NuScale Power, LLC, Design-Specific Review Standard and Scope and Safety Review Matrix, 75449-75452 [2016-26210]
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Federal Register / Vol. 81, No. 210 / Monday, October 31, 2016 / Notices
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[FR Doc. 2016–26169 Filed 10–28–16; 8:45 am]
BILLING CODE 7535–01–P
NUCLEAR REGULATORY
COMMISSION
[NRC–2015–0160]
NuScale Power, LLC, Design-Specific
Review Standard and Scope and
Safety Review Matrix
Nuclear Regulatory
Commission.
AGENCY:
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75449
NuScale design-specific review
standard; issuance.
ACTION:
The U.S. Nuclear Regulatory
Commission (NRC or Commission) has
issued the NuScale Power, LLC,
(NuScale), Design-Specific Review
Standard (DSRS) Sections, and is
issuing the final NuScale DSRS Scope
and Safety Review Matrix, for NuScale
Design Certification (DC), Combined
License (COL), and Early Site Permit
(ESP) reviews. The NRC staff is also
issuing the DSRS public comment
resolution matrices, which address the
comments received on the draft DSRS.
The NuScale DSRS provides guidance to
the NRC staff for performing safety
reviews for those specific areas where
existing NUREG–0800, ‘‘Standard
Review Plan [SRP] for the Review of
Safety Analysis Reports for Nuclear
Power Plants: LWR Edition,’’ sections
do not address the unique features of
the NuScale design.
DATES: The DSRS sections were effective
upon issuance between June 24 and
August 4, 2016.
ADDRESSES: Please refer to Docket ID
NRC–2015–0160 when contacting the
NRC about the availability of
information regarding this document.
You may obtain publically-available
information related to this document
using any of the following methods:
• Federal Rulemaking Web site: Go to
https://www.regulations.gov and search
for Docket ID NRC–2015–0160. Address
questions about NRC dockets to Carol
Gallagher; telephone: 301–415–3463;
email: Carol.Gallagher@nrc.gov. For
technical questions, contact the
individual listed in the FOR FURTHER
INFORMATION CONTACT section of this
document.
• NRC’s Agencywide Documents
Access and Management System
(ADAMS): You may obtain publiclyavailable documents online in the
ADAMS Public Documents collection at
https://www.nrc.gov/reading-rm/
adams.html. To begin the search, select
‘‘ADAMS Public Documents’’ and then
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, 301–415–4737, or by
email to pdr.resource@nrc.gov. The
ADAMS accession number for each
document referenced (if it is available in
ADAMS) is provided the first time that
it is mentioned in this document. The
DSRS is available in ADAMS Package
Accession No. ML15355A295 and the
final NuScale DSRS Scope and Safety
Review Matrix is also available in
ADAMS under Accession No.
ML16263A000. The resolution of
SUMMARY:
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comments on the draft DSRS is
documented in the DSRS Public
Comment Resolution Matrices (ADAMS
Package Accession No. ML16083A615).
In addition, for the convenience of the
reader, the ADAMS accession numbers
are provided in a table in the
‘‘Availability of Documents’’ section of
this document.
• NRC’s PDR: You may examine and
purchase copies of public documents at
the NRC’s PDR, Room O1–F21, One
White Flint North, 11555 Rockville
Pike, Rockville, Maryland 20852.
FOR FURTHER INFORMATION CONTACT:
Rajender Auluck, telephone: 301–415–
1025; email: Rajender.Auluck@nrc.gov
or Gregory Cranston, telephone: 301–
415–0546; email: Gregory.Cranston@
nrc.gov; both are staff members of the
Office of New Reactors, U.S. Nuclear
Regulatory Commission, Washington,
DC 20555–0001.
SUPPLEMENTARY INFORMATION:
I. Background
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In the Staff Requirements
Memorandum (SRM) COMGBJ–10–
0004/COMGEA–10–0001, ‘‘Use of Risk
Insights to Enhance Safety Focus of
Small Modular Reactor Reviews,’’ dated
August 31, 2010 (ADAMS Accession
No. ML102510405), the Commission
provided direction to the NRC staff on
the preparation for, and review of, small
modular reactor (SMR) applications,
with a near-term focus on integral
pressurized-water reactor designs. The
Commission directed the NRC staff to
more fully integrate the use of risk
insights into pre-application activities
and the review of applications and,
consistent with regulatory requirements
and Commission policy statements, to
align the review focus and resources to
risk-significant structures, systems, and
components and other aspects of the
design that contribute most to safety in
order to enhance the effectiveness and
efficiency of the review process. The
Commission directed the NRC staff to
develop a design-specific, risk-informed
review plan for each SMR design to
address pre-application and application
review activities. An important part of
this review plan is the DSRS. The DSRS
for the NuScale design is the result of
the implementation of the Commission’s
direction.
II. DSRS for the NuScale Design
The NuScale DSRS (available in
ADAMS Package Accession No.
ML15355A295) reflects current NRC
staff safety review methods and
practices which integrate risk insights
and, where appropriate, lessons learned
from the NRC’s reviews of DC and COL
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applications completed since the last
revision of the NUREG–0800, SRP
Introduction, Part 2, ‘‘Standard Review
Plan for the Review of Safety Analysis
Reports for Nuclear Power Plants: LightWater Small Modular Reactor Edition,’’
January 2014 (ADAMS Accession No.
ML13207A315). The NuScale DSRS
Scope and Safety Review Matrix
provides a complete list of SRP sections
and identifies which SRP sections will
be used for DC, COL, or ESP reviews
concerning the NuScale design; which
SRP sections are not applicable to the
NuScale design; which SRP sections
needed modification and were reissued
as DSRS sections; and which new DSRS
sections were added to address a unique
design consideration in the NuScale
design. The final NuScale DSRS Scope
and Safety Review Matrix is available in
ADAMS under Accession No.
ML16263A000.
The NRC staff developed the content
of the NuScale DSRS as an alternative
method for evaluating a NuScalespecific application and has determined
that the application may address the
DSRS in lieu of addressing the SRP,
with specified exceptions. These
exceptions include particular review
areas in which the DSRS directs
reviewers to consult the SRP and others
in which the SRP is used for the review
as identified in the final NuScale DSRS
Scope and Safety Review Matrix. If
NuScale chooses to address the DSRS,
the application should identify and
describe all differences between the
design features (DC and COL
applications only), analytical
techniques, and procedural measures
proposed in an application and the
guidance of the applicable DSRS section
(or SRP section, as specified in the
NuScale DSRS Scope and Safety Review
Matrix), and discuss how the proposed
alternative provides an acceptable
method of complying with the
regulations that underlie the DSRS
acceptance criteria. The staff has
accepted the content of the DSRS as an
alternative method for evaluating
whether an application complies with
NRC regulations for the NuScale Small
Modular Reactor applications, provided
that the application does not deviate
significantly from the design and siting
assumptions made by the NRC staff
while preparing the DSRS. If the design
or siting assumptions in a NuScale
application deviate significantly from
the design and siting assumptions the
staff used in preparing the DSRS, the
staff will use the more general guidance
in the SRP, as specified in sections
52.17(a)(1)(xii), 52.47(a)(9), or
52.79(a)(41) of title 10 of the Code of
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Federal Regulations, depending on the
type of application. Alternatively, the
staff may supplement the DSRS section
by adding appropriate criteria to address
new design or siting assumptions.
The NRC staff issued a Federal
Register notice on June 30, 2015 (80 FR
37312), to request public comments on
the draft NuScale DSRS Scope and
Safety review Matrix (ADAMS
Accession No. ML15156B063) and the
individual NuScale-specific DSRS
sections referenced in the table included
in the FRN. A correction Federal
Register notice was published on July 9,
2015 (80 FR 39454), to identify an
additional draft DSRS section for which
comments were requested. In response,
the NRC received comments from:
NuScale Power, LLC, by letter dated
August 31, 2015 (ADAMS Accession
No. ML15258A081), the Nuclear Energy
Institute (NEI) by letter dated August 31,
2015 (ADAMS Accession No.
ML15257A012), Mark Thomson by
electronic submission dated August 31,
2015 (ADAMS Accession No.
ML15292A309), an anonymous
submitter by electronic submission
dated August 31, 2015 (ADAMS
Accession No. ML15292A310), an
anonymous submitter by electronic
submission dated August 31, 2015
(ADAMS Accession No. ML15292A311),
Clinton Ferrara by electronic
submission dated August 31, 2015
(ADAMS Accession No. ML15292A333),
and Paula Ferrara by electronic
submission dated August 31, 2015,
(ADAMS Accession No. ML15292A334).
Several of these comments have been
previously discussed during public
meetings held in support of developing
the draft DSRS sections. These
comments and resolutions have been
documented in the DSRS Public
Comment Resolution Matrices and are
publicly available (ADAMS Package
Accession No. ML16083A615).
In the June 30, 2015 Federal Register
notice, the NRC requested public
comments on 115 DSRS sections. The
NRC staff determined whether to
develop a DSRS section after
considering whether significant
differences in the functions,
characteristics, or attributes of the
NuScale design required major revision
of the related SRP section guidance, or
whether structures, systems, and
components identified in the NuScale
design are unique and not addressed by
the current SRP. Following publication
of the draft version of the DSRS
sections, the NRC staff revisited these
criteria and determined, based on the
most recent NuScale design, that it is
appropriate to use the related SRP
section in lieu of a draft DSRS section
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in a number of cases. In these cases the
draft DSRS sections have not been
issued as final, and the related SRP
sections will be used for the NuScale
review. In deciding to use the related
SRP sections, the staff has not
necessarily determined that the SRP
sections are wholly applicable without
modification. For example, as the NRC
staff gains greater understanding of the
NuScale design or if the design changes
during the review, the staff would assess
whether different or supplemental
review criteria are needed. Stakeholders
who believe that different or
supplemental review criteria are needed
may provide these views to the NRC
staff for consideration during the
application review period.
The results of determinations to use
the related SRP sections rather than
draft DSRS sections, along with other
identified issues with the draft NuScale
DSRS Scope and Safety Review Matrix,
are documented in a separate
‘‘transitional’’ NuScale DSRS Scope and
Safety Review Matrix (ADAMS
Accession No. ML16076A048). The
‘‘transitional’’ Matrix shows the
differences between the draft and final
NuScale DSRS Scope and Safety Review
Matrices and describes the reasons for
these differences. The resulting final list
of DSRS titles with corresponding
section numbers and ADAMS references
are provided in the table below and in
ADAMS Package Accession No.
ML15355A295.
In the future, should additional SRP
sections be developed, the staff will
determine at that time their
applicability to the NuScale design. In
addition, the NRC disseminates
information regarding current safety
issues and proposed solutions through
various means, such as generic
communications and the process for
treating generic safety issues. When
current issues are resolved, the staff will
determine the need, extent, and nature
of revision that should be made to the
SRP and/or DSRS to reflect the new
NRC guidance.
III. Availability of Documents
ADAMS
accession No.
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Section
Design-specific review standard title
Matrix ................
Matrix ................
3.5.1.3 ...............
3.7.1 ..................
3.7.2 ..................
3.7.3 ..................
3.8.2 ..................
3.8.4 ..................
3.8.5 ..................
3.11 ...................
4.4 .....................
5.2.4 ..................
5.2.5 ..................
5.3.1 ..................
5.3.2 ..................
5.3.3 ..................
5.4.2.1 ...............
5.4.2.2 ...............
5.4.7 ..................
BTP 5–4 ............
6.2.1 ..................
6.2.1.1.A ...........
6.2.1.3 ...............
6.2.1.4 ...............
6.2.2 ..................
6.2.4 ..................
6.2.5 ..................
6.2.6 ..................
6.3 .....................
6.6 .....................
7.0 .....................
7.1 .....................
7.2 .....................
7.0, App A .........
7.0, App B .........
7.0, App C ........
7.0, App D ........
8.1 .....................
8.2 .....................
8.3.1 ..................
8.3.2 ..................
8.4 .....................
9.1.2 ..................
9.1.3 ..................
9.3.4 ..................
9.3.6 ..................
9.5.2 ..................
10.2.3 ................
10.3 ...................
10.4.7 ................
11.1 ...................
NuScale DSRS Scope and Safety Review Matrix (Transitional) ...................................................................
NuScale DSRS Scope and Safety Review Matrix (Final) .............................................................................
Turbine Missiles .............................................................................................................................................
Seismic Design Parameters ...........................................................................................................................
Seismic System Analysis ...............................................................................................................................
Seismic Subsystem Analysis .........................................................................................................................
Steel Containment ..........................................................................................................................................
Other Seismic Category I Structures .............................................................................................................
Foundations ....................................................................................................................................................
Environmental Qualification of Mechanical and Electrical Equipment ...........................................................
Thermal and Hydraulic Design .......................................................................................................................
Reactor Coolant Pressure Boundary Inservice Inspection and Testing ........................................................
Reactor Coolant Pressure Boundary Leakage Detection ..............................................................................
Reactor Vessel Materials ...............................................................................................................................
Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock .................................
Reactor Vessel Integrity .................................................................................................................................
Steam Generator Materials ............................................................................................................................
Steam Generator Program .............................................................................................................................
Decay Heat Removal (DHR) System .............................................................................................................
Design Requirements of the Decay Heat Removal System ..........................................................................
Containment Functional Design .....................................................................................................................
Containments .................................................................................................................................................
Mass and Energy Release Analysis for Postulated Loss-of-Coolant Accidents (LOCAs) ............................
Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures ...............................
Containment Heat Removal Systems ............................................................................................................
Containment Isolation System .......................................................................................................................
Combustible Gas Control in Containment .....................................................................................................
Containment Leakage Testing .......................................................................................................................
Emergency Core Cooling System ..................................................................................................................
Inservice Inspection and Testing of Class 2 and 3 Components ..................................................................
Instrumentation and Controls—Introduction and Overview of Review Process ............................................
Instrumentation and Controls—Fundamental Design Principles ...................................................................
Instrumentation and Controls—System Characteristics ................................................................................
Instrumentation and Controls—Hazard Analysis ...........................................................................................
Instrumentation and Controls—System Architecture .....................................................................................
Instrumentation and Controls—Simplicity ......................................................................................................
Instrumentation and Controls—References ...................................................................................................
Electric Power—Introduction ..........................................................................................................................
Offsite Power System .....................................................................................................................................
AC Power Systems (Onsite) ..........................................................................................................................
DC Power Systems (Onsite) ..........................................................................................................................
Station Blackout .............................................................................................................................................
New and Spent Fuel Storage .........................................................................................................................
Spent Fuel Pool Cooling and Cleanup System .............................................................................................
Chemical and Volume Control System ..........................................................................................................
Containment Evacuation and Flooding Systems ...........................................................................................
Communications Systems ..............................................................................................................................
Turbine Rotor Integrity ...................................................................................................................................
Main Steam Supply System ...........................................................................................................................
Condensate and Feedwater System ..............................................................................................................
Source Terms .................................................................................................................................................
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Section
11.2
11.3
11.4
11.5
11.6
...................
...................
...................
...................
...................
12.2 ...................
12.3–12.4 ..........
12.5 ...................
14.2 ...................
14.3.5 ................
15.0 ...................
15.0.3 ................
15.1.1—15.1.4 ..
15.1.5 ................
15.1.6 ................
15.2.1–15.2.5 ....
15.2.6 ................
15.2.7 ................
15.2.8 ................
15.5.1–15.5.2 ....
15.6.5 ................
15.6.6 ................
15.9A ................
16.0 ...................
Liquid Waste Management System ...............................................................................................................
Gaseous Waste Management System ..........................................................................................................
Solid Waste Management System .................................................................................................................
Process and Effluent Radiological Monitoring Instrumentation and Sampling Systems ...............................
Guidance on Instrumentation and Control Design Features for Process and Effluent Radiological Monitoring, and Area Radiation and Airborne Radioactivity Monitoring.
Radiation Sources ..........................................................................................................................................
Radiation Protection Design Features ...........................................................................................................
Operational Radiation Protection Program ....................................................................................................
Initial Plant Test Program—Design Certification and New License Applicants .............................................
Instrumentation and Controls—Inspections, Tests, Analyses, and Acceptance Criteria ..............................
Introduction—Transient and Accident Analyses ............................................................................................
Design Basis Accidents Radiological Consequence Analyses for NuScale SMR Design ............................
Decrease in FW Temperature, Increase in FW Flow, Increase in Steam Flow and Inadvertent Opening of
the Turbine Bypass System or Inadvertent Operation of the Decay Heat Removal System.
Steam System Piping Failures Inside and Outside of Containment .............................................................
Loss of Containment Vacuum ........................................................................................................................
Loss of External Load; Turbine Trip; Loss of Condenser Vacuum; Closure of Main Steam Isolation Valve
(BWR); and Steam Pressure Regulator Failure (Closed).
Loss of Non-Emergency AC Power to the Station Auxiliaries .......................................................................
Loss of Normal Feedwater Flow ....................................................................................................................
Feedwater System Pipe Breaks Inside and Outside Containment ...............................................................
Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory ...................
LOCAs Resulting From Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure
Boundary.
Inadvertent Opening of the Emergency Core Cooling System .....................................................................
Thermal-hydraulic Stability .............................................................................................................................
Technical Specifications .................................................................................................................................
Dated at Rockville, Maryland, this 21st day
of October 2016.
For the Nuclear Regulatory Commission.
Frank Akstulewicz,
Director, Division of New Reactor Licensing,
Office of New Reactors.
[FR Doc. 2016–26210 Filed 10–28–16; 8:45 am]
BILLING CODE 7590–01–P
NUCLEAR REGULATORY
COMMISSION
[Docket Nos.: 52–029 and 52–030; NRC–
2008–0558]
Duke Energy Florida, LLC; Levy
Nuclear Plant Units 1 and 2
Nuclear Regulatory
Commission
ACTION: Notice of intent to enter into a
modified indemnity agreement.
AGENCY:
The U.S. Nuclear Regulatory
Commission (NRC) is issuing a notice of
intent to enter into a modified
indemnity agreement with Duke Energy
Florida, LLC, (DEF) to operate Levy
Nuclear Plant Units 1 and 2 (LNP 1 and
2). The NRC is required to publish
notice of its intent to enter into an
indemnity agreement which contains
provisions different from the general
form found in the NRC’s regulations. A
modification to the general form is
necessary to accommodate the unique
timing provisions of a combined license
(COL).
sradovich on DSK3GMQ082PROD with NOTICES
SUMMARY:
VerDate Sep<11>2014
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accession No.
Design-specific review standard title
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On October 20, 2016, the
Commission authorized the Director of
the Office of New Reactors to issue
COLs to DEF to construct and operate
LNP 1 and 2. The modified indemnity
agreement would be effective upon
issuance of the COLs.
ADDRESSES: Please refer to Docket ID
NRC–2008–0558 when contacting the
NRC about the availability of
information regarding this document.
You may obtain publicly-available
information related to this document
using any of the following methods:
• Federal Rulemaking Web site: Go to
https://www.regulations.gov and search
for Docket ID NRC–2008–0558. Address
questions about NRC dockets to Carol
Gallagher; telephone: 301–415–3463;
email: Carol.Gallagher@nrc.gov. For
technical questions, contact the
individual listed in the FOR FURTHER
INFORMATION CONTACT section of this
document.
• NRC’s Agencywide Documents
Access and Management System
(ADAMS): You may obtain publiclyavailable documents online in the
ADAMS Public Documents collection at
https://www.nrc.gov/reading-rm/
adams.html. To begin the search, select
‘‘ADAMS Public Documents’’ and then
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, 301–415–4737, or by
email to pdr.resource@nrc.gov.
DATES:
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ML15363A397
ML15355A309
ML15355A310
ML15355A311
ML15355A312
• NRC’s PDR: You may examine and
purchase copies of public documents at
the NRC’s PDR, Room O1–F21, One
White Flint North, 11555 Rockville
Pike, Rockville, Maryland 20852.
FOR FURTHER INFORMATION CONTACT:
Donald Habib, Office of New Reactors,
U.S. Nuclear Regulatory Commission,
Washington DC 20555–0001; telephone:
301–415–1035, email: Donald.Habib@
nrc.gov.
SUPPLEMENTARY INFORMATION:
I. Background
On October 20, 2016, the Commission
authorized issuance of COLs to DEF for
LNP 1 and 2. These COLs would
include a license pursuant to part 70 of
title 10 of the Code of Federal
Regulations (10 CFR), ‘‘Domestic
Licensing of Special Nuclear Material.’’
Pursuant to 10 CFR 140.20(a)(1)(iii), the
NRC will execute and issue agreements
of indemnity effective on the date of a
license under 10 CFR part 70
authorizing the licensee to possess and
store special nuclear material at the site
of the nuclear reactor for use as fuel in
operation of the nuclear reactor after
issuance of an operating license for the
reactor. The general form of indemnity
agreement to be entered into by the NRC
with DEF is contained in 10 CFR 140.92,
‘‘Appendix B—Form of Indemnity
Agreement with licensees furnishing
insurance policies as proof of financial
protection.’’
E:\FR\FM\31OCN1.SGM
31OCN1
]]>Agencies
[Federal Register Volume 81, Number 210 (Monday, October 31, 2016)]
[Notices]
[Pages 75449-75452]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-26210]
=======================================================================
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
[NRC-2015-0160]
NuScale Power, LLC, Design-Specific Review Standard and Scope and
Safety Review Matrix
AGENCY: Nuclear Regulatory Commission.
ACTION: NuScale design-specific review standard; issuance.
-----------------------------------------------------------------------
SUMMARY: The U.S. Nuclear Regulatory Commission (NRC or Commission) has
issued the NuScale Power, LLC, (NuScale), Design-Specific Review
Standard (DSRS) Sections, and is issuing the final NuScale DSRS Scope
and Safety Review Matrix, for NuScale Design Certification (DC),
Combined License (COL), and Early Site Permit (ESP) reviews. The NRC
staff is also issuing the DSRS public comment resolution matrices,
which address the comments received on the draft DSRS. The NuScale DSRS
provides guidance to the NRC staff for performing safety reviews for
those specific areas where existing NUREG-0800, ``Standard Review Plan
[SRP] for the Review of Safety Analysis Reports for Nuclear Power
Plants: LWR Edition,'' sections do not address the unique features of
the NuScale design.
DATES: The DSRS sections were effective upon issuance between June 24
and August 4, 2016.
ADDRESSES: Please refer to Docket ID NRC-2015-0160 when contacting the
NRC about the availability of information regarding this document. You
may obtain publically-available information related to this document
using any of the following methods:
Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2015-0160. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
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 ``ADAMS Public Documents'' and
then 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, 301-415-4737, or by email to pdr.resource@nrc.gov. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in this
document. The DSRS is available in ADAMS Package Accession No.
ML15355A295 and the final NuScale DSRS Scope and Safety Review Matrix
is also available in ADAMS under Accession No. ML16263A000. The
resolution of
[[Page 75450]]
comments on the draft DSRS is documented in the DSRS Public Comment
Resolution Matrices (ADAMS Package Accession No. ML16083A615). In
addition, for the convenience of the reader, the ADAMS accession
numbers are provided in a table in the ``Availability of Documents''
section of this document.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
FOR FURTHER INFORMATION CONTACT: Rajender Auluck, telephone: 301-415-
1025; email: Rajender.Auluck@nrc.gov or Gregory Cranston, telephone:
301-415-0546; email: Gregory.Cranston@nrc.gov; both are staff members
of the Office of New Reactors, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001.
SUPPLEMENTARY INFORMATION:
I. Background
In the Staff Requirements Memorandum (SRM) COMGBJ-10-0004/COMGEA-
10-0001, ``Use of Risk Insights to Enhance Safety Focus of Small
Modular Reactor Reviews,'' dated August 31, 2010 (ADAMS Accession No.
ML102510405), the Commission provided direction to the NRC staff on the
preparation for, and review of, small modular reactor (SMR)
applications, with a near-term focus on integral pressurized-water
reactor designs. The Commission directed the NRC staff to more fully
integrate the use of risk insights into pre-application activities and
the review of applications and, consistent with regulatory requirements
and Commission policy statements, to align the review focus and
resources to risk-significant structures, systems, and components and
other aspects of the design that contribute most to safety in order to
enhance the effectiveness and efficiency of the review process. The
Commission directed the NRC staff to develop a design-specific, risk-
informed review plan for each SMR design to address pre-application and
application review activities. An important part of this review plan is
the DSRS. The DSRS for the NuScale design is the result of the
implementation of the Commission's direction.
II. DSRS for the NuScale Design
The NuScale DSRS (available in ADAMS Package Accession No.
ML15355A295) reflects current NRC staff safety review methods and
practices which integrate risk insights and, where appropriate, lessons
learned from the NRC's reviews of DC and COL applications completed
since the last revision of the NUREG-0800, SRP Introduction, Part 2,
``Standard Review Plan for the Review of Safety Analysis Reports for
Nuclear Power Plants: Light-Water Small Modular Reactor Edition,''
January 2014 (ADAMS Accession No. ML13207A315). The NuScale DSRS Scope
and Safety Review Matrix provides a complete list of SRP sections and
identifies which SRP sections will be used for DC, COL, or ESP reviews
concerning the NuScale design; which SRP sections are not applicable to
the NuScale design; which SRP sections needed modification and were
reissued as DSRS sections; and which new DSRS sections were added to
address a unique design consideration in the NuScale design. The final
NuScale DSRS Scope and Safety Review Matrix is available in ADAMS under
Accession No. ML16263A000.
The NRC staff developed the content of the NuScale DSRS as an
alternative method for evaluating a NuScale-specific application and
has determined that the application may address the DSRS in lieu of
addressing the SRP, with specified exceptions. These exceptions include
particular review areas in which the DSRS directs reviewers to consult
the SRP and others in which the SRP is used for the review as
identified in the final NuScale DSRS Scope and Safety Review Matrix. If
NuScale chooses to address the DSRS, the application should identify
and describe all differences between the design features (DC and COL
applications only), analytical techniques, and procedural measures
proposed in an application and the guidance of the applicable DSRS
section (or SRP section, as specified in the NuScale DSRS Scope and
Safety Review Matrix), and discuss how the proposed alternative
provides an acceptable method of complying with the regulations that
underlie the DSRS acceptance criteria. The staff has accepted the
content of the DSRS as an alternative method for evaluating whether an
application complies with NRC regulations for the NuScale Small Modular
Reactor applications, provided that the application does not deviate
significantly from the design and siting assumptions made by the NRC
staff while preparing the DSRS. If the design or siting assumptions in
a NuScale application deviate significantly from the design and siting
assumptions the staff used in preparing the DSRS, the staff will use
the more general guidance in the SRP, as specified in sections
52.17(a)(1)(xii), 52.47(a)(9), or 52.79(a)(41) of title 10 of the Code
of Federal Regulations, depending on the type of application.
Alternatively, the staff may supplement the DSRS section by adding
appropriate criteria to address new design or siting assumptions.
The NRC staff issued a Federal Register notice on June 30, 2015 (80
FR 37312), to request public comments on the draft NuScale DSRS Scope
and Safety review Matrix (ADAMS Accession No. ML15156B063) and the
individual NuScale-specific DSRS sections referenced in the table
included in the FRN. A correction Federal Register notice was published
on July 9, 2015 (80 FR 39454), to identify an additional draft DSRS
section for which comments were requested. In response, the NRC
received comments from: NuScale Power, LLC, by letter dated August 31,
2015 (ADAMS Accession No. ML15258A081), the Nuclear Energy Institute
(NEI) by letter dated August 31, 2015 (ADAMS Accession No.
ML15257A012), Mark Thomson by electronic submission dated August 31,
2015 (ADAMS Accession No. ML15292A309), an anonymous submitter by
electronic submission dated August 31, 2015 (ADAMS Accession No.
ML15292A310), an anonymous submitter by electronic submission dated
August 31, 2015 (ADAMS Accession No. ML15292A311), Clinton Ferrara by
electronic submission dated August 31, 2015 (ADAMS Accession No.
ML15292A333), and Paula Ferrara by electronic submission dated August
31, 2015, (ADAMS Accession No. ML15292A334). Several of these comments
have been previously discussed during public meetings held in support
of developing the draft DSRS sections. These comments and resolutions
have been documented in the DSRS Public Comment Resolution Matrices and
are publicly available (ADAMS Package Accession No. ML16083A615).
In the June 30, 2015 Federal Register notice, the NRC requested
public comments on 115 DSRS sections. The NRC staff determined whether
to develop a DSRS section after considering whether significant
differences in the functions, characteristics, or attributes of the
NuScale design required major revision of the related SRP section
guidance, or whether structures, systems, and components identified in
the NuScale design are unique and not addressed by the current SRP.
Following publication of the draft version of the DSRS sections, the
NRC staff revisited these criteria and determined, based on the most
recent NuScale design, that it is appropriate to use the related SRP
section in lieu of a draft DSRS section
[[Page 75451]]
in a number of cases. In these cases the draft DSRS sections have not
been issued as final, and the related SRP sections will be used for the
NuScale review. In deciding to use the related SRP sections, the staff
has not necessarily determined that the SRP sections are wholly
applicable without modification. For example, as the NRC staff gains
greater understanding of the NuScale design or if the design changes
during the review, the staff would assess whether different or
supplemental review criteria are needed. Stakeholders who believe that
different or supplemental review criteria are needed may provide these
views to the NRC staff for consideration during the application review
period.
The results of determinations to use the related SRP sections
rather than draft DSRS sections, along with other identified issues
with the draft NuScale DSRS Scope and Safety Review Matrix, are
documented in a separate ``transitional'' NuScale DSRS Scope and Safety
Review Matrix (ADAMS Accession No. ML16076A048). The ``transitional''
Matrix shows the differences between the draft and final NuScale DSRS
Scope and Safety Review Matrices and describes the reasons for these
differences. The resulting final list of DSRS titles with corresponding
section numbers and ADAMS references are provided in the table below
and in ADAMS Package Accession No. ML15355A295.
In the future, should additional SRP sections be developed, the
staff will determine at that time their applicability to the NuScale
design. In addition, the NRC disseminates information regarding current
safety issues and proposed solutions through various means, such as
generic communications and the process for treating generic safety
issues. When current issues are resolved, the staff will determine the
need, extent, and nature of revision that should be made to the SRP
and/or DSRS to reflect the new NRC guidance.
III. Availability of Documents
------------------------------------------------------------------------
Design-specific review ADAMS accession
Section standard title No.
------------------------------------------------------------------------
Matrix.................... NuScale DSRS Scope and ML16076A048
Safety Review Matrix
(Transitional).
Matrix.................... NuScale DSRS Scope and ML16263A000
Safety Review Matrix
(Final).
3.5.1.3................... Turbine Missiles......... ML15355A364
3.7.1..................... Seismic Design Parameters ML15355A384
3.7.2..................... Seismic System Analysis.. ML15355A389
3.7.3..................... Seismic Subsystem ML15355A402
Analysis.
3.8.2..................... Steel Containment........ ML15355A411
3.8.4..................... Other Seismic Category I ML15355A444
Structures.
3.8.5..................... Foundations.............. ML15355A451
3.11...................... Environmental ML15355A455
Qualification of
Mechanical and
Electrical Equipment.
4.4....................... Thermal and Hydraulic ML15355A468
Design.
5.2.4..................... Reactor Coolant Pressure ML15355A479
Boundary Inservice
Inspection and Testing.
5.2.5..................... Reactor Coolant Pressure ML15355A505
Boundary Leakage
Detection.
5.3.1..................... Reactor Vessel Materials. ML15355A513
5.3.2..................... Pressure-Temperature ML15355A526
Limits, Upper[dash]Shelf
Energy, and Pressurized
Thermal Shock.
5.3.3..................... Reactor Vessel Integrity. ML15355A530
5.4.2.1................... Steam Generator Materials ML15355A532
5.4.2.2................... Steam Generator Program.. ML15355A535
5.4.7..................... Decay Heat Removal (DHR) ML15355A536
System.
BTP 5-4................... Design Requirements of ML15355A313
the Decay Heat Removal
System.
6.2.1..................... Containment Functional ML15356A259
Design.
6.2.1.1.A................. Containments............. ML15355A544
6.2.1.3................... Mass and Energy Release ML15357A327
Analysis for Postulated
Loss-of-Coolant
Accidents (LOCAs).
6.2.1.4................... Mass and Energy Release ML15356A241
Analysis for Postulated
Secondary System Pipe
Ruptures.
6.2.2..................... Containment Heat Removal ML15356A267
Systems.
6.2.4..................... Containment Isolation ML15356A332
System.
6.2.5..................... Combustible Gas Control ML15356A356
in Containment.
6.2.6..................... Containment Leakage ML15356A388
Testing.
6.3....................... Emergency Core Cooling ML15356A393
System.
6.6....................... Inservice Inspection and ML15356A396
Testing of Class 2 and 3
Components.
7.0....................... Instrumentation and ML15356A416
Controls--Introduction
and Overview of Review
Process.
7.1....................... Instrumentation and ML15363A293
Controls--Fundamental
Design Principles.
7.2....................... Instrumentation and ML15363A347
Controls--System
Characteristics.
7.0, App A................ Instrumentation and ML15355A316
Controls--Hazard
Analysis.
7.0, App B................ Instrumentation and ML15355A318
Controls--System
Architecture.
7.0, App C................ Instrumentation and ML15355A319
Controls--Simplicity.
7.0, App D................ Instrumentation and ML15355A320
Controls--References.
8.1....................... Electric Power-- ML15356A473
Introduction.
8.2....................... Offsite Power System..... ML15356A516
8.3.1..................... AC Power Systems (Onsite) ML15356A533
8.3.2..................... DC Power Systems (Onsite) ML15356A552
8.4....................... Station Blackout......... ML15356A570
9.1.2..................... New and Spent Fuel ML15356A584
Storage.
9.1.3..................... Spent Fuel Pool Cooling ML15356A595
and Cleanup System.
9.3.4..................... Chemical and Volume ML15356A622
Control System.
9.3.6..................... Containment Evacuation ML15356A637
and Flooding Systems.
9.5.2..................... Communications Systems... ML15363A400
10.2.3.................... Turbine Rotor Integrity.. ML15356A700
10.3...................... Main Steam Supply System. ML15355A322
10.4.7.................... Condensate and Feedwater ML15355A331
System.
11.1...................... Source Terms............. ML15355A333
[[Page 75452]]
11.2...................... Liquid Waste Management ML15355A334
System.
11.3...................... Gaseous Waste Management ML15355A335
System.
11.4...................... Solid Waste Management ML15355A336
System.
11.5...................... Process and Effluent ML15355A337
Radiological Monitoring
Instrumentation and
Sampling Systems.
11.6...................... Guidance on ML15355A338
Instrumentation and
Control Design Features
for Process and Effluent
Radiological Monitoring,
and Area Radiation and
Airborne Radioactivity
Monitoring.
12.2...................... Radiation Sources........ ML15350A320
12.3-12.4................. Radiation Protection ML15350A339
Design Features.
12.5...................... Operational Radiation ML15350A341
Protection Program.
14.2...................... Initial Plant Test ML15355A339
Program--Design
Certification and New
License Applicants.
14.3.5.................... Instrumentation and ML15355A340
Controls--Inspections,
Tests, Analyses, and
Acceptance Criteria.
15.0...................... Introduction--Transient ML15355A302
and Accident Analyses.
15.0.3.................... Design Basis Accidents ML15355A341
Radiological Consequence
Analyses for NuScale SMR
Design.
15.1.1--15.1.4............ Decrease in FW ML15355A303
Temperature, Increase in
FW Flow, Increase in
Steam Flow and
Inadvertent Opening of
the Turbine Bypass
System or Inadvertent
Operation of the Decay
Heat Removal System.
15.1.5.................... Steam System Piping ML15355A304
Failures Inside and
Outside of Containment.
15.1.6.................... Loss of Containment ML15355A305
Vacuum.
15.2.1-15.2.5............. Loss of External Load; ML15355A306
Turbine Trip; Loss of
Condenser Vacuum;
Closure of Main Steam
Isolation Valve (BWR);
and Steam Pressure
Regulator Failure
(Closed).
15.2.6.................... Loss of Non-Emergency AC ML15363A348
Power to the Station
Auxiliaries.
15.2.7.................... Loss of Normal Feedwater ML15355A307
Flow.
15.2.8.................... Feedwater System Pipe ML15355A308
Breaks Inside and
Outside Containment.
15.5.1-15.5.2............. Chemical and Volume ML15363A397
Control System
Malfunction that
Increases Reactor
Coolant Inventory.
15.6.5.................... LOCAs Resulting From ML15355A309
Spectrum of Postulated
Piping Breaks Within the
Reactor Coolant Pressure
Boundary.
15.6.6.................... Inadvertent Opening of ML15355A310
the Emergency Core
Cooling System.
15.9A..................... Thermal-hydraulic ML15355A311
Stability.
16.0...................... Technical Specifications. ML15355A312
------------------------------------------------------------------------
Dated at Rockville, Maryland, this 21st day of October 2016.
For the Nuclear Regulatory Commission.
Frank Akstulewicz,
Director, Division of New Reactor Licensing, Office of New Reactors.
[FR Doc. 2016-26210 Filed 10-28-16; 8:45 am]
BILLING CODE 7590-01-P
