect, the license no longer authorizes operation of the
reactor or
[[Page 87110]]
emplacement or retention of fuel into the reactor system.
(2) Activities associated with decommissioning will be carried out
in accordance with the requirements and procedures in subpart G of this
part.
(3) The Commission shall terminate the license if it determines
that--
(i) The remaining dismantlement has been performed in accordance
with the approved license termination plan required in subpart G of
this part; and
(ii) The final radiation survey and associated documentation,
including an assessment of dose contributions associated with parts
released for use before approval of the license termination plan,
demonstrate that the facility and site have met the criteria for
decommissioning in subpart E of 10 CFR part 20.
(d) A holder of a CP or COL under this part may request the
termination of the license as well as licenses issued by the NRC under
parts 30, 40, or 70 of this chapter prior to plant operations. Such
requests may support an immediate NRC approval of the site for
unrestricted use.
Sec. 53.1580 Information requests.
Each licensee under this part must at any time before termination
of the license, upon request of the Commission, submit, as specified in
Sec. 53.040 written statements, signed under oath or affirmation, to
enable the Commission to determine whether or not the license should be
modified, suspended, or revoked. Except for information sought to
verify licensee compliance with the current licensing basis for that
facility, the NRC must prepare the reason or reasons for each
information request prior to issuance to ensure that the burden to be
imposed on respondents is justified in view of the potential safety
significance of the issue to be addressed in the requested information.
Each such justification provided for an evaluation performed by the NRC
staff must be approved by the Executive Director for Operations or his
or her designee prior to issuance of the request.
Sec. 53.1585 Revocation, suspension, modification of licenses and
approvals for cause.
A license or standard design approval issued under this part may be
revoked, suspended, or modified, in whole or in part, for any material
false statement in the application or in the supplemental or other
statement of fact required of the applicant; or because of conditions
revealed by the application or statement of fact of any report, record,
inspection, or other means which would warrant the Commission to refuse
to grant a license or approval on an original application; or for
failure to manufacture a reactor, or construct or operate a facility in
accordance with the terms of the license, provided, however, that
failure to make timely completion of the proposed construction or
alteration of a facility under a CP under this part shall be governed
by the provisions of Sec. 53.1342(b); or for violation of, or failure
to observe, any of the terms and provisions of the Act, regulations,
license, approval, or order of the Commission.
Sec. 53.1590 Backfitting.
(a)(1) Backfitting means the modification of or addition to
systems, structures, components, or design of a facility; or the design
approval or ML for a facility; or the procedures or organization
required to design, construct or operate a facility; any of which may
result from a new or amended provision in the Commission's regulations
or the imposition of a regulatory staff position interpreting the
Commission's regulations that is either new or different from a
previously applicable staff position after the date of the commercial
nuclear plant license issued under this part.
(2) Except as provided in paragraph (a)(4) of this section, the
Commission shall require a systematic and documented analysis pursuant
to paragraph (b) of this section for backfits which it seeks to impose.
(3) Except as provided in paragraph (a)(4) of this section, the
Commission shall require the backfitting of a facility only when it
determines, based on the analysis described in paragraph (b) of this
section, that there is a substantial increase in the overall protection
of the public health and safety or the common defense and security to
be derived from the backfit and that the direct and indirect costs of
implementation for that facility are justified in view of this
increased protection.
(4) The provisions of paragraphs (a)(2) and (a)(3) of this section
are inapplicable and, therefore, backfit analysis is not required and
the standards in paragraph (a)(3) of this section do not apply where
the Commission or staff, as appropriate, finds and declares, with
appropriate documented evaluation for its finding, either--
(i) That a modification is necessary to bring a facility into
compliance with a license or the rules or orders of the Commission, or
into conformance with written commitments by the licensee; or
(ii) That regulatory action is necessary to ensure that the
facility provides adequate protection to the health and safety of the
public and is in accord with the common defense and security; or
(iii) That the regulatory action involves defining or redefining
what level of protection to the public health and safety or common
defense and security should be regarded as adequate.
(5) The Commission must always require the backfitting of a
facility if it determines that such regulatory action is necessary to
ensure that the facility provides adequate protection to the health and
safety of the public and is in accord with the common defense and
security.
(6) The documented evaluation required by paragraph (a)(4) of this
section must include a statement of the objectives of and reasons for
the modification and the basis for invoking the exception. If
immediately effective regulatory action is required, then the
documented evaluation may follow rather than precede the regulatory
action.
(7) If there are two or more ways to achieve compliance with a
license or the rules or orders of the Commission, or with written
licensee commitments, or there are two or more ways to reach a level of
protection which is adequate, then ordinarily the applicant or licensee
is free to choose the way which best suits its purposes. However,
should it be necessary or appropriate for the Commission to prescribe a
specific way to comply with its requirements or to achieve adequate
protection, then cost may be a factor in selecting the way, provided
that the objective of compliance or adequate protection is met.
(b) In reaching the determination required by paragraph (a)(3) of
this section, the Commission will consider how the backfit should be
scheduled in light of other ongoing regulatory activities at the
facility and, in addition, will consider information available
concerning any of the following factors as may be appropriate and any
other information relevant and material to the proposed backfit:
(1) The statement of the specific objectives that the proposed
backfit is designed to achieve;
(2) The general description of the activity that would be required
by the licensee or applicant in order to complete the backfit;
(3) The potential change in the risk to the public from the
accidental off-site release of radioactive material;
(4) The potential impact on radiological exposure of facility
employees;
[[Page 87111]]
(5) The installation and continuing costs associated with the
backfit, including the cost of facility downtime or the cost of
construction delay;
(6) The potential safety impact of changes in plant or operational
complexity, including the relationship to proposed and existing
regulatory requirements;
(7) The estimated resource burden on the NRC associated with the
proposed backfit and the availability of such resources;
(8) The potential impact of differences in facility type, design or
age on the relevancy and practicality of the proposed backfit;
(9) Whether the proposed backfit is interim or final and, if
interim, the justification for imposing the proposed backfit on an
interim basis.
(c) No licensing action will be withheld during the pendency of
backfit analyses required by the Commission's rules.
(d) The Executive Director for Operations shall be responsible for
implementation of this section, and all analyses required by this
section shall be approved by the Executive Director for Operations or
his or her designee.
Sec. 53.1595 Renewal.
Licenses may be renewed by the Commission upon expiration of the
period of the license.
Subpart J--Reporting and Other Administrative Requirements
Sec. 53.1600 General information.
Each applicant and licensee under this part must ensure that U.S.
Nuclear Regulatory Commission (NRC) inspectors have unfettered access
to sites and facilities licensed or proposed to be licensed in Sec.
53.1610, must maintain records and make reports to the NRC in
accordance with requirements in Sec. Sec. 53.1620 through 53.1650,
must satisfy financial qualification and reporting requirements in
Sec. Sec. 53.1660 through 53.1700, and must obtain and maintain
required financial protections in case of an accident in Sec. Sec.
53.1720 and 53.1730.
Sec. 53.1610 Unfettered access for inspections.
(a) Each applicant for or holder of a manufacturing license (ML),
operating license (OL), combined license (COL), construction permit
(CP), or early site permit must permit inspection, by duly authorized
representatives of the Commission, of its records, premises,
activities, and of licensed materials in possession or use, related to
the license or CP or early site permit as may be necessary to
effectuate the purposes of the Atomic Energy Act of 1956, as amended,
(the Act) and the Energy Reorganization Act of 1974, as amended.
(b)(1) Each holder of an ML, OL, COL, or CP must, upon request by
the Director, Office of Nuclear Reactor Regulation, provide rent-free
office space for the exclusive use of the Commission inspection
personnel. Heat, air conditioning, light, electrical outlets, and
janitorial services must be furnished by each licensee and each holder
of a CP. The office must be convenient to and have full access to the
facility and must provide the inspectors both visual and acoustic
privacy.
(2) For a site or facility with an assigned resident inspector, the
space provided must be adequate to accommodate a full-time inspector, a
part-time secretary, and transient NRC personnel and must be generally
commensurate with other office facilities at the site. For sites or
facilities assigned multiple resident inspectors, additional space may
be requested. The office space that is provided must be subject to the
approval of the Director, Office of Nuclear Reactor Regulation. All
furniture, supplies, and communication equipment will be furnished by
the Commission.
(3) For a site or facility without an assigned resident inspector,
temporary space to accommodate periodic or special inspections must be
provided. The office space must be generally commensurate with other
office accommodations at the site.
(4) The licensee or permit holder must afford any NRC resident
inspector assigned to that site, or other NRC inspectors identified by
the Regional Administrator as likely to inspect the facility, immediate
unfettered access, equivalent to access provided regular plant
employees, following proper identification and compliance with
applicable access control measures for security, radiological
protection, and personal safety.
(5) The licensee or permit holder must ensure that the arrival and
presence of an NRC inspector, who has been properly authorized facility
access as described in paragraph (b)(4) of this section, is not
announced or otherwise communicated by its employees or contractors to
other persons at the facility unless specifically requested by the NRC
inspector.
Sec. 53.1620 Maintenance of records, making of reports.
(a) Each holder of an ML, OL, COL, CP, or early site permit must
maintain all records and make all reports, in connection with the
activity, as may be required by the conditions of the license or permit
or by the regulations and orders of the Commission in effectuating the
purposes of the Act and the Energy Reorganization Act of 1974, as
amended. Reports must be submitted in accordance with Sec. 53.040.
(b) [Reserved]
(c) Records that are required by the regulations in this part, by
license condition, or by technical specifications must be retained for
the period specified by the appropriate regulation, license condition,
or technical specification. If a retention period is not otherwise
specified, these records must be retained until the Commission
terminates the facility license or, in the case of an early site
permit, until the permit expires.
(d)(1) Records which must be retained under this part may be the
original or a reproduced copy or a microform if the reproduced copy or
microform is duly authenticated by authorized personnel and the
microform is capable of producing a clear and legible copy after
storage for the period specified by Commission regulations. The record
may also be stored in electronic media with the capability of producing
legible, accurate, and complete records during the required retention
period. Records such as letters, drawings, and specifications, must
include all pertinent information such as stamps, initials, and
signatures. The licensee must maintain adequate safeguards against
tampering with, and loss of records.
(2) If there is a conflict between the Commission's regulations in
this part, license condition, or technical specification, or other
written Commission approval or authorization pertaining to the
retention period for the same type of record, the retention period
specified in the regulations in this part for such records shall apply
unless the Commission, under Sec. 53.080 of this part, has granted a
specific exemption from the record retention requirements in the
regulations in this part.
(e) Each licensee must notify the Commission as specified in Sec.
53.040 of this part, of successfully completing power ascension testing
or startup testing as applicable within 30 calendar days of completing
the testing.
Sec. 53.1630 Immediate notification requirements for operating
commercial nuclear plants.
(a) General requirements.\1\ (1) Each holder of an OL under this
part or a COL under this part after the Commission makes the finding
under Sec. 53.1452(g), must notify the NRC Operations Center
[[Page 87112]]
via the Emergency Notification System (ENS) of--
(i) The declaration of any of the Emergency Classes specified in
the licensee's approved Emergency Plan; or
(ii) Those non-emergency events specified in paragraph (b) of this
section that occurred within 3 years of the date of discovery.
(2) If the ENS is inoperative, the licensee must make the required
notifications via commercial telephone service, other dedicated
telephone system, or any other method which will ensure that a report
is made as soon as practical to the NRC Headquarters Operations Center
at the numbers specified in appendix A to part 73 of this chapter.
(3) The licensee must notify the NRC immediately after notification
of the appropriate State or local agencies and not later than 1 hour
after the time the licensee declares one of the Emergency Classes.
(4) The licensee must activate the data links with the NRC as
specified in their emergency plans after declaring an Emergency Class
for events of actual or potential substantial degradation of plant
safety or security, probable risk to site personnel life, or site
equipment damage caused by hostile action. The data links may also be
activated by the licensee during emergency drills or exercises if the
licensee's computer system has the capability to transmit the exercise
data.
(5) When making a report under paragraph (a)(1) of this section,
the licensee must identify--
(i) The Emergency Class declared; or
(ii) Paragraph (b)(1), ``One-hour reports,'' paragraph (b)(2),
``Four-hour reports,'' or paragraph (b)(3), ``Eight-hour reports,'' as
the paragraph of this section requiring notification of the non-
emergency event.
(b) Non-emergency events. (1) One-hour reports. If not reported as
a declaration of an Emergency Class under paragraph (a) of this
section, the licensee must notify the NRC as soon as practical and in
all cases within one hour of the occurrence of any deviation from the
plant's Technical Specifications authorized under Sec. 53.740(h) of
this part.
(2) Four-hour reports. If not reported under paragraphs (a) or
(b)(1) of this section, the licensee must notify the NRC as soon as
practical, and in all cases, within 4 hours of the occurrence of any of
the following:
(i) The initiation of any commercial nuclear plant shutdown
required by the plant's Technical Specifications.
(ii) Any event or condition that results in actuation of the
reactor protection system when the reactor is critical except when the
actuation results from and is part of a pre-planned sequence during
testing or reactor operation.
(iii) Any event or condition that results in an unplanned actuation
of a safety-related (SR) standby cooling system or the unplanned sole
reliance on an SR standby cooling system for those systems that are in
constant operation.
(iv) Any event or condition that results in an unplanned movement
of, change of state in, or chemical interaction involving a significant
amount of radioactive material within the commercial nuclear plant.
(v) Any event or situation, related to the health and safety of the
public or onsite personnel, or protection of the environment, for which
a news release is planned or notification to other government agencies
has been or will be made. Such an event may include an onsite fatality
or inadvertent release of radioactively contaminated materials.
(3) Eight-hour reports. If not reported under paragraphs (a),
(b)(1), or (b)(2) of this section, the licensee must notify the NRC as
soon as practical and in all cases within 8 hours of the occurrence of
any of the following:
(i) Any event or condition that results in--
(A) The condition of the commercial nuclear plant, including its
principal safety barriers, being seriously degraded; or
(B) The commercial nuclear plant being in a condition not analyzed
under Sec. 53.450 that significantly degrades plant safety.
(ii) Any event or condition that results in valid actuation of an
SR system, except when the actuation results from and is part of a pre-
planned sequence during testing or reactor operation.
(iii) Any event or condition that at the time of discovery could
have prevented the fulfilment of the safety functions identified under
Sec. 53.230. Events covered may include one or more procedural errors,
equipment failures, and/or discovery of design, analysis, fabrication,
construction, and/or procedural inadequacies. However, individual
component failures need not be reported pursuant to this paragraph if
other equipment was operable and available to perform the required
safety function.
(iv) Any event requiring the transport of a radioactively
contaminated person to an offsite medical facility for treatment.
(v) Any event that results in a major loss of emergency assessment
capability, offsite response capability, or offsite communications
capability (e.g., significant portion of control room indication, ENS,
or offsite notification system).
(c) Follow-up notification: With respect to the notifications made
under paragraphs (a) and (b) of this section, in addition to making the
required initial notification, each licensee, must during the course of
the event--
(1) Immediately Report:
(i) any further degradation in the level of safety of the plant or
other worsening plant conditions, including those that require the
declaration of any of the Emergency Classes, if such a declaration has
not been previously made, or
(ii) any change from one Emergency Class to another, or
(iii) a termination of the Emergency Class.
(2) Immediately Report:
(i) the results of ensuing evaluations or assessments of plant
conditions,
(ii) the effectiveness of response or protective measures taken,
and
(iii) important information related to plant behavior that is not
understood.
(3) Maintain an open, continuous communication channel with the NRC
Operation Center upon request by the NRC.
\1\ Other requirements for immediate notification of the NRC by
licensed operating commercial nuclear plants are contained elsewhere
in this chapter, in particular Sec. Sec. 20.1906, 20.2202, 72.216,
73.77, and 73.1200 of this chapter.
Sec. 53.1640 Licensee event report system.
(a) Reportable events. (1) Each commercial nuclear plant licensee
holding an OL under this part or a COL under this part after the
Commission makes the finding under Sec. 53.1452(g), must submit a
Licensee Event Report (LER) for any event of the type described in this
paragraph within 60 days after discovery of the event. In the case of
an invalid actuation reported under Sec. 53.1640(a)(2), other than
automatic reactor shutdown when the reactor is critical, the licensee
may, at its option, provide a telephone notification to the NRC
Operations Center within 60 days after discovery of the event instead
of submitting a written LER. Unless otherwise specified in this
section, the licensee must report an event if it occurred within 3
years of the date of discovery regardless of the plant mode or power
level, and regardless of the significance of the structure, system, or
component that initiated the event.
(2) The licensee must report--
(i)(A) The completion of any commercial nuclear plant shutdown
required by the plant's Technical Specifications.
[[Page 87113]]
(B) Any operation or condition which was prohibited by the plant's
Technical Specifications except when--
(1) The Technical Specification is administrative in nature;
(2) The event consisted solely of a case of a late surveillance
test where the oversight was corrected, the test was performed, and the
equipment was found to be capable of performing its specified safety
functions; or
(3) The Technical Specification was revised prior to discovery of
the event such that the operation or condition was no longer prohibited
at the time of the event.
(C) Any deviation from the plant's Technical Specifications
authorized under Sec. 53.740(h).
(ii) Any event or condition that resulted in--
(A) The condition of the commercial nuclear plant, including its
principal safety barriers, being seriously degraded; or
(B) The commercial nuclear plant being in a condition not analyzed
under Sec. 53.450 that significantly degrades plant safety.
(iii) Any natural phenomena or other external condition that posed
an actual threat to the safety of the commercial nuclear plant or
significantly hampered site personnel in the performance of duties
necessary for the safe operation of the commercial nuclear plant.
(iv) Any event or condition that resulted in inadvertent operation
of any structures, systems, and component classified as SR for an
identified safety function under Sec. 53.460 or the unplanned sole
reliance on an SR system for those systems that are in constant
operation, except when--
(A) The actuation resulted from and was part of a pre-planned
sequence during testing; or
(B) The actuation was invalid and--
(1) Occurred while the system was properly removed from service; or
(2) Occurred after the safety function had been already completed.
(v) Any event or condition that could have prevented the
fulfillment of the safety functions identified under Sec. 53.230.
(vi) Events covered in paragraph (a)(2)(v) of this section may
include one or more procedural errors, equipment failures, and/or
discovery of design, fabrication, construction, and/or procedural
inadequacies. However, individual component failures need not be
reported pursuant to paragraph (a)(2)(v) of this section if any other
equipment was operable and available to perform the required safety
function.
(vii)(A) Any event or condition that as a result of a single cause
could have prevented the fulfillment of any of the safety functions
identified under Sec. 53.230.
(B) Events covered in paragraph (a)(2)(vii)(A) of this section may
include cases of procedural error, equipment failure, and/or discovery
of a design, analysis, fabrication, construction, and/or procedural
inadequacy. However, licensees are not required to report an event
pursuant to paragraph (a)(2)(vii)(A) of this section if the event
results from--
(1) A shared dependency among trains or channels that is a natural
or expected consequence of the approved plant design; or
(2) Normal and expected wear or degradation.
(viii)(A) Any airborne radioactive release that, when averaged over
a time period of 1-hour, resulted in airborne radionuclide
concentrations in an unrestricted area that exceeds 20 times the
applicable concentration limits specified in appendix B to 10 CFR part
20, table 2, column 1.
(B) Any liquid effluent release that, when averaged over a time
period of 1-hour, exceeds 20 times the applicable concentrations
specified in appendix B to 10 CFR part 20, table 2, column 2, at the
point of entry into the receiving waters (i.e., unrestricted area) for
all radionuclides except tritium and dissolved noble gases.
(ix) Any event that posed an actual threat to the safety of the
commercial nuclear plant or significantly hampered site personnel in
the performance of duties necessary for the safe operation of the
plant, including fires, toxic gas releases, or radioactive releases.
(b) Contents. The LER must contain--
(1) A brief abstract describing the major occurrences during the
event, including all component or system failures that contributed to
the event and significant corrective action taken or planned to prevent
recurrence.
(2)(i) A clear, specific narrative description of what occurred so
that knowledgeable readers conversant with the design of commercial
nuclear plants, but not familiar with the details of a particular
plant, can understand the complete event.
(ii) The narrative description must include the following specific
information as appropriate for the particular event:
(A) Plant operating conditions before the event.
(B) Status of systems, structures, or components that were
inoperable at the start of the event and that contributed to the event.
(C) Dates and approximate time of the occurrences.
(D) The cause of each component or system failure or personnel
error, if known.
(E) The failure mode, mechanism, and effect of each failed
component, if known.
(F) [Reserved]
(G) For failures of components with multiple functions, include a
list of systems or secondary functions that were also affected.
(H) For failure that rendered a component or system classified as
SR or non-safety-related but safety-significant inoperable, an estimate
of the elapsed time from the discovery of the failure until the
component or system was returned to service.
(I) The method of discovery of each component or system failure or
procedural error.
(J) For each human performance related root cause, the licensee
must discuss the cause(s) and circumstances.
(K) Automatically and manually initiated safety system responses.
(L) The manufacturer and model number (or other identification) of
each component that failed during the event.
(3) An assessment of the safety consequences and implications of
the event. This assessment must include--
(i) The availability of systems or components that could have
performed the same function as the components and systems that failed
during the event, and
(ii) For events that occurred when the reactor was shut down, the
availability of systems or components that are needed to shut down the
reactor and maintain safe shutdown conditions, remove residual heat,
control the release of radioactive material, or mitigate the
consequences of an accident.
(4) A description of any corrective actions planned as a result of
the event, including those to reduce the probability of similar events
occurring in the future.
(5) Reference to any previous similar events at the same plant that
are known to the licensee.
(6) The name and contact information of a person within the
licensee's organization who is knowledgeable about the event and can
provide additional information concerning the event and the plant's
characteristics.
(c) Supplemental Information. The Commission may require the
licensee to submit specific additional information beyond that required
by paragraph (b) of this section if the Commission finds that
supplemental material is necessary for complete understanding of an
unusually complex or significant event. These requests for supplemental
information will be made in writing and the licensee
[[Page 87114]]
must submit, as specified in Sec. 53.040, the requested information as
a supplement to the initial LER.
(d) Submission of Reports. Licensee Event Reports must be prepared
on Form NRC 366 and submitted to the NRC, as specified in Sec. 53.040.
(e) Report Legibility. The reports and copies that licensees are
required to submit to the Commission under the provisions of this
section must be of sufficient quality to permit legible reproduction
and micrographic processing.
53.1645 Reports of radiation exposure to members of the public.
(a) Each holder of an OL, and each holder of a COL after the
Commission has made the finding under Sec. 53.1452(g), must submit
radiological reports as required by 10 CFR part 20, as well as an
Annual Radioactive Effluent Release Report and an Annual Radiological
Environmental Operating Report. The Annual Radioactive Effluent Release
Report must specify the quantity of each of the principal radionuclides
released to unrestricted areas in liquid and in gaseous effluents and
an estimate of the dose received by the maximally exposed member of the
public in an unrestricted area from effluents and direct radiation from
contained sources during the previous calendar year. The Annual
Radiological Environmental Operating Report must provide data on
measurable levels of radiation and radioactive materials in the
environment, must include an evaluation of the relationship between
quantities of radioactive material released in effluents and resultant
radiation doses to individuals from principal pathways of exposure, and
must include the results of environmental monitoring during the
previous calendar year. These reports must also include any other
information as may be required by the Commission to estimate maximum
potential annual radiation doses to the public. The reports must be
submitted as specified in Sec. 53.040 by May 15 of each successive
year. If the total effective dose equivalent to members of the public
in unrestricted areas during the reporting period is greater than the
as low as is reasonably achievable (ALARA) design objectives
established under Sec. 53.425, the report must specify the causes for
exceeding the ALARA design objective and describe any corrective
actions. On the basis of these reports and any additional information
the Commission may obtain from the licensee or others, the Commission
may require the licensee to take action as the Commission deems
appropriate.
(b) If during any calendar quarter the radiation exposure to a
member of the public in the unrestricted areas, calculated on the same
basis as the respective ALARA design objective exposure, exceeds one-
half of the annual ALARA design objective exposure, the licensee must
submit a report as specified in Sec. 53.040. The report shall specify
the causes for exceeding one-half the annual ALARA design objective
exposure in a quarter and describe corrective actions that the licensee
will take to maintain radiation exposure to levels within the ALARA
design objectives for the remainder of the year. The report shall be
submitted within 30 days from the end of the quarter when one-half of
the annual ALARA design objective exposure was exceeded.
Sec. 53.1650 Facility information and verification.
(a) In response to a written request by the Commission, each
applicant for a CP or license and each recipient of a CP or a license
must submit facility information, as described in Sec. 75.10 of this
chapter, on International Atomic Energy Agency (IAEA) Design
Information Questionnaire forms and site information on DOC/NRC Form
AP-A and associated forms;
(b) As required by the Additional Protocol, must submit location
information described in Sec. 75.11 of this chapter on DOC/NRC Form
AP-1 and associated forms; and
(c) Must permit verification thereof by the IAEA and take other
action as necessary to implement the US/IAEA Safeguards Agreement, as
described in part 75 of this chapter.
Sec. 53.1660 Financial requirements.
Sections 53.1670 through 53.1700 set out the requirements and
procedures related to financial qualifications and related reporting
requirements.
Sec. 53.1670 Financial qualifications.
Except for an electric utility applicant for a license to operate a
commercial nuclear plant, an applicant for a CP, OL, or COL under this
part must possess or have reasonable assurance of obtaining the funds
necessary for the activities for which the permit or license is sought.
Sec. 53.1680 Annual financial reports.
With respect to any commercial nuclear plant of a type described in
Sec. 53.020, each licensee and each holder of a CP must submit its
annual financial report, including the certified financial statements,
to the Commission, as specified in Sec. 53.040, upon issuance of the
report. However, licensees and holders of a CP who submit a Form 10-Q
with the Securities and Exchange Commission or a Form 1 with the
Federal Energy Regulatory Commission need not submit the annual
financial report or the certified financial statement under this
section.
Sec. 53.1690 Licensee's change of status; financial qualifications.
(a) An electric utility licensee holding an OL or COL (including a
renewed license) for a commercial nuclear plant, no later than seventy-
five (75) days prior to ceasing to be an electric utility in any manner
not involving a license transfer under Sec. 53.1399 or Sec. 53.1456
must provide the NRC with the financial qualifications information that
would be required for obtaining an initial OL or COL under this part.
The financial qualifications information must address the first full 5
years of operation after the date the licensee ceases to be an electric
utility.
(b)(1) Any holder of a license issued under this part must notify
the appropriate NRC Regional Administrator, in writing, immediately
following the filing of a voluntary or involuntary petition for
bankruptcy under any chapter of title 11 (Bankruptcy) of the United
States Code by or against--
(i) The licensee;
(ii) An entity (as 11 U.S.C. 101(14) defines that term) controlling
the licensee or listing the license or licensee as property of the
estate; or
(iii) An affiliate (as 11 U.S.C. 101(2) defines that term) of the
licensee.
(2) This notification must indicate--
(i) The bankruptcy court in which the petition for bankruptcy was
filed; and
(ii) The date of the filing of the petition.
Sec. 53.1700 Creditor regulations.
(a) Pursuant to section 184 of the Act, the Commission consents,
without individual application, to the creation of any mortgage,
pledge, or other lien upon any facility not owned by the United States
which is the subject of a license or upon any leasehold or other
interest in such facility; provided--
(1) That the rights of any creditor so secured may be exercised
only in compliance with and subject to the same requirements and
restrictions as would apply to the licensee pursuant to the provisions
of the license, the Act, and regulations issued by the Commission under
the Act; and
(2) That no creditor so secured may take possession of the facility
pursuant to the provisions of this section prior to either the issuance
of a license from the Commission authorizing such possession or the
transfer of the license.
[[Page 87115]]
(b) Any creditor so secured may apply for transfer of the license
covering such facility by filing an application for transfer of the
license under Sec. 53.1570. The Commission will act upon such
application under subpart I of this part.
(c) Nothing contained in this regulation shall be deemed to affect
the means of acquiring, or the priority of, any tax lien or other lien
provided by law.
(d) As used in this section--
License includes any license under this part, which may be issued
by the Commission with regard to a facility.
Creditor includes, without implied limitation, the trustee under
any mortgage, pledge or lien on a facility made to secure any creditor,
any trustee or receiver of the facility appointed by a court of
competent jurisdiction in any action brought for the benefit of any
creditor secured by such mortgage, pledge or lien, any purchaser of
such facility at the sale thereof upon foreclosure of such mortgage,
pledge, or lien or upon exercise of any power of sale contained
therein, or any assignee of any such purchaser.
Facility includes, but is not limited to, a site which is the
subject of an early site permit under this part, and a reactor
manufactured under an ML under this part.
Sec. 53.1710 Financial protection.
Sections 53.1720 and 53.1730 set out the requirements and
procedures related to licensees obtaining and maintaining insurance to
cover stabilization and decontamination activities in the event of an
accident and financial protection in accordance with part 140,
``Financial Protection Requirements and Indemnity Agreements,'' of this
chapter.
Sec. 53.1720 Insurance required to stabilize and decontaminate plant
following an accident.
Each commercial nuclear plant licensee under this part must take
reasonable steps to obtain insurance available at reasonable costs and
on reasonable terms from private sources or to demonstrate that it
possesses an equivalent amount of protection covering the licensee's
obligation, in the event of an accident at the licensee's commercial
nuclear reactor, to stabilize and decontaminate the plant and the plant
site at which such an accident may occur, provided that--
(a) The insurance required by this section must have a minimum
coverage limit for each commercial nuclear plant site of $1.06 billion,
an amount based on plant-specific estimates of costs to stabilize and
decontaminate a plant, or whatever amount of insurance is generally
available from private sources, whichever is less. The required
insurance must clearly state that, as and to the extent provided in
paragraph (d)(1) of this section, any proceeds must be payable first
for stabilization of the plant and next for decontamination of the
plant and the plant site. If a licensee's coverage falls below the
required minimum, the licensee must within 60 days take all reasonable
steps to restore its coverage to the required minimum. The required
insurance may, at the option of the licensee, be included within
policies that also provide coverage for other risks, including, but not
limited to, the risk of direct physical damage.
(b)(1) With respect to policies issued or annually renewed, the
proceeds of such required insurance must be dedicated, as and to the
extent provided in this paragraph, to reimbursement or payment on
behalf of the insured of reasonable expenses incurred or estimated to
be incurred by the licensee in taking action to fulfill the licensee's
obligation, in the event of an accident at the licensee's plant, to
ensure that the plant is in, or is returned to, and maintained in, a
safe and stable condition and that radioactive contamination is removed
or controlled such that personnel exposures are consistent with the
occupational exposure limits in 10 CFR part 20. These actions must be
consistent with any other obligation the licensee may have under this
chapter and must be subject to paragraph (d) of this section. As used
in this section, an ``accident'' means an event that involves the
release of radioactive material from its intended place of confinement
within the commercial nuclear plant such that there is a present danger
of release off site in amounts that would pose a threat to the public
health and safety.
(2) The stabilization and decontamination requirements set forth in
paragraph (d) of this section must apply uniformly to all insurance
policies required under this section.
(c) The licensee shall report to the NRC on April 1 of each year
the current levels of this insurance or financial security it maintains
and the sources of this insurance or financial security.
(d)(1) In the event of an accident at the licensee's plant,
whenever the estimated costs of stabilizing the licensed plant and of
decontaminating the plant and the plant site exceed one tenth of the
minimum insurance under paragraph (a) of this section, the proceeds of
the insurance required by this section must be dedicated to and used,
first, to ensure that the licensed plant is in, or is returned to, and
can be maintained in, a safe and stable condition so as to prevent any
significant risk to the public health and safety and, second, to
decontaminate the plant and the plant site in accordance with the
licensee's cleanup plan as approved by order of the Director, Office of
Nuclear Reactor Regulation. This priority on insurance proceeds must
remain in effect for 60 days or, upon order of the Director, for such
longer periods, in increments not to exceed 60 days except as provided
for activities under the cleanup plan required in paragraphs (d)(3) and
(d)(4) of this section, as the Director may find necessary to protect
the public health and safety. Actions needed to bring the plant to and
maintain the plant in a safe and stable condition may include one or
more of the following, as appropriate:
(i) Shutdown of the reactor(s) and other processes at the plant;
(ii) Establishment and maintenance of long-term cooling with stable
decay heat removal;
(iii) Maintenance of sub-criticality;
(iv) Control of radioactive releases; and
(v) Securing of structures, systems, or components to minimize
radiation exposure to onsite personnel or to the offsite public or to
facilitate later decontamination or both.
(2) The licensee must inform the Director, Office of Nuclear
Reactor Regulation in writing when the plant is and can be maintained
in a safe and stable condition so as to prevent any significant risk to
the public health and safety. Within 30 days after the licensee informs
the Director that the plant is in this condition, or at such earlier
time as the licensee may elect or the Director may for good cause
direct, the licensee must prepare and submit a cleanup plan for the
Director's approval. The cleanup plan must identify and contain an
estimate of the cost of each cleanup operation that will be required to
decontaminate the reactor sufficiently to permit the licensee either to
resume operation of the reactor or to apply to the Commission under
subpart G of this part for authority to decommission the reactor and to
surrender the license voluntarily. Cleanup operations may include one
or more of the following, as appropriate:
(i) Processing any contaminated materials generated by the accident
and by decontamination operations to remove radioactive materials;
(ii) Decontamination of surfaces inside the plant buildings to
levels consistent with the Commission's occupational exposure limits in
10 CFR part 20, and decontamination or disposal of equipment;
[[Page 87116]]
(iii) Decontamination or removal and disposal of internal parts,
damaged fuel from the reactor coolant or fuel systems, or related
process or waste systems; and
(iv) Cleanup of the reactor coolant or fuel systems or related
process or waste systems.
(3) Following review of the licensee's cleanup plan, the Director
will order the licensee to complete all operations that the Director
finds are necessary to decontaminate the reactor sufficiently to permit
the licensee either to resume operation of the reactor or to apply to
the Commission under subpart G of this part for authority to
decommission the reactor and to surrender the license voluntarily. The
Director must approve or disapprove, in whole or in part for stated
reasons, the licensee's estimate of cleanup costs for such operations.
Such order may not be effective for more than one year, at which time
it may be renewed. Each subsequent renewal order, if imposed, may be
effective for not more than 6 months.
(4) Of the balance of the proceeds of the required insurance not
already expended to place the plant in a safe and stable condition
under paragraph (b)(1) of this section, an amount sufficient to cover
the expenses of completion of those decontamination operations that are
the subject of the Director's order must be dedicated to such use,
provided that, upon certification to the Director of the amounts
expended previously and from time to time for stabilization and
decontamination and upon further certification to the Director as to
the sufficiency of the dedicated amount remaining, policies of
insurance may provide for payment to the licensee or other loss payees
of amounts not so dedicated, and the licensee may proceed to use in
parallel (and not in preference thereto) any insurance proceeds not so
dedicated for other purposes.
Sec. 53.1730 Financial protection requirements.
Commercial nuclear plant licensees must satisfy the applicable
provisions of part 140, ``Financial Protection Requirements and
Indemnity Agreements,'' of this chapter.
Subparts K and L [Reserved]
Subpart M--Enforcement
Sec. 53.9000 Violations.
(a) The Commission may obtain an injunction or other court order to
prevent a violation of the provisions of--
(1) The Atomic Energy Act of 1954, as amended (the Act);
(2) Title II of the Energy Reorganization Act of 1974, as amended;
or
(3) A regulation or order issued under those Acts.
(b) The Commission may obtain a court order for the payment of a
civil penalty imposed under Section 234 of the Act:
(1) For violations of--
(i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of
the Act;
(ii) Section 206 of the Energy Reorganization Act of 1974, as
amended;
(iii) Any rule, regulation, or order issued under the sections
specified in paragraph (b)(1)(i) of this section;
(iv) Any term, condition, or limitation of any license issued under
the sections specified in paragraph (b)(1)(i) of this section.
(2) For any violation for which a license may be revoked under
section 186 of the Act.
Sec. 53.9010 Criminal penalties.
(a) Section 223 of the Act provides for criminal sanctions for
willful violation of, attempted violation of, or conspiracy to violate,
any regulation issued under sections 161b, 161i, or 161o of the Act.
For purposes of section 223, all the regulations in part 53 are issued
under one or more of sections 161b, 161i, or 161o, except for the
sections listed in paragraph (b) of this section.
(b) The regulations in 10 CFR part 53 that are not issued under
sections 161b, 161i, or 161o for the purposes of section 223 are as
follows: Sec. Sec. 53.000, 53.015, 53.020, 53.040, 53.080, 53.090,
53.100, 53.110, 53.120, 53.600, 53.725, 53.726, 53.735, 53.760, 53.775,
53.790, 53.795, 53.820, 53.910, 53.1000, 53.1050, 53.1100, 53.1103,
53.1106, 53.1109, 53.1112, 53.1115, 53.1118, 53.1120, 53.1121, 53.1124,
53.1140, 53.1143, 53.1144, 53.1146, 53.1149, 53.1155, 53.1158, 53.1164,
53.1170, 53.1173, 53.1176, 53.1179, 53.1188, 53.1200, 53.1203, 53.1206,
53.1209, 53.1210, 53.1212, 53.1215, 53.1218, 53.1221, 53.1230, 53.1236,
53.1239, 53.1241, 53.1242, 53.1245, 53.1248, 53.1251, 53.1254, 52.1257,
52.1260, 53.1263, 53.1270, 53.1273, 53.1276, 53.1279, 53.1282, 53.1285,
53.1286, 53.1287, 53.1288, 53.1291, 53.1293, 53.125, 53.1300, 53.1306,
53.1309, 53.1312, 53.1315, 53.1318, 53.1324, 53.1330, 53.1333, 53.1336,
53.1348, 53.1360, 53.1366, 53.1369, 53.1372, 53.1375, 53.1381, 53.1384,
53.1387, 53.1390, 53.1396, 53.1401, 53.1405, 53.1410, 53.1416, 53.1419,
53.1422, 53.1425, 53.1431, 53.1437, 53.1440, 53.1443, 53.1452, 53.1455,
53.1456, 53.1458, 53.1461, 53.1470, 53.1500, 53.1510, 53.1515, 53.1520,
53.1525, 53.1530, 53.1535, 53.1540, 53.1560, 53.1585, 53.1590, 53.1595,
53.1600, 53.1660, 53.1670, 53.1700, 53.1710, 53.1730, 53.9000, 53.9010.
PART 70--DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL
0
129. The authority citation for 10 CFR part 70 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 51, 53, 57(d), 108,
122, 161, 182, 183, 184, 186, 187, 193, 223, 234, 274, 1701 (42
U.S.C. 2071, 2073, 2077(d), 2138, 2152, 2201, 2232, 2233, 2234,
2236, 2237, 2243, 2273, 2282, 2021, 2297f); Energy Reorganization
Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846,
5851); Nuclear Waste Policy Act of 1982, secs. 135, 141 (42 U.S.C.
10155, 10161); 44 U.S.C. 3504 note.
Sec. 70.20a [Amended]
0
130. In Sec. 70.20a, in paragraph (b) remove the phrase ``parts 30
through 36, 39, 40, 50, 72, 110,'' and add in its place the phrase
``parts 30 through 36, 39, 40, 50, 53, 72, 110''.
Sec. 70.22 [Amended]
0
131. In Sec. 70.22, wherever it appears, remove the phrase ``part 50''
and add in its place the phrase ``parst 50 or 53''.
0
132. In Sec. 70.24, revise paragraph (d) to read as follows:
Sec. 70.24 Criticality accident requirements.
* * * * *
(d)(1) The requirements in paragraphs (a) through (c) of this
section do not apply to a holder of a construction permit or operating
license for a nuclear power reactor issued under part 50 or part 53 of
this chapter or a combined license issued under part 52 or part 53 of
this chapter, if the holder complies with the requirements of paragraph
(b) of 10 CFR 50.68 or paragraph (m)(2) of 10 CFR 53.440,as applicable.
(2) An exemption from Sec. 70.24 held by a licensee who thereafter
elects to comply with requirements of paragraph (b) of 10 CFR 50.68 or
paragraph (m)(2) of 10 CFR 53.440 does not exempt that licensee from
complying with any of the requirements in Sec. 50.68 or Sec.
53.440(m) of this chapter but shall be ineffective so long as the
licensee elects to comply with Sec. 50.68(b) or Sec. 53.440(m)(2) of
this chapter, as applicable.
Sec. 70.32 [Amended]
0
133. In Sec. 70.32, in paragraph (c)(1) introductory text, remove the
phrase ``part 50 of this chapter'' and add in its place the phrase
``parts 50 or 53 of this chapter''; and in paragraph (d) remove the
phrase ``or Sec. 70.34 of this chapter, as appropriate.'' and add in
its place the phrase ``, Sec. Sec. 74.34 or 53.1510 of this chapter,
as appropriate.''.
[[Page 87117]]
0
134. In Sec. 70.50, revise paragraph (d) to read as follows:
Sec. 70.50 Reporting requirements.
* * * * *
(d) The provisions of Sec. 70.50 do not apply to licensees subject
to Sec. Sec. 50.72 or 53.1630 of this chapter. They do apply to those
10 CFR parts 50 or 53 licensees possessing material licensed under 10
CFR part 70 that are not subject to the notification requirements in
Sec. Sec. 50.72 or 53.1630 of this chapter.
PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF
SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-
RELATED GREATER THAN CLASS C WASTE
0
135. The authority citation for 10 CFR part 72 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 51, 53, 57, 62, 63,
65, 69, 81, 161, 182, 183, 184, 186, 187, 189, 223, 234, 274 (42
U.S.C. 2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2210e,
2232, 2233, 2234, 2236, 2237, 2238, 2273, 2282, 2021); Energy
Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C.
5841, 5842, 5846, 5851); National Environmental Policy Act of 1969
(42 U.S.C. 4332); Nuclear Waste Policy Act of 1982, secs. 117(a),
132, 133, 134, 135, 137, 141, 145(g), 148, 218(a) (42 U.S.C.
10137(a), 10152, 10153, 10154, 10155, 10157, 10161, 10165(g), 10168,
10198(a)); 44 U.S.C. 3504 note.
0
136. In Sec. 72.3, revise the definition for ``Independent spent fuel
storage installation or ISFSI'' to read as follows:
Sec. 72.3 Definitions.
* * * * *
Independent spent fuel storage installation or ISFSI means a
complex designed and constructed for the interim storage of spent
nuclear fuel, solid reactor-related GTCC waste, and other radioactive
materials associated with spent fuel and reactor-related GTCC waste
storage. An ISFSI which is located on the site of another facility
licensed under this part or a facility licensed under part 50 or part
53 of this chapter and which shares common utilities and services with
that facility or is physically connected with that other facility may
still be considered independent.
* * * * *
0
137. In Sec. 72.30, revise paragraph (e)(5) to read as follows:
Sec. 72.30 Financial assurance and recordkeeping for decommissioning.
* * * * *
(e) * * *
(5) In the case of licensees who are issued a power reactor license
under parts 50 or 53 of this chapter or ISFSI licensees who are an
electric utility, as defined in parts 50 or 53 of this chapter, with a
specific license issued under this part, the methods of Sec. Sec.
50.75(b), (e), and (h) or 53.1010, 53.1040, 53.1045(b), and 53.1060 of
this chapter, as applicable. In the event that funds remaining to be
placed into the licensee's ISFSI decommissioning external sinking fund
are no longer approved for recovery in rates by a competent rate making
authority, the licensee must make changes to provide financial
assurance using one or more of the methods stated in paragraphs (a)(1)
through (4) of this section.
* * * * *
0
138. In Sec. 72.32, revise paragraph (c)(2) to read as follows:
Sec. 72.32 Emergency plan.
* * * * *
(c) * * *
(2)(i) Located within the exclusion area as defined in 10 CFR part
100, of a nuclear power reactor licensed for operation by the
Commission, the emergency plan that meets either the requirements in
Sec. 50.160 of this chapter or the requirements in appendix E to part
50 of this chapter and Sec. 50.47(b) of this chapter shall be deemed
to satisfy the requirements of this section.
(ii) Located within the exclusion area, as defined in 10 CFR part
53, of a commercial nuclear plant licensed for operation by the
Commission, the emergency plan that meets either the requirements in
Sec. 50.160 of this chapter or the requirements in appendix E to part
50 of this chapter and Sec. 50.47(b) of this chapter shall be deemed
to satisfy the requirements of this section.
* * * * *
Sec. 72.40 [Amended]
0
139. In Sec. 72.40, in paragraph (c) remove the phrase ``under part 50
of this chapter,'' and add in its place the phrase ``under parts 50 or
53 of this chapter,''.
0
140. In Sec. 72.75, revise paragraph (i)(1)(ii) to read as follows:
Sec. 72.75 Reporting requirements for specific events and conditions.
* * * * *
(i) * * *
(1) * * *
(ii) Licensees issued a general license under Sec. 72.210, after
the licensee has placed spent fuel on the ISFSI storage pad (if the
ISFSI is located inside the collocated protected area, for a reactor
licensed under parts 50 or 53 of this chapter) or after the licensee
has transferred spent fuel waste outside the reactor licensee's
protected area to the ISFSI storage pad (if the ISFSI is located
outside the collocated protected area, for a reactor licensed under
parts 50 or 53 of this chapter).
* * * * *
Sec. 72.184 [Amended]
0
141. In Sec. 72.184, in paragraph (a) remove the phrase ``under part
50 of this chapter'' and add in its place the phrase ``under parts 50
or 53 of this chapter''.
0
142. Revise Sec. 72.210 to read as follows:
Sec. 72.210 General license issued.
A general license is hereby issued for the storage of spent fuel in
an independent spent fuel storage installation at power reactor sites
to persons authorized to possess or operate nuclear power reactors
under 10 CFR parts 50, 52, or 53.
0
143. In Sec. 72.212, revise paragraph (b)(8) to read as follows:
Sec. 72.212 Conditions of general license issued under Sec. 72.210.
* * * * *
(b) * * *
(8) Before use of the general license, determine whether activities
related to storage of spent fuel under this general license involve a
change in the facility Technical Specifications or require a license
amendment for the facility pursuant to Sec. Sec. 50.59(c) or 53.1550
of this chapter. Results of this determination must be documented in
the evaluations made in paragraph (b)(5) of this section.
* * * * *
0
144. In Sec. 72.218, revise paragraphs (a) and (b) to read as follows:
Sec. 72.218 Termination of licenses.
(a) The notification regarding the program for the management of
spent fuel at the reactor required by Sec. Sec. 50.54(bb) or 53.1060
of this chapter must include a plan for removal of the spent fuel
stored under this general license from the reactor site. The plan must
show how the spent fuel will be managed before starting to decommission
systems and components needed for moving, unloading, and shipping this
spent fuel.
(b) An application for termination of a reactor operating license
issued under 10 CFR part 50 and submitted under Sec. 50.82 of this
chapter, or a combined license issued under 10 CFR part 52 and
submitted under Sec. 52.110 of this chapter, or a reactor operating or
combined license under 10 CFR part 53 and submitted under Sec. 53.1070
of this chapter must contain a description of how the spent fuel stored
under this
[[Page 87118]]
general license will be removed from the reactor site.
* * * * *
PART 73--PHYSICAL PROTECTION OF PLANTS AND MATERIALS
0
145. The authority citation for 10 CFR part 73 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 53, 147, 149, 161,
161A, 170D, 170E, 170H, 170I, 223, 229, 234, 1701 (42 U.S.C. 2073,
2167, 2169, 2201, 2201a, 2210d, 2210e, 2210h, 2210i, 2273, 2278a,
2282, 2297f); Energy Reorganization Act of 1974, secs. 201, 202 (42
U.S.C. 5841, 5842); Nuclear Waste Policy Act of 1982, secs. 135, 141
(42 U.S.C. 10155, 10161); 44 U.S.C. 3504 note.
Section 73.37(b)(2) also issued under sec. 301, Pub. L. 96-295,
94 Stat. 789 (42 U.S.C. 5841 note).
0
146. In Sec. 73.1, revise paragraph (b)(1)(i) to read as follows:
Sec. 73.1 [Amended]
* * * * *
(b) * * *
(1) * * *
(i) The physical protection of production and utilization
facilities licensed under parts 50, 52, or 53 of this chapter,
* * * * *
0
147. In Sec. 73.2, revise the introductory text and paragraph (a) to
read as follows:
Sec. 73.2 Definitions.
As used in this part:
(a) Terms defined in parts 50, 52, 53, 70, and 95 of this chapter
have the same meaning when used in this part.
* * * * *
0
148. In Sec. 73.8, revise paragraph (b) to read as follows:
Sec. 73.8 Information collection requirements: OMB approval.
* * * * *
(b) The approved information collection requirements contained in
this part appear in Sec. Sec. 73.5, 73.15, 73.17, 73.20, 73.21, 73.24,
73.25, 73.26, 73.27, 73.37, 73.40, 73.45, 73.46, 73.50, 73.54, 73.55,
73.56, 73.57, 73.58, 73.60, 73.67, 73.70, 73.72, 73.73, 73.74, 73.77,
73.100, 73.110, 73.120, 73.1200, 73.1205, 73.1210, 73.1215, and
appendices B and C to this part.
* * * * *
0
149. In Sec. 73.50, revise the introductory text to read as follows:
Sec. 73.50 Requirements for physical protection of licensed
activities.
Each licensee who is not subject to Sec. 73.51, but who possesses,
uses, or stores formula quantities of strategic special nuclear
material that are not readily separable from other radioactive material
and which have a total external radiation level in excess of 1 gray
(100 rad) per hour at a distance of 1 meter (3.3 feet) from any
accessible surfaces without intervening shielding other than at a
nuclear reactor facility licensed under parts 50, 52, or 53 of this
chapter, shall comply with the following:
* * * * *
0
150. In Sec. 73.55, revise paragraphs (a)(4) and (6), (i)(4)(iii),
(l)(1), (l)(7)(ii), (p)(1)(i), (r)(2) and (r)(4)(iii) to read as
follows:
Sec. 73.55 Requirements for physical protection of licensed
activities in nuclear power reactors against radiological sabotage.
(a) * * *
(4) Applicants for an operating license under the provisions of
part 50 or part 53 of this chapter or holders of a combined license
under the provisions of part 52 or part 53 of this chapter shall
implement the requirements of this section before fuel is allowed
onsite (protected area).
* * * * *
(6) Applicants for an operating license under the provisions of
part 50 or part 53 of this chapter, or holders of a combined license
under the provisions of part 52 or part 53 of this chapter that do not
reference a standard design certification or reference a standard
design certification issued after May 26, 2009, shall meet the
requirement of Sec. 73.55(i)(4)(iii).
* * * * *
(i) * * *
(4) * * *
(iii) Applicants for an operating license under the provisions of
part 50 of this chapter, or holders of a combined license under the
provisions of part 52 of this chapter, or licensees under part 53 of
this chapter that elect to demonstrate compliance with Sec. 73.55,
consistent with Sec. 53.860(a)(2) of this chapter, shall construct,
locate, protect, and equip both the central and secondary alarm
stations to the standards for the central alarm station contained in
this section. Both alarm stations shall be equal and redundant, such
that all functions needed to satisfy the requirements of this section
can be performed in both alarm stations.
* * * * *
(l) * * *
(1) Commercial nuclear power reactors licensed under 10 CFR parts
50, 52, or 53 and authorized to use special nuclear material in the
form of MOX fuel assemblies containing up to 20 weight percent
PuO2 shall, in addition to demonstrating compliance with the
requirements of this section, protect un-irradiated MOX fuel assemblies
against theft or diversion as described in this paragraph.
* * * * *
(7) * * *
(ii) Additional measures for the physical protection of un-
irradiated MOX fuel assemblies containing greater than 20 weight
percent PuO2 shall be determined by the Commission on a
case-by-case basis and documented through license amendment in
accordance with Sec. Sec. 50.90 or 53.1510 of this chapter.
* * * * *
(p) * * *
(1) * * *
(i) Under Sec. Sec. 50.54(x) and (y) or 53.740(h) of this chapter,
the licensee may suspend any security measures under this section in an
emergency when this action is immediately needed to protect the public
health and safety and no action consistent with license conditions and
technical specifications that can provide adequate or equivalent
protection is immediately apparent. This suspension of security
measures must be approved as a minimum by a licensed senior operator
before taking this action.
* * * * *
(r) * * *
(2) The licensee shall submit proposed alternative measure(s) to
the Commission for review and approval under Sec. Sec. 50.4 and 50.90,
or Sec. Sec. 53.040 and 53.1510 of this chapter before implementation.
* * * * *
(4) * * *
(iii) Based on comparison of the costs of the alternative measures
to the costs of demonstrating compliance with the Commission's
requirements using the essential elements of Sec. Sec. 50.109 or
53.1590 of this chapter, the costs of fully demonstrating compliance
with the Commission's requirements are not justified by the protection
that would be provided.
0
151. In Sec. 73.56, revise paragraph (a)(3) to read as follows:
Sec. 73.56 Personnel access authorization requirements for nuclear
power plants.
(a) * * *
(3) Each applicant for an operating license under the provisions of
part 50 of this chapter, each holder of a combined license under the
provisions of part 52 of this chapter, and applicants for an operating
license or holders of a combined license under part 53 of this chapter
that do not meet the requirements of Sec. 53.860(a)(2) of this
chapter, shall implement the
[[Page 87119]]
requirements of this section before fuel is allowed on site (protected
area).
* * * * *
0
152. In Sec. 73.57, revise paragraph (a)(3) to read as follows:
Sec. 73.57 Requirements for criminal history records checks of
individuals granted unescorted access to a nuclear power facility, a
non-power reactor, or access to Safeguards Information.
(a) * * *
(3) Before receiving its operating license under 10 CFR parts 50 or
53 or before the Commission makes its finding under Sec. Sec.
52.103(g) or 53.1452(g) of this chapter, each applicant for a license
to operate a nuclear power reactor (including an applicant for a
combined license) or a non-power reactor may submit fingerprints for
those individuals who will require unescorted access to the nuclear
power facility or non-power reactor facility.
* * * * *
0
153. In Sec. 73.58, revise paragraph (a) to read as follows:
Sec. 73.58 Safety/security interface requirements for nuclear power
reactors.
(a) Each operating nuclear power reactor licensee with a license
issued under parts 50, 52, or 53 of this chapter shall comply with the
requirements of this section.
* * * * *
0
154. In Sec. 73.67, revise paragraphs (d) introductory text and (f)
introductory text to read as follows:
Sec. 73.67 Licensee fixed site and in-transit requirements for the
physical protection of special nuclear material of moderate and low
strategic significance.
* * * * *
(d) Fixed site requirements for special nuclear material of
moderate strategic significance. Each licensee who possesses, stores,
or uses quantities and types of special nuclear material of moderate
strategic significance at a fixed site or contiguous sites, except as
allowed by paragraph (b)(2) of this section and except those who are
licensed to operate a nuclear power reactor pursuant to part 50 or part
53, provided that the special nuclear material is located within a
protected area and protected under Sec. 73.55 or Sec. 73.100, shall:
* * * * *
(f) Fixed site requirements for special nuclear material of low
strategic significance. Each licensee who possesses, stores, or uses
special nuclear material of low strategic significance at a fixed site
or contiguous sites, except those who are licensed to operate a nuclear
power reactor pursuant to part 50 or part 53, provided that the special
nuclear material is located within a protected area and protected under
Sec. 73.55 or Sec. 73.100, shall:
* * * * *
0
155. In Sec. 73.77, revise paragraphs (a), (b)(1), (c)(6) and (7) to
read as follows:
Sec. 73.77 Cybersecurity event notifications.
(a) Each licensee subject to the provisions of Sec. Sec. 73.54 or
73.110 shall notify the NRC Headquarters Operations Center via the
Emergency Notification System (ENS), under paragraph (c) of this
section:
(1) Within one hour after discovery of a cyberattack that adversely
impacted:
(i) Safety-related or important-to-safety functions, security
functions, or emergency preparedness functions (including offsite
communications); or that compromised support systems and equipment
resulting in adverse impacts to safety, security, or emergency
preparedness functions within the scope of Sec. 73.54; or,
(ii) Functions performed by digital assets that would prevent a
postulated fission product release resulting in offsite doses exceeding
the values in Sec. 53.210 of this chapter, or functions performed by
digital assets used by the licensee for implementing the physical
security requirements in Sec. 53.860(a) of this chapter.
(2) Within 4 hours:
(i) After discovery of a cyberattack that could have caused an
adverse impact to:
(A) Safety-related or important-to-safety functions, security
functions, or emergency preparedness functions (including offsite
communications); or that could have compromised support systems and
equipment, which if compromised, could have adversely impacted safety,
security, or emergency preparedness functions within the scope of Sec.
73.54; or,
(B) Functions performed by digital assets that would prevent a
postulated fission product release resulting in offsite doses exceeding
the values in Sec. 53.210 of this chapter, or functions performed by
digital assets used by the licensee for implementing the physical
security requirements in Sec. 53.860(a) of this chapter.
(ii) After discovery of a suspected or actual cyberattack initiated
by personnel with physical or electronic access to digital computer and
communication systems and networks within the scope of Sec. Sec. 73.54
or 73.110.
(iii) After notification of a local, State, or other Federal agency
(e.g., law enforcement, Federal Bureau of Investigation (FBI), etc.) of
an event related to the licensee's implementation of their
cybersecurity program for digital computer and communication systems
and networks within the scope of Sec. Sec. 73.54 or 73.110 that does
not otherwise require a notification under paragraph (a) of this
section.
(3) Within 8 hours after receipt or collection of information
regarding observed behavior, activities, or statements that may
indicate intelligence gathering or pre-operational planning related to
a cyberattack against digital computer and communication systems and
networks within the scope of Sec. Sec. 73.54 or 73.110.
(b) Twenty-four hour recordable events. (1) The licensee shall use
the site corrective action program to record vulnerabilities,
weaknesses, failures and deficiencies in their Sec. 73.54 or Sec.
73.110 cybersecurity program within 24 hours of their discovery.
* * * * *
(c) * * *
(6) Declaration of emergencies. Notifications made to the NRC for
the declaration of an emergency class shall be performed in accordance
with Sec. Sec. 50.72 or 53.1630 of this chapter, as applicable.
(7) Elimination of duplication. Separate notifications and reports
are not required for events that are also reportable under Sec. Sec.
50.72 and 50.73 or Sec. Sec. 53.1630 and 53.1640 of this chapter.
However, these notifications should also indicate the applicable Sec.
73.77 reporting criteria.
* * * * *
0
156. Add Subpart J consisting of Sec. Sec. 73.100 through 73.120 to
read as follows:
Subpart J--Security Requirements at Commercial Nuclear Plants
Sec.
73.100 Technology-inclusive requirements for physical protection of
licensed activities at commercial nuclear plants against
radiological sabotage.
73.110 Technology-inclusive requirements for protection of digital
computer and communication systems and networks.
73.120 Access authorization program for commercial nuclear plants.
Subpart J--Security Requirements at Commercial Nuclear Plants
Sec. 73.100 Technology-inclusive requirements for physical protection
of licensed activities at commercial nuclear plants against
radiological sabotage.
(a) Introduction. (1) Each licensee that is licensed to operate a
commercial nuclear plant under 10 CFR part 53 and elects to implement
the requirements of this section must do so through its physical
security plan, training and qualification plan, safeguards
[[Page 87120]]
contingency plan, and cybersecurity plan, referred to collectively
hereafter as ``security plans,'' before initial fuel load into the
reactor (or, for a fueled manufactured reactor, before initiating the
physical removal of any one of the independent physical mechanisms to
prevent criticality required under Sec. 53.620(d)(1) of this chapter).
(2) The security plans must identify, describe, and account for
site-specific conditions that affect the licensee's capability to
satisfy the requirements of this section.
(b) General performance objective and requirements. (1) The
licensee must establish, implement, and maintain a physical protection
program and a security organization, which will have as their objective
to provide reasonable assurance that activities involving special
nuclear material are not inimical to the common defense and security
and do not constitute an unreasonable risk to the public health and
safety.
(2) To satisfy the general performance objective of paragraph
(b)(1) of this section, the physical protection program must protect
against the design basis threat of radiological sabotage as stated in
Sec. 73.1. Specifically, the licensee must--
(i) Ensure that the physical protection program capabilities to
protect against the design basis threat of radiological sabotage are
maintained at all times; and
(ii) Provide defense in depth in achieving performance requirements
through the integration of engineered systems, administrative controls,
and management measures.
(3) The physical protection program must be designed and
implemented to achieve and maintain the reliability and availability of
structures, systems, and components (SSCs) required for demonstrating
compliance with the following performance requirements at all times:
(i) Intrusion detection. The licensee must be capable of detecting
attempted and actual unauthorized access to interior and exterior areas
containing SSCs needed to implement safety and security functions.
(ii) Intrusion assessment. The licensee must be capable of timely
assessment for determining the cause of a detected intrusion.
(iii) Security communication. The licensee must be capable of
continuous security communications. Communication systems must account
for design basis threats that can interrupt or interfere with
continuity or integrity of communications.
(iv) Security response. The physical protection program must be
designed to provide timely security response to interdict and
neutralize adversary attacks up to and including the design basis
threat of radiological sabotage. The physical protection program must
be designed to provide layers of security response, with each layer
assuring that a single failure does not result in the loss of
capability to neutralize the design basis threat adversary. Structures,
systems, and components relied on for delay functions must be designed
to allow for timely security responses to adversary attacks with
adequate defense in depth.
(A) The security response may rely on the use of onsite responders,
law enforcement or other offsite armed responders, or a combination
thereof, to fulfill the interdiction and neutralization functions
required by paragraph (b)(3)(iv) of this section. A licensee relying
entirely or partially on law enforcement or other offsite armed
responders must--
(1) Maintain the capability to detect, assess, interdict, and
neutralize threats as required by paragraphs (b)(3)(i), (b)(3)(ii), and
(b)(3)(iv) of this section;
(2) Provide adequate delay to enable law enforcement or other
offsite armed responders to fulfill the interdiction and neutralization
functions for threats up to and including the design basis threat of
radiological sabotage;
(3) Provide necessary information about the facility and make
available periodic training to law enforcement or other offsite armed
responders who will fulfill the interdiction and neutralization
functions for threats up to and including the design basis threat of
radiological sabotage;
(4) Fully describe in the safeguards contingency plan the role that
law enforcement or other offsite armed responders will play in the
licensee's protective strategy. The description must provide sufficient
detail to enable the NRC to determine that the licensee's physical
protection program provides reasonable assurance of adequate protection
against threats up to and including the design basis threat of
radiological sabotage; and
(5) Identify criteria and measures to compensate for the
degradation or absence of law enforcement or other offsite armed
responders and propose suitable compensatory measures that meet the
requirements of paragraph (h)(3) of this section to address this
degradation.
(B) For licensees relying entirely or partially on law enforcement
responders to fulfill the interdiction and neutralization functions
required by paragraph (b)(3)(iv) of this section, the training and
qualification requirements related to armed response personnel in
paragraphs (c) and (e) of this section do not apply to law enforcement
responders. The licensee shall continue to satisfy the performance
evaluation requirements in paragraph (g) of this section for all armed
response personnel, including law enforcement.
(v) Protecting against land and waterborne vehicle bomb assaults.
The licensee must be capable of protecting the plant against the design
basis threat vehicle bomb assault. The methods that are relied on to
protect against a design basis threat land vehicle and waterborne
vehicle bomb assault must be designed to protect the reactor building
and structures containing safety- or security-related systems, and
components from explosive effects.
(vi) Access control portals. The licensee must be capable of
detecting and denying unauthorized access to persons and pass-through
of contraband materials (e.g., weapons, incendiaries, explosives) to
protected areas.
(4) The licensee must meet the requirements related to target sets
in Sec. 73.55(f).
(5) The licensee must identify and analyze site-specific
conditions, including target sets, that may affect the physical
protection program needed to implement the requirements of this
section. The licensee must account for these conditions in
demonstrating compliance with the requirements of this section.
(6) The licensee must establish, implement, and maintain a
performance evaluation program to assess the effectiveness of the
licensee's implementation of the physical protection program to protect
against the design basis threat of radiological sabotage.
(7) The licensee must establish, implement, and maintain an access
authorization program under Sec. 73.56 and must describe the program
in the physical security plan.
(8) The licensee must establish, implement, and maintain a
cybersecurity program under Sec. Sec. 73.54 or 73.110 and must
describe the program in the cybersecurity plan.
(9) The licensee must establish, implement, and maintain an insider
mitigation program and must describe the program in the physical
security plan.
(i) The insider mitigation program must monitor the initial and
continuing trustworthiness and reliability of individuals granted or
retaining unescorted access or unescorted access authorization to a
protected or vital area, and implement defense-in-depth methodologies
to minimize the potential for an insider (active, passive, or both)
[[Page 87121]]
to adversely affect, either directly or indirectly, the licensee's
capability to protect against radiological sabotage.
(ii) The insider mitigation program must integrate elements of--
(A) The access authorization program under Sec. 73.56;
(B) The fitness-for-duty program under 10 CFR part 26;
(C) The cybersecurity program under Sec. Sec. 73.54 or 73.110; and
(D) The physical protection program under this section.
(10) The licensee must have the capability to track, trend,
correct, and prevent recurrence of failures and deficiencies in the
implementation of the requirements of this section.
(11) Implementation of security plans and associated procedures
must be coordinated with other onsite plans and procedures to preclude
conflict during both normal and emergency conditions and ensure the
adequate management of the safety and security interface.
(12)(i) The licensee must ensure that the firearms background check
requirements of Sec. 73.17 of this part are met for all members of the
security organization whose official duties require access to covered
weapons or who inventory enhanced weapons.
(ii) The provisions of this paragraph are only applicable to
licensees subject to this section that are also subject to the firearms
background check provisions of Sec. 73.17 of this part.
(c) Security organization. The licensee must establish and maintain
a security organization that is staffed, trained, qualified, and
equipped to implement the physical protection program under the
requirements of this section.
(1) The licensee must establish a management system for maintaining
and implementing security policies and procedures to implement the
requirements of this section and the security plans.
(2) Implementing procedures must document the conduct of security
operations, security design and configuration controls, maintenance,
training and qualification, and contingency responses.
(3) The licensee must--
(i) Establish a process for the approval of designs, policies,
processes, and procedures and changes by the individual with overall
responsibility for the physical protection program; and
(ii) Ensure that revisions and changes to the physical protection
program and implementing policies, processes, and procedures satisfy
the requirements of this section.
(4) The licensee must retain, in accordance with Sec. 73.70, all
analyses, assessments, calculations, and descriptions of the technical
basis for demonstrating compliance with the performance requirements of
Sec. 73.100(b). The licensee must protect these records in accordance
with the requirements for protecting safeguards information in
Sec. Sec. 73.21 and 73.22.
(5) The licensee may not permit any individual to implement any
part of the physical protection program unless the individual has been
trained, equipped, and qualified to perform their assigned duties and
responsibilities in accordance with the training and qualification
plan.
(d) Search requirements. The licensee must establish and implement
searches of individuals, vehicles, and materials to detect and prevent
the introduction into the protected area of firearms, explosives,
incendiary devices, or other items and material which could be used to
commit radiological sabotage.
(e) Training and qualification program. The licensee must establish
and maintain a training and qualification program that ensures
personnel who are responsible for the physical protection of the
facility against radiological sabotage are able to effectively perform
their assigned security-related job duties for implementing the
requirements of this section and must describe the program in the
training and qualification plan.
(f) Security reviews. The licensee must establish and implement
security reviews to assess the effectiveness of the implementation of
the physical protection program. Security reviews must be performed by
individuals independent of those personnel responsible for program
management and any individual who has direct responsibility for
implementing the onsite physical protection program.
(1) The licensee must review each element of the physical
protection program at a frequency commensurate with the importance or
significance to safety of plant operations to ensure timely
identification and documentation of vulnerabilities, improvements, and
corrective actions. The objective of these reviews must be maintaining
effective implementation of the engineered and administrative controls
required to achieve the physical protection program functions and the
management system required to implement programs and requirements in
this section.
(2) The licensee must establish and perform self-assessments to
ensure the effective implementation of the physical protection program
functions of detection, assessment, communication, delay, and
interdiction and neutralization to protect against the design basis
threat of radiological sabotage. The licensee must perform design
verification and assessments of the capabilities of active and passive
engineering systems relied on to protect against the design basis
threat.
(3) Reviews of the security program must include, but are not
limited to, an audit of the effectiveness of the physical protection
program, security plans, implementing procedures, cybersecurity
programs, safety/security interface activities, the testing,
maintenance, and calibration program, and response commitments by
local, State, and Federal law enforcement authorities.
(4) The results and recommendations of the onsite physical
protection program reviews, management's findings regarding program
effectiveness, and any actions taken as a result of recommendations
from prior program reviews, must be documented in a report and must be
maintained in an auditable form and available for inspection.
(g) Performance evaluation. Licensee performance evaluations must
include methods appropriate and necessary to assess, test, and
challenge the integration of the physical protection program's
functions to protect against the design basis threat, including
measures to protect against cyberattack and engineered systems designed
to protect against the design basis threat standalone ground vehicle
bomb attack.
(1) The licensee must establish the frequencies for performance
evaluations commensurate with the security significance of the physical
protection program.
(2) The licensee must document processes and procedures for
implementing the performance evaluations. The licensee must maintain
records, including results, findings, and corrective actions identified
during the performance evaluations.
(h) Maintenance, testing, and calibration and corrective actions.
(1) The licensee must ensure that security SSCs, including supporting
systems, are inspected, tested, and calibrated for operability and
performance at intervals necessary and sufficient to meet the
requirements of this section.
(2) The licensee must implement corrective actions to ensure
resolution of identified vulnerabilities and deficiencies to meet the
requirements of this section.
(3) The licensee must establish and implement timely compensatory
measures for degraded or inoperable security SSCs to meet the
requirements of this section. Compensatory measures must provide a
level of protection that is equivalent to the protection that was
provided prior to the degradation or
[[Page 87122]]
inoperability of the security structures, systems, or components.
(4) The licensee must document processes and procedures and
maintain records for implementing the corrective actions, compensatory
measures, and maintenance, inspection, testing, and calibration of
security SSCs.
(i) Suspension of security measures. (1) The licensee may suspend
implementation of affected requirements of this section in accordance
with Sec. 53.740(h) of this chapter under the following conditions:
(i) In an emergency, when action is immediately needed to protect
the public health and safety; and
(ii) During severe weather, when the suspension of affected
security measures is immediately needed to protect the personal health
and safety of personnel.
(2) Suspended security measures must be reinstated as soon as
conditions permit.
(3) The suspension of security measures must be reported and
documented in accordance with the provisions of Sec. Sec. 73.1200 and
73.1205.
(j) Records. (1) The Commission may inspect, copy, retain, and
remove all reports, records, and documents required to be kept by
Commission regulations, orders, or license conditions, whether the
reports, records, and documents are kept by the licensee or a
contractor.
(2) The licensee must maintain all records required to be kept by
Commission regulations, orders, or license conditions, until the
Commission terminates the license for which the records were developed
and must maintain superseded portions of these records for at least 3
years after the record is superseded, unless otherwise specified by the
Commission.
(3) If a contracted security force is used to implement the onsite
physical protection program, the licensee's written agreement with the
contractor must be retained by the licensee as a record for the
duration of the contract.
(4) Review and audit reports must be available for inspection, for
a period of 3 years.
Sec. 73.110 Technology-inclusive requirements for protection of
digital computer and communication systems and networks.
(a) Each licensee that is licensed to operate a commercial nuclear
plant under 10 CFR part 53 and elects to implement the requirements of
this section must establish, implement, and maintain a cybersecurity
program that is commensurate with the potential consequences resulting
from cyberattacks, up to and including the design basis threat as
described in Sec. 73.1. The cybersecurity program must provide
reasonable assurance that digital computer and communication systems
and networks are adequately protected against cyberattacks that are
capable of causing the following consequences:
(1) Adversely impacting the functions performed by digital assets
that would prevent a postulated fission product release resulting in
offsite doses exceeding the values in Sec. 53.210 of this chapter.
(2) Adversely impacting the functions performed by digital assets
used by the licensee for implementing the physical security
requirements in Sec. 53.860(a) of this chapter.
(b) To protect digital computer and communication systems and
networks associated with the functions described in paragraphs (a)(1)
and (2), the licensee must--
(1) Analyze the potential consequences resulting from cyberattacks
on digital computer and communication systems and networks and identify
those assets that must be protected to demonstrate compliance with
paragraph (a) of this section; and
(2) Implement the cybersecurity program in accordance with
paragraph (d) of this section.
(c) The licensee must comply with the requirements in Sec.
73.54(a)(2) for the systems and networks identified in paragraph (b)(1)
of this section in a manner that is commensurate with the potential
consequences resulting from cyberattacks.
(d) The cybersecurity program must be designed in a manner that is
commensurate with the potential consequences resulting from
cyberattacks through the following steps:
(1) Implement security controls to protect the assets identified
under paragraph (b)(1) of this section from cyberattacks, commensurate
with their safety and security significance;
(2) Apply and maintain defense-in-depth protective strategies to
ensure the capability to detect, delay, respond to, and recover from
cyberattacks capable of causing the consequences identified in
paragraph (a) of this section;
(3) Mitigate the adverse effects of cyberattacks capable of causing
the consequences identified in paragraph (a) of this section; and
(4) Ensure that the functions of protected assets identified under
paragraph (b)(1) of this section are not adversely impacted due to
cyberattacks.
(e) The licensee must implement the following requirements in a
manner that is commensurate with the potential consequences resulting
from cyberattacks:
(1) As part of the cybersecurity program, the licensee must comply
with the requirements in Sec. 73.54(d)(1), (2), and (4), and must
ensure that modifications to assets, identified under paragraph (b)(1)
of this section are evaluated before implementation to ensure that the
cybersecurity performance objectives identified in paragraph (a) of
this section are maintained.
(2) The licensee must establish, implement, and maintain a
cybersecurity plan that implements the cybersecurity program
requirements of this section.
(i) The cybersecurity plan must describe how the requirements of
this section will be implemented and must account for the site-specific
conditions that affect implementation.
(ii) The cybersecurity plan must include measures for incident
response and recovery for cyberattacks. The cybersecurity plan must
include the analysis identified under paragraph (b)(1) of this section
and describe how the licensee will--
(A) Apply and maintain defense-in-depth protective strategies as
required in paragraph (d)(2) of this section;
(B) Maintain the capability for timely detection and response to
cyberattacks;
(C) Mitigate the consequences of cyberattacks;
(D) Correct exploited vulnerabilities; and
(E) Restore affected systems, networks, and/or equipment affected
by cyberattacks.
(3) The licensee must develop and maintain written policies and
implementing procedures to implement the cybersecurity plan. Policies,
implementing procedures, and other supporting technical information
used by the licensee need not be submitted for Commission review and
approval as part of the cybersecurity plan but are subject to
inspection by NRC staff on a periodic basis.
(4) The licensee must establish and implement cybersecurity reviews
to assess the effectiveness of the implementation of the cybersecurity
program.
(i) The licensee must review each element of the cybersecurity
program at a frequency commensurate with the importance or significance
to safety of plant operations to ensure timely identification and
documentation of vulnerabilities, improvements, and corrective actions.
(ii) Cybersecurity reviews must be performed by individuals
independent
[[Page 87123]]
of those personnel responsible for program management and any
individual who has direct responsibility for implementing the
cybersecurity program.
(iii) The licensee must establish and perform self-assessments to
ensure the effective implementation of the cybersecurity program.
(iv) The results and recommendations of the cybersecurity program
reviews, management's findings regarding program effectiveness, and any
actions taken as a result of recommendations from prior program
reviews, must be documented in a report and must be maintained in an
auditable form and available for inspection.
(5) The licensee must retain all records and supporting technical
documentation required to demonstrate compliance with the requirements
of this section as a record until the Commission terminates the license
for which the records were developed and must maintain superseded
portions of these records for at least three (3) years after the record
is superseded, unless otherwise specified by the Commission.
Sec. 73.120 Access authorization program for commercial nuclear
plants.
(a) Introduction and scope. Each applicant for an operating license
or a holder of a combined license under 10 CFR part 53 must establish,
maintain, and implement an access authorization program before initial
fuel load into the reactor (or, for a fueled manufactured reactor,
before initiating the physical removal of any one of the independent
physical mechanisms to prevent criticality required under Sec.
53.620(d)(1) of this chapter). The requirements in this section apply
to licensees satisfying the criterion in Sec. 53.860(a)(2)(i) of this
chapter.
(b) Applicability. (1) The following individuals must be subject to
an access authorization program under this section:
(i) Any individual to whom a licensee intends to grant unescorted
access to a commercial nuclear plant protected area, vital area, or
controlled access area where licensed material is used or stored;
(ii) Any individual whose duties and responsibilities permit the
individual to take actions by electronic means, either on site or
remotely, that could adversely impact the licensee's or applicant's
operational safety, security, or emergency preparedness;
(iii) Any individual who has responsibilities for implementing a
licensee's or applicant's protective strategy, including armed security
force officers, alarm station operators, and tactical response team
leaders but not including Federal, State, or local law enforcement
personnel; and
(iv) The licensee or applicant access authorization program
reviewing official or contractor or vendor access authorization program
reviewers.
(2) The licensee or applicant may subject other individuals,
including employees of a contractor or a vendor who are designated in
access authorization program procedures, to an access authorization
program that demonstrates compliance with the requirements of this
section.
(c) General performance objectives and requirements. Each
licensee's or applicant's access authorization program under this
section must demonstrate that the individuals who are specified in
paragraph (b) of this section are trustworthy and reliable, such that
they do not constitute an unreasonable risk to public health and safety
or the common defense and security. The licensee's access authorization
program must maintain the capabilities for demonstrating compliance
with the following performance requirements:
(1) Background investigation. (i)(A) Licensees and applicants must
ensure that any individual seeking initial unescorted access or to
maintain unescorted access is subject to a background investigation.
(B) Background investigations must include the program elements
contained under Sec. 37.25 of this chapter and must also include a
credit history evaluation.
(ii) Background investigations must include fingerprinting and an
FBI identification and criminal history records check in accordance
with Sec. 37.27 of this chapter.
(iii) Licensees must have the informed and signed consent of the
subject individual to initiate a background investigation. This consent
must include authorization to share personal information with other
individuals or organizations as necessary to complete the background
investigation. A signed consent must be obtained prior to any
reinvestigation. The subject individual may withdraw his or her consent
at any time. Licensees must inform the individual that--
(A) If an individual withdraws his or her consent, the licensee may
not initiate any elements of the background investigation that were not
in progress at the time the individual withdrew his or her consent; and
(B) The withdrawal of consent for the background investigation is
sufficient cause for denial or termination of unescorted access
authorization.
(2) Behavioral observation. Licensees, applicants, contractors, and
vendors must ensure the access authorization program includes
provisions that the individuals specified in paragraph (b) of this
section are subject to behavioral observation.
(i) Each person subject to behavioral observation must communicate
to the licensee or applicant observed behaviors or activities of
individuals that may constitute an unreasonable risk to the health and
safety of the public and common defense and security.
(ii) Behavioral observation must include visual observation, in
person or remotely by video, to detect and promptly report to plant
supervision any concerns arising from behavioral observation,
including, but not limited to, concerns related to any questionable
behavior patterns or activities of others.
(3) Self-reporting of legal actions. Licensees or applicants must
inform personnel who are granted and who maintain unescorted access of
their responsibilities to self-report to plant supervision legal
actions taken by a law enforcement authority or court of law against
the individual that could result in incarceration or a court order or
that requires a court appearance, including but not limited to an
arrest, an indictment, the filing of charges, or a conviction, but
excluding minor civil actions or misdemeanors such as parking
violations or speeding tickets, for any individual who has applied for
unescorted access or who maintains unescorted access.
(4) Unescorted access. Licensees or applicants must grant
unescorted access only after the licensee has verified an individual is
trustworthy and reliable. A list of persons currently approved for
unescorted access to a protected area, vital area, or controlled access
area must be maintained at all times. Unescorted access determinations
must be reviewed annually by the reviewing official. Licensees and
applicants must complete an FBI criminal history record check update
for each individual maintaining unescorted access, within 10 years of
the last review.
(5) Termination of unescorted access. Licensees and applicants must
promptly terminate unescorted access when this access is no longer
required or a reviewing official determines an individual is no longer
trustworthy and reliable in accordance with this section.
(6) Determination basis for access. (i) The licensee's or
applicant's reviewing official must determine whether to permit, deny,
unfavorably terminate, maintain, or administratively withdraw an
individual's unescorted access based on an evaluation of all of the
[[Page 87124]]
information collected to demonstrate compliance with the requirements
of this section.
(ii) Licensees and applicants must provide individuals subject to
this section, prior to any final adverse determination, the right to
complete, correct, and explain information obtained as a result of the
licensee's background investigation pursuant to Sec. 37.23(g) of this
chapter.
(iii) The licensee's or applicant's reviewing officials are the
only individuals authorized to make unescorted access determination
decisions. Each licensee or applicant must name one or more individuals
to be reviewing officials pursuant to the requirements of Sec.
37.23(b)(2) of this chapter.
(7) Review procedures. Review procedures must be established in
accordance with Sec. 37.23(f) of this chapter, to include provisions
for the notification in writing of individuals who are denied
unescorted access or who are unfavorably terminated.
(8) Protection of information. Licensees, applicants, contractors,
or vendors must establish and maintain a system of files and procedures
in accordance with Sec. 37.31 of this chapter, to ensure personal
information is not disclosed to unauthorized persons.
(9) Access authorization reviews and corrective action. Licensees
and applicants must develop, implement, and maintain procedures for
conduct of access authorization reviews and corrective actions in
accordance with Sec. 37.33 of this chapter to ensure the continuing
effectiveness of the access authorization program and to ensure that
the access authorization program and program elements are in compliance
with the requirements of this section. Each licensee and applicant must
be responsible for the continuing effectiveness of the access
authorization program, including access authorization program elements
that are provided by the contractors or vendors, and the access
authorization programs of any of the contractors or vendors that are
accepted by the licensee or applicant.
(10) Records. Licensees, applicants, and contractors or vendors
must document the processes and procedures for maintaining records used
or created to establish an individual's trustworthiness and reliability
or to document access determinations. Licensees, applicants, and
contractor or vendors must--
(i) Retain documentation regarding the trustworthiness and
reliability of individual employees for 3 years from the date the
individual no longer requires unescorted access;
(ii) Retain a copy of the current access authorization program
procedures as a record for 3 years after the procedure is no longer
needed. If any portion of the procedure is superseded, retain the
superseded material for 3 years after the record is superseded; and
(iii) Retain the list of persons approved for unescorted access for
3 years after the list is superseded or replaced. Records maintained in
any database(s) must be available for NRC review.
0
157. In Sec. 73.1200, revise paragraphs (a) introductory text, (c)(1)
introductory text, (e)(1) introductory text, (e)(3) and (4), (g)(1)
introductory text, (o)(5)(i) and (o)(6)(i), (r) and (s) to read as
follows:
Sec. 73.1200 Notification of physical security events.
(a) 15-minute notifications--facilities. Each licensee subject to
the provisions of Sec. 73.20, Sec. 73.45, Sec. 73.46, Sec. 73.51,
Sec. 73.55, or Sec. 73.100 must notify the NRC Headquarters
Operations Center, as soon as possible but within 15 minutes after--
* * * * *
(c) * * *
(1) Each licensee subject to the provisions of Sec. Sec. 73.20,
73.45, 73.46, 73.50, 73.51, 73.55, 73.60, 73.67, or 73.100 must notify
the NRC Headquarters Operations Center as soon as possible but no later
than 1 hour after the time of discovery of the following significant
facility security events involving--
* * * * *
(e) * * *
(1) Each licensee subject to the provisions of Sec. Sec. 73.20,
73.45, 73.46, 73.50, 73.51, 73.55, 73.60, 73.67, or 73.100 must notify
the NRC Headquarters Operations Center within 4 hours after time of
discovery of the following facility security events involving--
* * * * *
(3)(i) An event involving a law enforcement response to the
facility that could reasonably be expected to result in public or media
inquiries and that does not otherwise require a notification under
paragraphs (a) through (h) of this section, or in other NRC regulations
such as Sec. 50.72(b)(2)(xi) or Sec. 53.1630(b)(2)(v) of this
chapter.
(ii) As an exemption, licensees need not report law enforcement
responses to minor incidents, such as traffic accidents.
(4) For licensees subject to the provisions of Sec. 73.55 or Sec.
73.100 of this part, an event involving the licensee's suspension of
security measures.
* * * * *
(g) * * *
(1) Each licensee subject to the provisions of Sec. 73.20, Sec.
73.45, Sec. 73.46, Sec. 73.50, Sec. 73.51, Sec. 73.55, Sec. 73.60,
Sec. 73.67, or Sec. 73.100 must notify the NRC Headquarters
Operations Center within 8 hours after time of discovery of the
following facility security program failures involving--
* * * * *
(o) * * *
(5) * * *
(i) Licensees must establish the requested continuous
communications channel once the licensee has completed other required
notifications under this section, Sec. 50.72 of this chapter, appendix
E to part 50 of this chapter, Sec. 53.1630 of this chapter, Sec.
70.50 of this chapter; or Sec. 72.75 of this chapter; as appropriate.
* * * * *
(6) * * *
(i) Licensees must establish the requested continuous
communications channel once the licensee or the movement control center
has completed other required notifications under this section, Sec.
50.72 of this chapter, appendix E to part 50 of this chapter, Sec.
53.1630 of this chapter, Sec. 70.50 of this chapter; Sec. 72.75 of
this chapter; or requested assistance from the LLEA, as appropriate.
* * * * *
(r) Declaration of emergencies. Licensees notifying the NRC of the
declaration of an emergency class must do so in accordance with
Sec. Sec. 50.72, 53.1630, 63.73, 70.50, and 72.75 of this chapter, as
applicable.
(s) Elimination of duplication. Licensees with notification
obligations under paragraphs (a) through (h), (m), and (n) of this
section and Sec. Sec. 50.72, 53.1630, 63.73, 70.50, and 72.75 of this
chapter may notify the NRC of events in a single communication. This
communication must identify each regulation under which the licensee is
reporting.
* * * * *
0
158. In Sec. 73.1205, revise paragraph (b)(2) to read as follows:
Sec. 73.1205 Written follow-up reports of physical security events.
* * * * *
(b) * * *
(2)(i) Licensees subject to Sec. 50.73 or Sec. 53.1640 of this
chapter must prepare the written follow-up report on NRC Form 366.
(ii) Licensees not subject to Sec. 50.73 or Sec. 53.1640 of this
chapter must prepare the written follow-up report in a letter format.
* * * * *
[[Page 87125]]
0
159. In Sec. 73.1210, revise paragraphs (a)(1) and (b)(3)(i) to read
as follows:
Sec. 73.1210 Recordkeeping of physical security events.
(a) * * *
(1) Licensees with facilities or shipment activities subject to the
provisions of Sec. 73.20, Sec. 73.25, Sec. 73.26, Sec. 73.27, Sec.
73.37, Sec. 73.45, Sec. 73.46, Sec. 73.50, Sec. 73.51, Sec. 73.55,
Sec. 73.60, Sec. 73.67, or Sec. 73.100, must record the physical
security events and conditions adverse to security that are specified
in paragraphs (c) through (f) of this section.
* * * * *
(b) * * *
(3)(i) Licensees must record these physical security events and
conditions adverse to security in either a stand-alone safeguards event
log or as part of the licensee's corrective action program, as
specified under the applicable quality assurance program provisions of
parts 50, 52, 53, 60, 63, 70, and 72 of this chapter, or both.
* * * * *
0
160. In Sec. 73.1215, revise paragraph (d)(1) introductory text to
read as follows:
Sec. 73.1215 Suspicious activity reports.
* * * * *
(d) * * *
(1) For licensees subject to the provisions of Sec. Sec. 73.20,
73.45, 73.46, 73.50, 73.51, 73.55, 73.60, 73.67, or 73.100, the
licensees must report activities they assess are suspicious. Examples
include, but are not limited to, the following:
* * * * *
0
161. In appendix B to part 73, revise Definitions introductory text to
read as follows:
Appendix B to Part 73--General Criteria for Security Personnel
* * * * *
Definitions
Terms defined in parts 50, 53, 70, and 73 of this chapter have
the same meaning when used in this appendix.
* * * * *
PART 74--MATERIAL CONTROL AND ACCOUNTING OF SPECIAL NUCLEAR
MATERIAL
0
162. The authority citation for 10 CFR part 74 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 53, 57, 161, 182,
223, 234, 1701 (42 U.S.C. 2073, 2077, 2201, 2232, 2273, 2282,
2297f); Energy Reorganization Act of 1974, secs. 201, 202 (42 U.S.C.
5841, 5842); 44 U.S.C. 3504 note.
0
163. In Sec. 74.31, revise paragraph (a) introductory text to read as
follows:
Sec. 74.31 Nuclear material control and accounting for special
nuclear material of low strategic significance.
(a) General performance objectives. Each licensee who is authorized
to possess and use more than one effective kilogram of special nuclear
material of low strategic significance, excluding sealed sources, at
any site or contiguous sites subject to control by the licensee, other
than a production or utilization facility licensed pursuant to parts
50, 53, or 70 of this chapter, or operations involved in waste
disposal, shall implement and maintain a Commission approved material
control and accounting system that will achieve the following
objectives:
* * * * *
0
164. In Sec. 74.41, revise paragraph (a) introductory text to read as
follows:
Sec. 74.41 Nuclear material control and accounting for special
nuclear material of moderate strategic significance.
(a) General performance objectives. Each licensee who is authorized
to possess special nuclear material (SNM) of moderate strategic
significance or SNM in a quantity exceeding one effective kilogram of
strategic special nuclear material in irradiated fuel reprocessing
operations other than as sealed sources and to use this material at any
site other than a nuclear reactor licensed pursuant to parts 50 or 53
of this chapter; or as reactor irradiated fuels involved in research,
development, and evaluation programs in facilities other than
irradiated fuel reprocessing plants; or an operation involved with
waste disposal, shall establish, implement, and maintain a Commission-
approved material control and accounting (MC&A) system that will
achieve the following performance objectives:
* * * * *
0
165. In Sec. 74.51, revise paragraph (a) introductory text to read as
follows:
Sec. 74.51 Nuclear material control and accounting for strategic
special nuclear material.
(a) General performance objectives. Each licensee who is authorized
to possess five or more formula kilograms of strategic special nuclear
material (SSNM) and to use such material at any site, other than a
nuclear reactor licensed pursuant to parts 50 or 53 of this chapter, an
irradiated fuel reprocessing plant, an operation involved with waste
disposal, or an independent spent fuel storage facility licensed
pursuant to part 72 of this chapter shall establish, implement, and
maintain a Commission-approved material control and accounting (MC&A)
system that will achieve the following objectives:
* * * * *
PART 75--SAFEGUARDS ON NUCLEAR MATERIAL--IMPLEMENTATION OF
SAFEGUARDS AGREEMENTS BETWEEN THE UNITED STATES AND THE
INTERNATIONAL ATOMIC ENERGY AGENCY
0
166. The authority citation for 10 CFR part 75 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 53, 63, 103, 104,
122, 161, 223, 234, 1701 (42 U.S.C. 2073, 2093, 2133, 2134, 2152,
2201, 2273, 2282, 2297f); Energy Reorganization Act of 1974, sec.
201 (42 U.S.C. 5841); Nuclear Waste Policy Act of 1982, secs. 135,
141 (42 U.S.C. 10155, 10161); 44 U.S.C. 3504 note.
0
167. In Sec. 75.4, revise the introductory text and paragraph (6) of
the definition for ``Facility'' to read as follows:
Sec. 75.4 Definitions.
* * * * *
Unless otherwise defined in this section, the terms defined in
Sec. Sec. 40.4, 50.2, 53.020, and 70.4 of this chapter have the same
meaning when used in this part.
* * * * *
Facility means:
* * * * *
(6) Any plant or location where the possession of more than 1
effective kilogram of nuclear material is licensed pursuant to 10 CFR
part 40, 50, 53, 60, 61, 63, 70, 72, 76, or 150 of this chapter or an
Agreement State license.
* * * * *
PART 95--FACILITY SECURITY CLEARANCE AND SAFEGUARDING OF NATIONAL
SECURITY INFORMATION AND RESTRICTED DATA
0
168. The authority citation for 10 CFR part 95 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 145, 161, 223, 234
(42 U.S.C. 2165, 2201, 2273, 2282); Energy Reorganization Act of
1974, sec. 201 (42 U.S.C. 5841); 44 U.S.C. 3504 note; E.O. 10865, as
amended, 25 FR 1583, 3 CFR, 1959-1963 Comp., p. 398; E.O. 12829, 58
FR 3479, 3 CFR, 1993 Comp., p. 570; E.O. 12968, 60 FR 40245, 3 CFR,
1995 Comp., p. 391; E.O. 13526, 75 FR 707, 3 CFR, 2009 Comp., p.
298.
0
169. In Sec. 95.5, revise the definition for ``License'' to read as
follows:
Sec. 95.5 Definitions.
* * * * *
License means a license issued under 10 CFR part 50, 52, 53, 54,
60, 63, 70, or 72.
* * * * *
[[Page 87126]]
Sec. 95.39 [Amended]
0
170. In Sec. 95.39(a), remove ``part 52'' and add in its place ``parts
52 or 53.''
PART 140--FINANCIAL PROTECTION REQUIREMENTS AND INDEMNITY
AGREEMENTS
0
171. The authority citation for 10 CFR part 140 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 161, 170, 223, 234
(42 U.S.C. 2201, 2210, 2273, 2282); Energy Reorganization Act of
1974, secs. 201, 202 (42 U.S.C. 5841, 5842); 44 U.S.C. 3504 note.
0
172. In Sec. 140.2, revise paragraphs (a)(1) and (2) to read as
follows:
Sec. 140.2 Scope.
(a) * * *
(1) To each person who is an applicant for or holder of a license
issued under 10 CFR part 50, 52, 53, or 54 to operate a nuclear
reactor, and
(2) With respect to an extraordinary nuclear occurrence, to each
person who is an applicant for or holder of a license to operate a
production facility or a utilization facility (including an operating
license issued under part 50 or part 53 of this chapter and a combined
license under part 52 or part 53 of this chapter), and to other persons
indemnified with respect to the involved facilities.
* * * * *
0
173. Revise Sec. 140.10 to read as follows:
Sec. 140.10 Scope.
This subpart applies to each person who is an applicant for or
holder of a license issued under 10 CFR parts 50, 53 or 54 to operate a
nuclear reactor, or is the applicant for or holder of a combined
license issued under 10 CFR parts 52, 53, or 54, except licenses held
by persons found by the Commission to be Federal agencies or nonprofit
educational institutions licensed to conduct educational activities.
This subpart also applies to persons licensed to possess and use
plutonium in a plutonium processing and fuel fabrication plant.
0
174. In Sec. 140.11, revise paragraph (b) to read as follows:
Sec. 140.11 Amounts of financial protection for certain reactors.
* * * * *
(b) In any case where a person is authorized under 10 CFR parts 50,
52, 53, or 54 to operate two or more nuclear reactors at the same
location, the total primary financial protection required of the
licensee for all such reactors is the highest amount which would
otherwise be required for any one of those reactors; provided, that
such primary financial protection covers all reactors at the location.
0
175. In Sec. 140.12, revise paragraph (c) to read as follows:
Sec. 140.12 Amount of financial protection required for other
reactors.
* * * * *
(c) In any case where a person is authorized under 10 CFR parts 50,
52, 53, or 54 to operate two or more nuclear reactors at the same
location, the total financial protection required of the licensee for
all such reactors is the highest amount which would otherwise be
required for any one of those reactors; provided, that such financial
protection covers all reactors at the location.
* * * * *
0
176. Revise Sec. 140.13 to read as follows:
Sec. 140.13 Amount of financial protection required of certain
holders of construction permits and combined licenses under 10 CFR part
52.
Each holder of a 10 CFR part 50 or 10 CFR part 53 construction
permit, or a holder of a combined license under parts 52 or 53 of this
chapter before the date that the Commission had made the finding under
Sec. Sec. 52.103(g) or 53.1452(g) of this chapter, who also holds a
license under part 70 of this chapter authorizing ownership, possession
and storage only of special nuclear material at the site of the nuclear
reactor for use as fuel in operation of the nuclear reactor after
issuance of either an operating license under 10 CFR part 50 or 53, or
a combined license under 10 CFR part 52 or 53, shall, during the period
before issuance of a license authorizing operation under 10 CFR part 50
or 53, or the period before the Commission makes the finding under
Sec. 52.103(g) or Sec. 53.1452(g) of this chapter, as applicable,
have and maintain financial protection in the amount of $1,000,000.
Proof of financial protection shall be filed with the Commission in the
manner specified in Sec. 140.15 before issuance of the license under
part 70 of this chapter.
0
177. In Sec. 140.20, revise paragraphs (a)(1)(i) and (ii) to read as
follows:
Sec. 140.20 Indemnity agreements and liens.
(a) * * *
(1)(i) The effective date of the license (issued under part 50 or
part 53 of this chapter) authorizing the licensee to operate the
nuclear reactor involved; or
(ii) The date that the Commission makes the finding under
Sec. Sec. 52.103(g) or 53.1452(g) of this chapter; or
* * * * *
PART 150--EXEMPTIONS AND CONTINUED REGULATORY AUTHORITY IN
AGREEMENT STATES AND IN OFFSHORE WATERS UNDER SECTION 274
0
178. The authority citation for 10 CFR part 150 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 11, 53, 81, 83, 84,
122, 161, 181, 223, 234, 274 (42 U.S.C. 2014, 2201, 2231, 2273,
2282, 2021); Energy Reorganization Act of 1974, sec. 201 (42 U.S.C.
5841); Nuclear Waste Policy Act of 1982, secs. 135, 141 (42 U.S.C.
10155, 10161); 44 U.S.C. 3504 note.
0
179. In Sec. 150.15, revise paragraphs (a)(7)(iii) and (a)(8) to read
as follows:
Sec. 150.15 Persons not exempt.
(a) * * *
(7) * * *
(iii) Greater than Class C (GTCC) waste, as defined in part 72 of
this chapter, in an ISFSI or an MRS licensed under part 72 of this
chapter; the GTCC waste must originate in, or be used by, a facility
licensed under parts 50, 52, or 53 of this chapter.
(8) Greater than Class C waste, as defined in part 72 of this
chapter, that originates in, or is used by, a facility licensed under
parts 50, 52, or 53 of this chapter and is licensed under part 30 and/
or part 70 of this chapter.
* * * * *
PART 170--FEES FOR FACILITIES, MATERIALS, IMPORT AND EXPORT
LICENSES, AND OTHER REGULATORY SERVICES UNDER THE ATOMIC ENERGY ACT
OF 1954, AS AMENDED
0
180. The authority citation for 10 CFR part 170 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 11, 161(w) (42
U.S.C. 2014, 2201(w)); Energy Reorganization Act of 1974, sec. 201
(42 U.S.C. 5841); 42 U.S.C. 2215; 31 U.S.C. 901, 902, 9701; 44
U.S.C. 3504 note.
0
181. In Sec. 170.3, revise the definitions for ``Manufacturing
License,'' ``Part 55 Reviews,'' ``Power reactor,'' and ``Special
projects'' to read as follows:
Sec. 170.3 Definitions.
* * * * *
Manufacturing license means a license under subpart F of part 52 of
this chapter or subpart H of part 53 of this chapter to manufacture a
nuclear power reactor(s) to be operated at sites not identified in the
license application.
* * * * *
Part 55 Reviews as used in this part means those services provided
by the Commission to administer requalification and replacement
examinations and tests for reactor
[[Page 87127]]
operators licensed under 10 CFR part 55 or 53 of the Commission's
regulations and employed by part 50 or 53 licensees. These services
also include related items such as the preparation, review, and grading
of the examinations and tests.
* * * * *
Power reactor means a nuclear reactor designed to produce
electrical or heat energy licensed by the Commission under the
authority of section 103 or subsection 104b of the Act, and under the
provisions of Sec. Sec. 50.21(b), 50.22, or part 53 of this chapter.
* * * * *
Special projects means specific services provided by the Commission
for which fees are not otherwise specified in this chapter. This
includes, but is not limited to, contested hearings on licensing
actions directly related to U.S. Government national security
initiatives (as determined by the NRC), topical report reviews, early
site reviews, waste solidification activities, activities related to
the tracking and monitoring of shipment of classified matter, services
provided to certify licensee, vendor, or other private industry
personnel as instructors for 10 CFR part 55 or 53 reactor operators,
reviews of financial assurance submittals that do not require a license
amendment, reviews of responses to Confirmatory Action Letters, reviews
of uranium recovery licensees' land-use survey reports, and reviews of
Sec. Sec. 50.71 or 53.1545 of this chapter Final Safety Analysis
Reports. Special projects does not include activities otherwise exempt
from fees under this part. It also does not include those contested
hearings for which a fee exemption is granted in Sec. 170.11(a)(2),
including those related to individual plant security modifications.
* * * * *
0
182. In Sec. 170.12, revise paragraph (d)(1)(v) to read as follows:
Sec. 170.12 Payment of fees.
* * * * *
(d) * * *
(1) * * *
(v) 10 CFR 50.71 or 53.1545 final safety analysis reports;
* * * * *
Sec. 170.21 [Amended]
0
183. In Sec. 170.21, in footnote 1 remove the phrase ``(e.g., 10 CFR
50.12, 10 CFR 73.5)'' and add in its place the phrase ``(e.g., 10 CFR
50.12, 10 CFR 53.080, 10 CFR 73.5)''.
0
184. Revise Sec. 170.41to read as follows:
Sec. 170.41 Failure by an applicant or licensee to pay prescribed
fees.
If the Commission determines that an applicant or a licensee has
failed to pay a prescribed fee required in this part, the Commission
will not process any application and may suspend or revoke any license
or approval issued to the applicant or licensee. The Commission may
issue an order with respect to licensed activities that the Commission
determines to be appropriate or necessary to carry out the provisions
of this part, parts 30, 31, 32 through 35, 40, 50, 53, 61, 70, 71, 72,
73, and 76 of this chapter, and of the Act.
PART 171--ANNUAL FEES FOR REACTOR LICENSES AND FUEL CYCLE LICENSES
AND MATERIALS LICENSES, INCLUDING HOLDERS OF CERTIFICATES OF
COMPLIANCE, REGISTRATIONS, AND QUALITY ASSURANCE PROGRAM APPROVALS
AND GOVERNMENT AGENCIES LICENSED BY THE NRC
0
185. The authority citation for 10 CFR part 171 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 11, 161(w), 223, 234
(42 U.S.C. 2014, 2201(w), 2273, 2282); Energy Reorganization Act of
1974, sec. 201 (42 U.S.C. 5841); 42 U.S.C. 2215; 44 U.S.C. 3504
note.
0
186. Revise Sec. 171.3 to read as follows:
Sec. 171.3 Scope.
The regulations in this part apply to any person holding an
operating license for a test reactor or research reactor issued under
part 50 of this chapter, and to any person holding an operating license
for a power reactor licensed under 10 CFR part 50 or 53, or a combined
license issued under 10 CFR part 52 or 53, that has provided
notification to the U.S. Nuclear Regulatory Commission (NRC) that the
licensee has successfully completed power ascension testing. The
regulations in this part also apply to any person holding a materials
license as defined in this part, a Certificate of Compliance, a sealed
source or device registration, a quality assurance program approval,
and to a Government agency as defined in this part. Notwithstanding the
other provisions in this section, the regulations in this part do not
apply to uranium recovery and fuel facility licensees until after the
Commission verifies through inspection that the facility has been
constructed in accordance with the requirements of the license.
0
187. In Sec. 171.5, revise the definitions for ``Operating license,''
and ``Power reactor'' to read as follows:
Sec. 171.5 Definitions.
* * * * *
Operating license means having a license issued under Sec. Sec.
50.57 or 53.1387 of this chapter. It does not include licenses that
only authorize possession of special nuclear material after the
Commission has received a request from the licensee to amend its
licensee to permanently withdraw its authority to operate or the
Commission has permanently revoked such authority.
* * * * *
Power reactor means a nuclear reactor designed to produce
electrical or heat energy and licensed by the Commission under the
authority of section 103 or subsection 104b of the Atomic Energy Act of
1954, as amended, and under the provisions of Sec. Sec. 50.21(b) or
50.22, or part 53 of this chapter.
* * * * *
0
188. In Sec. 171.15, revise paragraphs (a), (b)(2)(iii), (c)(1), and
(d)(1) to read as follows:
Sec. 171.15 Annual fees: Non-power production or utilization
licenses, reactor licenses, and independent spent fuel storage
licenses.
(a) Each person holding an operating license for one or more non-
power production or utilization facilities under 10 CFR part 50 that
has provided notification to the NRC of the successful completion of
startup testing; each person holding an operating license for a power
reactor licensed under 10 CFR part 50 or a combined license under 10
CFR part 52, or an operating license or combined license for a
commercial nuclear plant under 10 CFR part 53, that has provided
notification to the NRC of the successful completion of power ascension
testing; each person holding a 10 CFR part 50 or 52, power reactor
license, or a 10 CFR part 53 commercial nuclear plant license that is
in decommissioning or possession only status, except those that have no
spent fuel onsite; and each person holding a 10 CFR part 72 license who
does not hold a 10 CFR part 50, 52, or 53 license and provides
notification under Sec. 72.80(g) of this chapter, shall pay the annual
fee for each license held during the Federal fiscal year in which the
fee is due. This paragraph (a) does not apply to test or research
reactors exempted under Sec. 171.11(b).
(b) * * *
(2) * * *
(iii) Generic activities required largely for NRC to regulate power
reactors (e.g., updating part 50, part 52, or part 53 of this chapter,
operating the Incident Response Center, new reactor regulatory
infrastructure). The base annual fee for operating power reactors does
not
[[Page 87128]]
include generic activities specifically related to reactor
decommissioning.
(c)(1) The FY 2022 annual fee for each power reactor holding a 10
CFR part 50 operating license or combined license issued under 10 CFR
part 52 or part 53 that is in a decommissioning or possession-only
status and has spent fuel onsite, and for each independent spent fuel
storage 10 CFR part 72 licensee who does not hold a 10 CFR part 50 or
part 53 operating license, or a 10 CFR part 52 or part 53 combined
license, is $227,000.
* * * * *
(d)(1) Each person holding an operating license for an SMR issued
under 10 CFR part 50 or part 53, or a combined license issued under 10
CFR part 52 or part 53, that has provided notification to the NRC of
the successful completion startup testing, shall pay the annual fee for
all licenses held for an SMR site. The annual fee will be determined
using the cumulative licensed thermal power rating of all SMR units and
the bundled unit concept, during the fiscal year in which the fee is
due. For a given site, the use of the bundled unit concept is
independent of the number of SMR plants, the number of SMR licenses
issued, or the sequencing of the SMR licenses that have been issued.
* * * * *
0
189. In Sec. 171.17, revise paragraphs (a) introductory text,
(a)(1)(ii), and (a)(2) to read as follows:
Sec. 171.17 Proration.
(a) Reactors, 10 CFR part 72 licensees who do not hold 10 CFR part
50, 52, or 53 licenses, and materials licenses with annual fees of
$100,000 or greater for a single fee category. The NRC will base the
proration of annual fees for terminated and downgraded licenses on the
fee rule in effect at the time the action is official. The NRC will
base the determinations on the proration requirements under paragraphs
(a)(2) and (3) of this section.
(1) * * *
(ii) The annual fees for new licenses for non-power production or
utilization facilities, 10 CFR part 72 licensees who do not hold 10 CFR
part 50, 52, or 53 licenses, and materials licenses with annual fees of
$100,000 or greater for a single fee category for the current FY, that
are subject to fees under this part and are granted a license to
operate on or after October 1 of a FY, are prorated on the basis of the
number of days remaining in the FY. Thereafter, the full annual fee is
due and payable each subsequent FY.
(2) Terminations. The base operating power reactor annual fee for
operating reactor licensees or the annual fee for small modular reactor
licensees, who have requested amendment to withdraw operating authority
permanently during the FY will be prorated based on the number of days
during the FY the license was in effect before docketing of the
certifications for permanent cessation of operations and permanent
removal of fuel from the reactor vessel or when a final legally
effective order to permanently cease operations has come into effect.
The spent fuel storage/reactor decommissioning annual fee for reactor
licensees who permanently cease operations and have permanently removed
fuel from the site during the FY will be prorated on the basis of the
number of days remaining in the FY after docketing of both the
certifications of permanent cessation of operations and permanent
removal of fuel from the site. The spent fuel storage/reactor
decommissioning annual fee will be prorated for those 10 CFR part 72
licensees who do not hold a 10 CFR part 50, 52, or 53 license who
request termination of the 10 CFR part 72 license and permanently cease
activities authorized by the license during the FY based on the number
of days the license was in effect before receipt of the termination
request. The annual fee for materials licenses with annual fees of
$100,000 or greater for a single fee category for the current FY will
be prorated based on the number of days remaining in the FY when a
termination request or a request for a possession-only license is
received by the NRC, provided the licensee permanently ceased licensed
activities during the specified period. The annual fee for non-power
production or utilization facilities will be prorated based on the
number of days remaining in the FY when the authorization to operate
the facility has been permanently removed from the license during the
FY.
* * * * *
Dated: October 7, 2024.
For the Nuclear Regulatory Commission.
Carrie Safford,
Secretary of the Commission.
[FR Doc. 2024-23434 Filed 10-23-24; 8:45 am]
BILLING CODE 7590-01-P
