Economic Simplified Boiling-Water Reactor Design Certification, 61943-61988 [2014-24362]

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• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 79, Number 199 (Wednesday, October 15, 2014)]
[Rules and Regulations]
[Pages 61943-61988]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2014-24362]



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Vol. 79

Wednesday,

No. 199

October 15, 2014

Part II





 Nuclear Regulatory Commission





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10 CFR Part 52





Economic Simplified Boiling Water Reactor Design Certification; Final 
Rule

Federal Register / Vol. 79 , No. 199 / Wednesday, October 15, 2014 / 
Rules and Regulations

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Nuclear Regulatory Commission

10 CFR Part 52

[NRC-2010-0135]
RIN 3150-AI85


Economic Simplified Boiling-Water Reactor Design Certification

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is adopting a new 
rule certifying the Economic Simplified Boiling-Water Reactor (ESBWR) 
standard plant design. This action is necessary so that applicants or 
licensees intending to construct and operate an ESBWR design may do so 
by referencing this design certification rule (DCR). The applicant for 
certification of the ESBWR design is GE-Hitachi Nuclear Energy (GEH).

DATES: This final rule is effective on November 14, 2014. The 
incorporation by reference of certain publications listed in this 
regulation is approved by the Director of the Office of the Federal 
Register (OFR) as of November 14, 2014.

ADDRESSES: Please refer to Docket ID NRC-2010-0135 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 Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2010-0135. Address 
questions about NRC dockets to Carol Gallagher, telephone: 301-287-
3422; email: Carol.Gallagher@nrc.gov. For technical questions, contact 
the individuals 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. For 
the convenience of the reader, instructions about obtaining materials 
referenced in this document are provided in a table in Section VII, 
``Availability of Documents,'' 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: George M. Tartal, Office of New 
Reactors, telephone: 301-415-0016, email: George.Tartal@nrc.gov; or 
David Misenhimer, Office of New Reactors, telephone: 301-415-6590, 
email: David.Misenhimer@nrc.gov; U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001.

SUPPLEMENTARY INFORMATION:

Executive Summary

A. Need for the Regulatory Action

    The NRC is amending its regulations related to licenses, 
certifications, and approvals for nuclear power plants. This final rule 
certifies the ESBWR standard plant design. This action is necessary so 
that applicants or licensees intending to construct and operate an 
ESBWR design may do so by referencing this DCR.

B. Major Provisions

    Major provisions of the final rule include changes to:
     specify which documents contain the requirements for the 
ESBWR design,
     specify how a nuclear power plant license applicant can 
reference the ESBWR design,
     describe how the NRC considers matters within the scope of 
the design to be resolved for proceedings involving a license or 
application referencing the ESBWR design, and
     describe the processes for changes to and departures from 
the ESBWR design.

C. Costs and Benefits

    The NRC did not prepare a regulatory analysis to determine the 
expected quantitative or qualitative costs and benefits of the final 
rule. The NRC prepares regulatory analyses for rulemakings that 
establish generic regulatory requirements applicable to all licensees. 
Design certifications are not generic rulemakings in the sense that 
design certifications do not establish standards or requirements with 
which all licensees must comply. Rather, design certifications are NRC 
approvals of specific nuclear power plant designs by rulemaking, which 
then may be voluntarily referenced by an applicant for a combined 
license (COL). Furthermore, design certification rulemakings are 
initiated by an applicant for a design certification, rather than the 
NRC. Preparation of a regulatory analysis in this circumstance would 
not be useful because the design to be certified is proposed by the 
applicant rather than the NRC. For these reasons, the NRC concludes 
that preparation of a regulatory analysis is neither required nor 
appropriate.

Table of Contents

I. Background
II. Summary and Analysis of Public Comments on the ESBWR Proposed 
Rule and Supplemental Proposed Rule
    A. Overview of Public Comments
    B. Comments Regarding Technical Content in the Design Control 
Document
    C. Comments Regarding NRC's Response to Fukushima Dai-ichi 
Accident
III. Regulatory and Policy Issues
    A. How the ESBWR Design Addresses Fukushima Near Term Task Force 
(NTTF) Recommendations
    B. Incorporation by Reference of Public Documents and Issue 
Resolution Associated With Non-Public Documents
    C. Changes to Tier 2* Information
    D. Change Control for Severe Accident Design Features
    E. Access to Safeguards Information (SGI) and Sensitive 
Unclassified Non-Safeguards Information (SUNSI)
    F. Human Factors Engineering (HFE) Operational Program Elements 
Exclusion From Finality
    G. Other Changes to the ESBWR Rule Language and Difference 
Between the ESBWR Rule and Other DCRs
IV. Technical Issues
    A. Regulatory Treatment of Nonsafety Systems (RTNSS)
    B. Containment Performance
    C. Control Room Cooling
    D. Feedwater Temperature Operating Domain
    E. Steam Dryer Analysis Methodology
    F. Aircraft Impact Assessment (AIA)
    G. American Society of Mechanical Engineers (ASME) Code Case N-
782
    H. Exemption for the Safety Parameter Display System
    I. Hurricane-Generated Winds and Missiles
    J. Loss of One or More Phases of Offsite Power
    K. Spent Fuel Assembly Integrity in Spent Fuel Racks
    L. Turbine Building Offgas System Design Requirements
    M. ASME Boiler and Pressure Vessel Code (BPV Code) Statement in 
Chapter 1 of the ESBWR Design Control Document (DCD)
    N. Clarification of ASME Component Design Inspections, Tests, 
Analyses, and Acceptance Criteria (ITAACs)
    O. Corrections, Editorial, and Conforming Changes
V. Rulemaking Procedure
    A. Exclusions From Issue Finality and Issue Resolution for Spent 
Fuel Pool Instrumentation
    B. Incorporation by Reference of Public Documents
    C. Changes to Tier 2* Information
    D. Other Changes to the ESBWR Rule Language and Difference From 
Other DCRs
    E. Exclusions From Issue Finality and Issue Resolution for 
Hurricane-Generated Winds and Missiles

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    F. Loss of One or More Phases of Offsite Power
    G. Spent Fuel Assembly Integrity in Spent Fuel Racks
    H. Turbine Building Offgas System Design Requirements
    I. ASME BPV Code Statement in Chapter 1 of the ESBWR DCD
    J. Clarification of ASME Component Design Inspections, Tests, 
Analyses, and Acceptance Criteria (ITAACs)
    K. Changes to the Supplemental FSER After Publication of the 
Supplemental Proposed Rule
    L. Corrections, Editorial, and Conforming Changes
VI. Planned Withdrawal of the ESBWR Standard Design Approval (SDA)
VII. Section-by-Section Analysis
    A. Introduction (Section I)
    B. Definitions (Section II)
    C. Scope and Contents (Section III)
    D. Additional Requirements and Restrictions (Section IV)
    E. Applicable Regulations (Section V)
    F. Issue Resolution (Section VI)
    G. Duration of This Appendix (Section VII)
    H. Processes for Changes and Departures (Section VIII)
    I. Inspections, Tests, Analyses, and Acceptance Criteria 
(Section IX)
    J. Records and Reporting (Section X)
VIII. Agreement State Compatibility
IX. Availability of Documents
X. Voluntary Consensus Standards
XI. Finding of No Significant Environmental Impact: Availability
XII. Paperwork Reduction Act
XIII. Regulatory Analysis
XIV. Regulatory Flexibility Certification
XV. Backfitting and Issue Finality
XVI. Congressional Review Act
XVII. Plain Writing
XVIII. Availability of Guidance

I. Background

    Part 52 of Title 10 of the Code of Federal Regulations (10 CFR), 
``Licenses, Certifications, and Approvals for Nuclear Power Plants,'' 
subpart B, presents the process for obtaining standard design 
certifications. On August 24, 2005, GEH tendered its application for 
certification of the ESBWR standard plant design (ADAMS Accession No. 
ML052450245) with the NRC. The NRC published a notice of receipt of the 
application in the Federal Register (70 FR 56745; September 28, 2005). 
GEH submitted this application in accordance with subpart B of 10 CFR 
part 52. On December 1, 2005, the NRC formally accepted the application 
as a docketed application for design certification (Docket No. 52-010) 
(70 FR 73311; December 9, 2005). The pre-application information 
submitted before the NRC formally accepted the application can be found 
in ADAMS under Docket No. PROJ0717 (Project No. 717).
    The NRC staff issued a final safety evaluation report (FSER) for 
the ESBWR design in March 2011. The FSER is available in ADAMS under 
Accession No. ML103470210. The NRC subsequently published the FSER in 
April 2014 as NUREG-1966, ``Final Safety Evaluation Report Related to 
the Certification of the Economic Simplified Boiling-Water Reactor 
Standard Design'' (ADAMS Accession No. ML14100A304). The NRC also 
published a proposed rule to certify the ESBWR design in the Federal 
Register on March 24, 2011 (76 FR 16549), and a supplemental proposed 
rule on May 6, 2014 (79 FR 25715). The FSER and the proposed rule were 
based on the NRC's review of Revision 9 of the ESBWR DCD.
    On April 17, 2014, the NRC issued an advanced supplemental safety 
evaluation report (SER) (ADAMS Accession No. ML14043A134) to address 
several matters identified by the NRC and revisions to the ESBWR DCD in 
Revision 10. The advanced supplemental SER was referenced in the 
supplemental proposed rule (79 FR 25715; May 6, 2014). The supplemental 
FSER will be published as Supplement No. 1 to NUREG-1966 before this 
final rule becomes effective. Because Revision 10 of the DCD was issued 
after the ESBWR proposed rule was published, all of the substantive 
changes in Revision 10 of the DCD are addressed in the SUPPLEMENTARY 
INFORMATION section of this document, including a discussion of why the 
change was or was not addressed in a supplemental proposed rule.
    In its application for design certification, GEH also requested the 
NRC to provide an SDA for the ESBWR design. An SDA for the ESBWR design 
was issued in March 2011 (ADAMS Accession No. ML110540310) following 
the NRC staff's issuance of the ESBWR FSER. On June 3, 2014, GEH 
requested that the NRC retire the SDA at the time of issuance of the 
final ESBWR design certification rule (ADAMS Accession No. 
ML14154A094). After this final rule is published, the NRC intends, as a 
separate action from this rulemaking, to withdraw the SDA.
    The application for design certification of the ESBWR design has 
been referenced in the following COL applications as of the date of 
this document: (1) Detroit Edison Company, Fermi Unit 3, Docket No. 52-
033 (73 FR 73350; December 2, 2008); (2) Dominion Virginia Power, North 
Anna Unit 3, Docket No. 52-017 (73 FR 6528; February 4, 2008); (3) 
Entergy Operations, Inc., Grand Gulf Unit 3, Docket No. 52-024 (73 FR 
22180; April 24, 2008) (APPLICATION SUSPENDED); (4) Entergy Operations, 
Inc., River Bend Unit 3, Docket No. 52-036 (73 FR 75141; December 10, 
2008) (APPLICATION SUSPENDED); and (5) Exelon Nuclear Texas Holdings, 
LLC, Victoria County Station Units 1 and 2, Docket Nos. 52-031 and 52-
032 (73 FR 66059; November 6, 2008) (APPLICATION WITHDRAWN).

II. Summary and Analysis of Public Comments on the ESBWR Proposed Rule 
and Supplemental Proposed Rule

A. Overview of Public Comments

    The NRC published a proposed rule to certify the ESBWR design in 
the Federal Register on March 24, 2011 (76 FR 16549). The period for 
submitting comments on the proposed DCR, ESBWR DCD, or draft 
environmental assessment (EA) closed on June 7, 2011. The NRC received 
a total of 10 public comments on the proposed rule. The types of 
comments, the organization of comments, the comment identification 
format, and comment responses follow.
    The NRC also published a supplemental proposed rule to request 
public comments on two specific topics regarding the ESBWR design 
certification. The supplemental proposed rule was published in the 
Federal Register on May 6, 2014 (79 FR 25715). The period for 
submitting comments on these specific topics closed on June 5, 2014. 
The NRC received no public comments on the supplemental proposed rule.
Types of Comments
    The NRC received two types of comment submissions on the proposed 
rule for the ESBWR design certification. A comment submission means a 
communication or document, submitted to the NRC by an individual or 
entity, with one or more individual comments addressing a subject or an 
issue. The two types of comment submissions were:
    1. Comment submissions that were not identical or similar in 
content (unique comment submissions); and
    2. Comment submissions self-characterized as ``petitions'' or 
comment submissions related to such ``petitions'' (petitions).
    The NRC received four unique comment submissions, including three 
comment submissions from private citizens and one comment submission 
from a non-government organization. Table 1 provides summary 
information on the unique comment submissions and their ADAMS Accession 
numbers.
    In addition, in light of the Fukushima Dai-ichi accident and during 
the public comment period on the proposed rule, the NRC received a 
series of petitions to suspend adjudicatory, licensing, and

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rulemaking activities, including the ESBWR design certification 
rulemaking. The NRC subsequently authorized responsive and supplemental 
filings on these petitions. In its Memorandum and Order, CLI-11-05, 
September 9, 2011, 74 NRC 141 (2011) (this decision is available on the 
NRC Web site in Volume 74 at https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr0750/), the Commission addressed the 
petitions and the responsive and supplemental filings and determined 
that the petitions should be denied in the relevant adjudicatory 
proceedings; and, on its own motion referred the petitions to the NRC 
staff for consideration as comments in the ESBWR rulemaking. The staff 
considered the petitions and the responsive and supplemental filings 
and identified six comment submissions applicable to the ESBWR 
rulemaking. Table 2 provides summary information on these ``petition-
related'' comment submissions and their ADAMS Accession numbers. Four 
of those comment submissions were ``petitions'' filed during the public 
comment period. One of the comment submissions was a responsive filing 
to the ``petitions.''
    The sixth of these comment submissions, self-characterized as a 
``petition'' and referred to the NRC staff in CLI-11-05, was received 
on August 15, 2011, after the close of the public comment period. As 
stated in the proposed rule, comments received after June 7, 2011, 
``will be considered if it is practical to do so, but assurance of 
consideration cannot be given'' to comments received after this date. 
The NRC determined that it was practical to consider this comment. This 
comment opposed issuance of the final ESBWR rule.

                   Table 1--Unique Comment Submissions
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 Comment submission No.          Commenter          ADAMS Accession No.
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1.......................  Paul Daugherty.........  ML110880057
2.......................  Farouk Baxter..........  ML110880315
3.......................  Patricia T. Birnie,      ML11158A088
                           Chairman, General
                           Electric Stockholders'
                           Alliance.
4.......................  Anonymous..............  ML11187A303
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    Table 2--Comment Submissions Self-Characterized as Petitions and
                           Responsive Filings
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 Comment submission No.          Commenter          ADAMS Accession No.
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1 (Note 1)..............  Various organizations    ML111040472
                           and individuals.
2 (Note 1)..............  Various organizations    ML111080855
                           and individuals.
3.......................  Various organizations    ML111100618
                           and individuals.
4.......................  Jerald G. Head, Senior   ML11124A103
                           VP, Regulatory
                           Affairs, GE Hitachi
                           Nuclear Energy.
5.......................  Various organizations    ML111260637
                           and individuals.
6.......................  ESBWR Intervenors......  ML112430118
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Note 1: Petition comment submission 2 was submitted as an amendment to
  petition comment submission 1. Therefore, the NRC is only addressing
  comments on petition comment submission 2 in this final rule and no
  further response is needed on petition comment submission 1.

Organization of Comments and Responses
    Comments and the NRC's responses are organized into two categories: 
Comments on technical issues presented in the DCD, and comments 
regarding Fukushima lessons learned. Comments on technical issues 
include the inclusion of beyond-design-basis accidents into the design, 
design of the ancillary diesel generators, safety-related battery 
design, control rod drive design, and control room flood protection. 
Comments regarding Fukushima lessons learned include delaying 
certification of the ESBWR design until lessons learned have been 
incorporated and the NRC's obligation under the National Environmental 
Policy Act (NEPA) to evaluate new information (such as the NTTF report, 
ADAMS Accession No. ML111861807) relevant to the environmental impact 
of its actions prior to certifying the ESBWR design. The NRC received 
comments related to the draft EA for this rule but those comments did 
not include anything to suggest that: (i) A rule certifying the ESBWR 
standard design would be a major Federal action, or (ii) the severe 
accident mitigation design alternatives (SAMDA) evaluation omitted a 
design alternative that should have been considered or incorrectly 
considered the costs and benefits of the alternatives it did consider. 
Therefore, no change to the EA was warranted. The NRC received no 
comments on the two specific topics in the supplemental proposed rule. 
The detailed comment summaries and the NRC's responses are provided in 
Sections II.B and II.C of this document.
Comment Identification Format
    All comments are identified uniquely by using the format [W][X]-
[Y], where:
    [W] represents the comment submission type (S = unique comment 
submission, P = petition).
    [X] represents the comment submission identification number (refer 
to the comment submission tables).
    [Y] represents the comment number, which the NRC assigned to the 
comment. In some instances, lower-case alphabetic characters [Ya, Yb, 
Yc * * *] were added to a comment number after the initial designation 
of comments.
    The NRC has created a document (ADAMS Accession No. ML113130141) 
which compiles all comment submissions and annotates each comment 
submission with the comment number indicated in the right hand margin.

B. Comments Regarding Technical Content in the DCD

Design-Basis Accidents
    Comment: Beyond-Design-Basis Accidents (DBAs) should be included in 
the design, final safety analysis report (FSAR), and Technical 
Specifications (TS). (S1-1)
    NRC Response: The NRC agrees that beyond-DBAs should be considered 
in the ESBWR design and the FSAR. In its 1985 policy statement on 
severe accidents (50 FR 32138), the Commission defined the term 
``severe accident'' as an event that is ``beyond

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the substantial coverage of design basis events,'' (DBE) including 
events in which there is substantial damage to the reactor core 
(whether or not there are serious offsite consequences). Consistent 
with the objectives of standardization and early resolution of design 
issues, 10 CFR 52.47(a)(23) requires applicants for design 
certification to include a description and analysis of severe accident 
prevention and mitigation features in the new reactor designs. These 
features are discussed in Chapter 19 of the DCD (equivalent to an 
FSAR), and the staff's evaluation of them is found in Chapter 19 of the 
FSER.
    The NRC disagrees that beyond-DBAs should be included in the TS. 
The TS prescribe safety limits, limiting safety system settings, 
limiting conditions for operation, surveillance requirements, and 
administrative controls associated with DBEs, but need not prescribe 
limits or settings for conditions that could be experienced during a 
beyond-DBE.
    No change was made to the rule, the DCD, or the EA as a result of 
this comment.
    Comment: The NRC's current regulatory scheme requires significant 
re-evaluation and revision in order to expand or upgrade the design-
basis for reactor safety as recommended by its NTTF report. (P6-1)
    NRC Response: The NRC considers this comment to be outside the 
scope of the ESBWR design certification rulemaking. The comment deals 
with the adequacy of the NRC's overall regulatory scheme for nuclear 
power reactors and does not directly address the adequacy of the ESBWR 
design certification.
    Nonetheless, the NRC disagrees with the comment. The NRC's rules 
and regulations provide reasonable assurance of adequate protection of 
public health and safety and the common defense and security. However, 
the Commission has ``initiated a comprehensive examination of the 
implications of the Fukushima accident. . . . As a result [of that 
examination], the NRC may implement changes to its regulations and 
regulatory processes.'' CLI-11-05, 74 NRC at 168. If such changes are 
warranted, the NRC's ``regulatory processes provide sufficient time and 
avenues to ensure that design certifications and COLs satisfy any 
Commission-directed changes before any new power plant commences 
operations. . . . Whether [the Commission] adopt[s] the Task Force 
recommendations or require[s] more, or different, actions associated 
with certified designs or COL applications, [the Commission has] the 
authority to ensure that certified designs and combined licenses 
include appropriate Commission-directed changes before operation.'' Id. 
at 162-163.
    No change was made to the rule, the DCD, or the EA as a result of 
this comment.
    Comment: The ESBWR environmental documents do not address the 
radiological consequences of DBAs or demonstrate that those reactors 
can be operated without undue risk to the health and safety of the 
public and conclude that any health effects resulting from the DBAs are 
negligible. This conclusion is based on a review of the DBAs considered 
in the ESBWR DCD (WEC 2008) and NUREG-0800, Standard Review Plan (SRP). 
The findings of the Fukushima NTTF report call into question whether 
this represents a full, accurate description and examination of all 
DBAs having the potential for releases to the environment. See 
Makhijani Declaration at 7. If the design-basis for the reactors does 
not incorporate accidents that should be considered in order to satisfy 
the adequate protection standard, then it is not possible to reach a 
conclusion that the design of the reactor adequately protects against 
accident risks. See Makhijani Declaration at 9. (P6-3)
    NRC Response: The NRC disagrees with this comment. The NRC notes 
that the Makhijani Declaration citations do not address DBAs as 
discussed in the comment, but rather the declaration specifically 
refers to beyond-DBEs. The NRC interprets the comment to be referring 
to the environmental report required to be provided by the design 
certification applicant per 10 CFR 52.47, ``Contents of applications; 
technical information,'' and 10 CFR 51.55, ``Environmental report--
standard design certification.'' The environmental report (NEDO-33306; 
ADAMS Accession No. ML102990433) referenced in Chapter 19 of the ESBWR 
DCD and evaluated in Chapter 19 of the FSER, as well as the NRC's EA, 
addresses costs and benefits of severe accident mitigation design 
alternatives. Conversely, DBAs for the ESBWR, and their associated 
radiological consequences, are not addressed in the environmental 
report but rather are addressed in Chapter 15 of the ESBWR DCD and 
evaluated in Chapter 15 of the FSER. The environmental report addresses 
the costs and benefits of severe accident mitigation design 
alternatives but does not address the design basis accidents discussed 
in the comment. In any event, the Commission has stated that, if 
warranted and after ``a comprehensive examination of the implications 
of the Fukushima accident . . ., the NRC may implement changes to its 
regulations and regulatory processes.'' CLI-11-05, 74 NRC at 168. The 
NRC's ``regulatory processes provide sufficient time and avenues to 
ensure that design certifications and COLs satisfy any Commission-
directed changes before any new power plant commences operations. . . 
.'' Id. at 162-163.
    No change was made to the rule, the DCD, or the EA as a result of 
this comment.
Electrical Systems
    Comment: The ESBWR design is flawed because it has failed to comply 
with the requirements of Institute of Electrical and Electronics 
Engineers (IEEE) Standard 603, which requires the electrical portion of 
the safety systems that perform safety functions--specifically, 
alternating current (ac) power from the Ancillary Diesel Generators 
(ADGs)--be classified as Class 1E. The DCD acknowledges that ac power 
from the ADGs is not needed for the first 72 hours of an accident, but 
are needed to perform Class 1E functions (recharging the Class 1E 
direct current (dc) batteries that provide power during the first 72 
hours of an accident) when no other sources of power are available. The 
ESBWR design has classified these ac power sources as commercial grade, 
nonsafety-related, and non-Class 1E (S2-1, referencing ADAMS Accession 
No. ML102350160).
    NRC Response: The NRC disagrees with the comment. The NRC's 
position remains as stated in the separate correspondence between the 
commenter and the NRC that is attached to the comment letter. 
Specifically, the NRC stated that the events described in the 
commenter's previous letters (no ac power available to the plant for 72 
hours after initiation of the accident and all batteries are depleted) 
are not DBEs but are beyond design-basis, for which the requirements of 
IEEE Standard 603 do not apply. As stated in the staff requirements 
memorandum (SRM), dated January 15, 1997, concerning SECY-96-128, 
``Policy and Key Technical Issues Pertaining to the Westinghouse AP600 
Standardized Passive Reactor Design,'' dated June 12, 1996, the 
Commission approved Item IV--Post-72 Hour Actions. The approval 
specified that the post-72 hour systems, structures, and components 
(SSCs) are not required to be safety-related. In addition, as stated in 
NUREG-1242, Volume 3, Part 1, ``NRC Review of Electric Power Research 
Institute's Advanced Light Water Reactor Utility Requirements Document: 
Passive Plant

[[Page 61948]]

Designs, Chapter 1,'' August 1994, a passive advanced light-water 
reactor, such as the ESBWR design, need not include or rely upon an 
active safety-related ac power source to support safety system 
functions after 72 hours from the onset of an accident, but may rely on 
electrical power sources that are not safety-related after that time. 
Specifically, the ESBWR is designed so that safety-related passive 
systems are able to perform all safety functions for 72 hours after 
initiation of a DBE without the need for operator actions. The DBE is 
assumed to be resolved (except for long-term cooling) within 72 hours, 
and thus, the Class 1E batteries are designed for and need only 
function for 72 hours without being recharged.
    In the ESBWR, the ADGs, which are the subject of the commenter's 
concern, are not used to recharge the Class 1E batteries. Rather, the 
ADGs provide power directly to post accident monitoring 
instrumentation, main control room lighting, the reactor pressure 
vessel (RPV) makeup pump, and containment cooling systems, among 
others. After 72 hours, consistent with NUREG-1242, nonsafety-related 
systems other than the ADGs are used to replenish safety-related 
passive systems so that they will perform long-term core cooling and 
containment integrity functions. These nonsafety-related systems are 
designed in accordance with quality standards commensurate with the 
importance of these functions and that provide reasonable assurance 
they will function when needed. In the event that the ADGs are not 
available, the Seismic Category I firewater storage tanks and Seismic 
Category I diesel pump and fire protection piping can be used to 
provide post-accident makeup water to the Isolation Condenser and 
Passive Containment Cooling System (PCCS) pools and Spent Fuel Pool 
(SFP) using the Fuel and Auxiliary Plant Cooling System (FAPCS) for 
long-term cooling beyond 72 hours.
    The NRC also stated in its May 15, 2009, letter (in the referenced 
document) that the offsite power system, a nonsafety-related power 
source, is the preferred source of power for safety-related systems at 
all current plants. Further, the station blackout (SBO) rule, 10 CFR 
50.63, ``Loss of all alternating current power,'' does not require the 
use of safety-related alternative ac power sources to cope with an SBO. 
Therefore, neither of these ac power sources--offsite power or 
alternate ac power source--is required to be safety-related or 
classified as Class 1E under IEEE 603. Thus, the ADGs need not be 
classified as Class 1E power sources as well.
    In summary, the design bases of the passive safety systems are 
centered on the 72-hour capability and these safety-related systems 
must remain functional to assure the integrity of the reactor coolant 
pressure boundary and the capability to shut down the reactor and 
maintain it in a safe shutdown condition without operator action or 
support from nonsafety systems for the first 72 hours following the 
initiation of a DBE. Beyond 72 hours, these systems must continue to 
remain functional to provide such assurance for the following 4 days, 
with allowance for operator actions and support from nonsafety SSCs 
consistent with NUREG-1242.
    No change was made to the rule, the DCD, or the EA as a result of 
this comment.
    Comment: The NRC should require GEH to relocate the safety-related 
dc batteries and their related systems above grade level so that they 
are not subject to external flooding. This recommendation is supported 
by the following points:
    1. There is a fair chance of a failure of the dc supply as safety-
related battery banks (Class-1E grade batteries) are housed below grade 
in the reactor building, as well as their electrical penetration to 
primary containment. In a natural disaster they may not remain 
watertight, as water may enter through the doors and incapacitate the 
battery banks.
    2. Water may also enter the battery rooms if those doors are open 
for maintenance, testing, or replacement of cells.
    3. ESBWR emergency core cooling systems (ECCS) are dependent on 
this dc supply. If the dc supply is lost, emergency cooling and 
depressurization systems will fail. There is no diversity for the core 
cooling and depressurization systems if the dc supply fails. (S4-1)
    NRC Response: The NRC disagrees with the comment. The safety-
related dc batteries and their related systems do not need to be 
relocated above grade level. The NRC has reviewed the ESBWR DCD and has 
determined that the ESBWR safety-related SSCs (including the reactor 
building, which houses the dc batteries) are designed to withstand the 
effects of external flooding. With the exception of loads due to 
hurricane winds and wind-generated missiles beyond those considered in 
the ESBWR DCD, the NRC concluded that the ESBWR DCD meets the 
requirements of 10 CFR part 50, appendix A, ``General Design Criteria 
for Nuclear Power Plants,'' (GDC) 2, which requires the design bases of 
SSCs important to safety to include protection against natural 
phenomena (including earthquakes, tornadoes, floods, hurricanes, and 
tsunami) such that these SSCs will not lose the capability to perform 
their safety functions as a result of such phenomena. This conclusion 
is documented in the NRC's FSER for the ESBWR design.
    In the following paragraphs, the NRC addresses each of the three 
supporting points for the comment.
    Supporting Point 1: The NRC agrees that safety-related batteries 
are located below grade per the ESBWR DCD, Tier 2, Figure 1.2-2. This 
is acceptable because all components of safety-related dc electric 
systems are housed in structures which provide protection against 
external flood damage. The structures that may be subjected to a 
design-basis flood are designed to withstand the flood level by 
locating the plant grade elevation 1 ft. (0.30 m) above the flood level 
and incorporating structural provisions into the plant design to 
protect the SSCs from the postulated flood conditions. GEH's 
application for design certification was submitted with proposed 
vendor-specified site parameters. These values are provided in Table 
2.0-1 (Tier 2) and in Table 5.1-1 (Tier 1) of the DCD. For the ESBWR 
design, the maximum groundwater level is 2 ft. (0.61 m) below plant 
grade and the maximum flood level is 1 ft. (0.30 m) below plant grade. 
The ESBWR design was evaluated using the vendor-specified flood levels 
and found to be safe. All exterior access openings are above flood 
level. The flood design incorporates reinforced concrete walls designed 
to resist the static and dynamic forces of the design-basis flood and 
water stops at construction joints to prevent in-leakage. External 
surfaces below flood and ground water levels are waterproofed. 
Penetrations are sealed and also capable of withstanding the static and 
dynamic forces of the design-basis flood. Watertight doors provide 
physical separation of flood zones. In addition, the applicant has 
specified the site parameters, design characteristics, and any 
additional requirements and restrictions necessary for a COL applicant 
to ensure that safety-related SSCs will be adequately protected from 
the site-specific probable maximum flood conditions. Based on the 
evaluation in Section 3.4 of the FSER, the NRC concludes that the ESBWR 
design regarding flood protection provides reasonable assurance that 
safety-related SSCs (including the safety-related dc batteries and 
their

[[Page 61949]]

related systems) will maintain their structural integrity or are 
located within structures that will maintain their integrity, and will 
perform their intended safety functions when subjected to a design-
basis flood, and therefore, satisfy the requirements of GDC 2.
    Supporting Point 2: The comment stated that water may enter the 
battery rooms if the watertight doors are open for maintenance, 
testing, or replacement of the battery cells. The NRC agrees that this 
scenario is possible for one division of safety-related battery banks. 
The ESBWR TS, under limiting condition of operation 3.8.1, restricts 
maintenance, testing, or replacement of the battery cells during plant 
operation to only one required division of safety-related battery 
banks. In addition, the COL applicant is required to develop plant 
operating and maintenance procedures that provide control for 
activities that are important to the safe operation of the facility, 
including limiting conditions of operation. However, there are four 
divisions of safety-related battery banks, which are physically 
separated by concrete walls and watertight doors. Only two divisions of 
dc systems are required for safe shutdown of the plant. If one of the 
safety-related battery room doors is open during a flood, as suggested 
in the comment, the other batteries will still be adequately protected 
by design features for physical separation to ensure the safety-related 
SSCs can perform their functions.
    Supporting Point 3: The comment stated that the ESBWR ECCS is 
dependent on dc power, and if dc power is lost, emergency cooling and 
depressurization systems will fail. The ESBWR ECCS consists of the 
Gravity Driven Cooling System, the Isolation Condenser System, the 
Standby Liquid Control System, and the Automatic Depressurization 
System. The Gravity Driven Cooling System, Standby Liquid Control 
System, and the Automatic Depressurization System do rely on dc power 
for actuation (as pointed out in the comment). The four trains of 
Isolation Condenser System, on the other hand, automatically begin 
removal of decay heat and control RPV level above the top of active 
fuel upon loss of all ac and dc power because the only valve in the 
system relied upon to change position upon initiation of the system 
fails in the safe (open) position upon loss of power. Beginning 4 hours 
after the start of an accident, the Isolation Condenser System upper 
and lower header vent valves are opened periodically to remove non-
condensable gases to maintain optimum heat removal and allow continued 
reactor cooldown. These valves are solenoid-operated valves and rely 
upon electric power to open.
    The comment also suggests that there is no diversity for several 
systems that rely on the dc power supply. The NRC agrees that the 
Automatic Depressurization System, Gravity Driven Cooling System, the 
Suppression Pool Equalization Line Valves, and the Standby Liquid 
Control System all require safety-related dc power in order to perform 
their safety functions and therefore lack diversity in that regard, but 
does not agree that the Basemat Internal Melt Arrest Coolability 
(BiMAC) cooling system requires safety-related dc power to perform its 
safety function. As discussed below, the BiMAC cooling system--a non-
safety system--is designed to automatically fire squib valves and drain 
water to the area below the RPV upon sensing high temperatures in the 
BiMAC without dependence on any of the four safety-related power 
sources. Also, as discussed above, the four trains of the Isolation 
Condenser System automatically begin removal of decay heat and control 
RPV level above the top of active fuel upon loss of all ac and dc power 
because the only valve in the system relied upon to change position 
upon initiation of the system fails in the safe (open) position upon 
loss of power. Decay heat can be removed with the Isolation Condenser 
System for 72 hours without any additional action. The ESBWR is 
designed such that the Isolation Condenser System heat exchanger pool 
can be replenished after 72 hours with the diesel driven fire pump to 
allow continued cooling with the Isolation Condenser System. Safety-
related dc power is not needed to operate this pump. In light of these 
facts, the NRC concludes that the capability of the ESBWR to remove 
decay heat from the reactor core following an accident is sufficiently 
diverse. It should also be noted that the ESBWR safety-related 120 
volts ac uninterruptible power supply (UPS) input is normally supplied 
by offsite power or a nonsafety-related onsite power system. During a 
loss of offsite and nonsafety-related onsite power, the UPS gets its 
power from 250 volts dc batteries. The ESBWR design includes an offsite 
power system, nonsafety-related standby diesel generators, and ADGs, 
any of which can mitigate the consequences of an accident if available. 
Safety-related UPS systems are housed in seismic Category I structures 
and meet GDCs 2, 4, and 17.
    Common cause failure of the safety-related batteries in the ESBWR 
design would clearly be an event of substantial safety significance 
because dc power is used to power the distributed control and 
instrumentation system, which is used to actuate passive safety 
systems. However, the ESBWR design includes a number of defense-in-
depth features for reducing the likelihood of losing all ability to 
accomplish key safety functions. As previously stated, the Isolation 
Condenser System automatically begins removal of decay heat and 
controls RPV level above the top of active fuel upon loss of all ac and 
dc power. All safety divisions (including concrete walls and watertight 
doors that separate the four safety-related battery banks) are 
physically separated.
    The ESBWR design also includes design features specifically for the 
purpose of injecting water into the containment to flood the 
containment floor and cover core debris. The BiMAC cooling system is 
designed to automatically fire squib valves and drain water to the area 
below the RPV upon sensing high temperatures in the BiMAC, indicating 
core debris below the RPV. This occurs without operator action and 
without dependence on any of the four safety-related power sources.
    No change was made to the rule, the DCD, or the EA as a result of 
this comment.
Control Rod Drive System
    Comment: Two Control Rod Drives (CRD) are scrammed by one hydraulic 
control unit (HCU). A single failure of one HCU will affect the scram 
function of two CRDs. It is done for cost saving. This is not 
acceptable in a safety system. (S4-2)
    NRC Response: The NRC disagrees with the comment. In Section 4.6.3 
of the FSER, the NRC stated that a single failure in an HCU may result 
in the failure of two control rods. The DCD describes that the control 
rods are assigned to HCUs in a manner such that no 4X4 array of rods 
contain both rods connected to the same HCU. This arrangement assures 
that shutdown is achieved (among other things) assuming a single 
failure of an HCU. The NRC reviewed the effects of an HCU failure and 
concluded in Section 4.3 of the FSER that sufficient shutdown margin 
exists in the case of an HCU failure. In addition, TS 3.1.5 requires 
that all control rod scram accumulators are operable during Modes 1 
(Power Operation) and 2 (Start-Up). If an accumulator is inoperable, 
the associated control rod pair is declared inoperable and Limiting 
Condition of Operation (LCO) 3.1.3, Control Rod Operability, is 
entered. This would

[[Page 61950]]

result in requiring the affected control rod to be fully inserted and 
disarmed, thereby satisfying the intended function in accordance with 
actions of LCO 3.1.3. If an accumulator is inoperable, TS require the 
affected control rod to be inserted and hence the scram function of two 
CRDs is satisfied. Finally, the ESBWR has a diverse method to scram the 
reactor. An electric motor is provided for each CRD for scram in 
addition to the hydraulic scram using the accumulator. Accordingly, the 
NRC has determined that the CRD system design is adequate.
    No change was made to the rule, the DCD, or the EA as a result of 
this comment.
Control Room
    Comment: For safety reasons, the Control Room should be located at 
a sufficient height from the ground to prevent its flooding during a 
tsunami, tornado, hurricane, heavy rain, etc. (S4-3)
    NRC Response: The NRC agrees that the control room should be 
protected from flooding. GEH's application for SDA and design 
certification was submitted with proposed vendor-specified site 
parameters. The values for maximum groundwater is 2 feet (0.61 m) below 
plant grade as provided in Table 2.0-1 (Tier 2) of the DCD and the 
maximum flood level is 1 foot (0.30 m) below plant grade as provided in 
Table 5.1-1 (Tier 1) of the DCD.
    The ESBWR design was evaluated using the vendor-specified flood 
levels and found to be safe. As described in Chapter 3 of the DCD, the 
ESBWR construction incorporates several water proofing features: The 
external walls below groundwater and flood levels are designed to 
withstand hydrostatic loads, construction and expansion joints have 
water stops, external surfaces below groundwater and flood levels are 
waterproofed, penetrations below groundwater and flood levels are 
sealed, and there are no exterior openings below grade.
    If a COL application referencing the ESBWR design is submitted to 
the NRC, the COL applicant must demonstrate that the site-specific 
characteristics are bounded by the DCD site parameters. During the 
review of a COL application using this design, the staff will perform 
an independent analysis to verify that the flood levels and other 
relevant site characteristics are within the DCD parameters.
    No change was made to the rule, the DCD, or the EA as a result of 
this comment.
Spent Fuel Pool
    Comment: The ESBWR design has an elevated SFP. This is a 
particularly troublesome feature in common with the Mark I BWR design, 
which is the design of the Fukushima reactors. (P2-2)
    NRC Response: The NRC disagrees with this comment. The ESBWR SFP 
design is different from the Mark I BWR design in that the ESBWR SFP is 
located entirely below grade. The ESBWR design does include an 
additional buffer pool located above grade in the reactor building. The 
buffer pool contains a small array of spent fuel racks that is used for 
temporary storage of spent fuel during refueling operations and also 
includes a location to store new fuel assemblies during power 
operations.
    GDC 2 requires that the ESBWR spent fuel storage facilities (SFP 
and buffer pool) and the structure within which they are housed, as 
SSCs important to safety, be protected against the effects of natural 
phenomena without loss of their safety function. In addition, GDC 61 
requires that the design prevents drainage of coolant inventory below 
an adequate shielding depth, provides adequate coolant flow to the 
spent fuel racks, and provides a system for detecting and containing 
pool liner leakage.
    The reactor building and the concrete containment, which houses the 
SFP and additional buffer pool, are seismic Category I structures that 
are designed to meet the requirements of GDC 2 for protection against 
natural phenomena such as an earthquake, tornado, or hurricane in 
combination with normal and accident condition loads considering the 
effects due to the elevated location of the buffer pool. Information 
relating to the analysis and design of the reactor building is provided 
in DCD Sections 3.7 and 3.8 and Appendices 3A, 3B, 3F, and 3G. Through 
analysis and review of the design, the NRC determined that the reactor 
building and the concrete containment are structurally adequate to 
withstand all design-basis loads. The NRC concluded in the FSER that 
both pools are adequately protected from the effects of natural 
phenomena without loss of capability to perform their safety functions.
    The NRC also concluded in its FSER that, because the SFP and buffer 
pools have anti-siphoning devices on all submerged Fuel and Auxiliary 
Pools Cooling System (FAPCS) piping, and there are no other drainage 
paths by which the level in the SFP or buffer pool could be reduced, 
coolant will not drain below an adequate shielding depth in either 
pool.
    Cooling of spent fuel located in either the SFP or buffer pool is 
provided by the FAPCS. In the unlikely event that a loss of active 
cooling to the spent fuel assemblies occurs, there is enough water to 
keep the fuel assemblies cooled for a minimum of 72 hours before 
operator actions are needed. After 72 hours, additional water can be 
provided through safety-related connections to the fire protection 
system or another onsite or offsite water source. The NRC concluded in 
the FSER that cooling for both ESBWR SFP and buffer pools will be 
maintained.
    Finally, the NRC concluded in the FSER that, because the spent fuel 
pool and buffer pool are equipped with stainless steel liners, concrete 
walls, and leak detection drains, both detection and containment of 
pool liner leakage capability are provided.
    No change was made to the rule, the DCD, or the EA as a result of 
this comment.

C. Comments Regarding the NRC's Response to Fukushima Dai-ichi Accident

    Some commenters favored delaying (in some fashion) the ESBWR 
rulemaking until lessons are learned from the Fukushima Dai-ichi 
Nuclear Power Plant (Fukushima) accident that occurred on March 11, 
2011, and the NRC applies the lessons learned to United States (U.S.) 
nuclear power plants, including the ESBWR design. Background on how the 
Commission responded to the Fukushima accident and how the ESBWR design 
addresses Fukushima NTTF recommendations is discussed in Section III of 
the SUPPLEMENTARY INFORMATION section of this document.
    As discussed in Section III of the SUPPLEMENTARY INFORMATION 
section of this document, the NRC concludes that no changes to the 
ESBWR design are warranted at this time to provide reasonable assurance 
of adequate protection of public health and safety. Moreover, even if 
the Commission concludes at a later time that some additional action is 
needed for the ESBWR design, the NRC has ample opportunity and legal 
authority to modify the ESBWR DCR to implement design changes, as well 
as to take any necessary action to ensure that COLs that reference the 
ESBWR make any necessary design changes.
    Comment: The NRC should suspend the certification of the ESBWR 
reactor design and rescind the final design approval it granted on 
March 9, 2011. Based on the recent events at the Fukushima Dai-ichi 
site, the NRC should first undertake a far more

[[Page 61951]]

rigorous, long-term review of the design and the regulatory implication 
of the events, implement new regulations to protect public health and 
safety, and revise the environmental analyses to evaluate the potential 
health, environmental and economic costs of reactor and SFP accidents. 
(S3-1, P3-1, P3-2)
    NRC Response: The NRC declines to suspend the ESBWR rulemaking. See 
Memorandum and Order, CLI-11-05, 74 NRC 141 (2011) (ADAMS Accession No. 
ML112521106).
    Background on how the Commission responded to the Fukushima 
accident and how the ESBWR design addresses Fukushima NTTF 
recommendations is discussed in Section III of the SUPPLEMENTARY 
INFORMATION section of this document. In that section, the NRC 
concludes that no changes to the ESBWR design are required at this time 
to provide reasonable assurance of adequate protection of public health 
and safety. If the Commission concludes at a later time that some 
additional action is needed for the ESBWR design, the NRC has ample 
opportunity and legal authority to modify the ESBWR DCR to implement 
design changes, as well as to take any necessary action to ensure that 
COLs that reference the ESBWR also make any necessary design changes.
    For these reasons the NRC does not regard delays in the ESBWR 
design certification process to be appropriate. No change was made to 
the rule, the DCD, or the EA as a result of this comment.
    Comment: The Atomic Energy Act (AEA) and NEPA preclude the NRC from 
approving standardized plant designs until it has completed the 
investigation of the Fukushima accident and considered the safety and 
environmental implications of the accident with respect to its 
regulatory program. NEPA imposes on agencies a continuing obligation to 
gather and evaluate new information relevant to the environmental 
impact of its actions. The need to supplement under NEPA when there is 
new and significant information is also found throughout the NRC 
regulations, e.g., 10 CFR 51.92(a)(2), 51.50(c)(iii), 51.53(b), and 
51.53(c)(3)(iv). The conclusions and recommendations presented in the 
NTTF report constitute ``new and significant information'' whose 
environmental implications must be considered before the NRC may 
certify the ESBWR design and operating procedures. (P2-2, P6-2)
    NRC Response: The NRC disagrees with this comment. The comment did 
not explain what particular provision of the AEA precludes the NRC from 
issuing a standard DCR. Furthermore, NEPA has no ``continuing 
obligation'' to gather and evaluate new information relevant to the 
environmental impact of its actions, because the Commission has 
determined that issuance of a standard DCR is not a major Federal 
action significantly affecting the quality of the human environment. 
See the EA at page 1 (ADAMS Accession No. ML111730382).
    No change was made to the rule, the DCD, or the EA as a result of 
this comment.
    Comment: The whole nuclear culture must be reviewed before any 
reactor designs are certified for potential construction, and that all 
licensing of new reactor designs be put on hold until the NRC's systems 
of regulations, oversight, and enforcement are thoroughly reviewed and, 
where required, are made more restrictive. (S3-2)
    NRC Response: The NRC considers this comment to be outside the 
scope of the ESBWR design certification rulemaking. The comment 
addresses overall nuclear industry safety culture and does not directly 
address the adequacy of the ESBWR design certification.
    Nonetheless, the NRC disagrees with the comment. The NRC considers 
that its regulatory framework and requirements provide a rigorous and 
comprehensive design certification and license review process that 
examines the full extent of siting, system design, and operations of 
nuclear power plants.
    The NRC will continue to process existing applications for new 
design certifications and licenses in accordance with the schedules 
that have been established.
    Background on how the Commission responded to the Fukushima 
accident and how the ESBWR design addresses Fukushima near-term task 
force recommendations is discussed in Section III of the SUPPLEMENTARY 
INFORMATION section of this document. In that section, the NRC 
concludes that no changes to the ESBWR design are warranted at this 
time to provide reasonable assurance of adequate protection of public 
health and safety. Moreover, even if the Commission concludes at a 
later time that some additional action is needed for the ESBWR design, 
the NRC has ample opportunity and legal authority to modify the ESBWR 
DCR to implement design changes, as well as to take any necessary 
action to ensure that COLs that reference the ESBWR also make any 
necessary design changes.
    For these reasons the NRC does not regard delays in the ESBWR 
design certification process to be appropriate. No change was made to 
the rule, the DCD, or the EA as a result of this comment.
    Comment: The NRC should include a review of public health 
challenges worldwide from radiation in its decision-making process. 
(S3-3)
    NRC Response: The NRC considers this comment to be outside the 
scope of the ESBWR DCR. The comment addresses the NRC's generic process 
and criteria for regulatory decision making, and does not directly 
address the adequacy of the ESBWR design.
    Nonetheless, the NRC disagrees with the comment. The NRC interprets 
the comment's reference to the ``decision-making process'' to mean the 
Commission's decision whether to certify the ESBWR design. The NRC 
reviewed the design and has found that it complies with the NRC's 
regulations, which provide reasonable assurance of adequate protection 
of public health and safety, including protection of the public from 
radiation. The comment did not provide any data, analyses, or other 
technical information to suggest why the EBSWR design would be unable 
to provide adequate protection of the public from radiation. No change 
was made to the rule, the DCD, or the EA as a result of this comment.
    Comment: The NTTF recommended that licensees reevaluate the seismic 
and flooding hazards at their sites and if necessary update the design-
basis and SSCs important to safety to protect against the updated 
hazards. NTTF Report, page 30. The ESBWR environmental documents must 
be supplemented in light of this new and significant information. The 
NTTF's findings and recommendations are directly relevant to 
environmental concerns and have a bearing on the proposed action and 
its impacts. They demonstrate a need to reevaluate the seismic and 
flooding hazards on the ESBWR reactors, the environmental consequences 
such hazards could pose, and what, if any, design measures could be 
implemented (i.e., through NEPA's requisite ``alternatives'' analysis) 
to ensure that the public is adequately protected from these risks. 
(P6-4)
    NRC Response: The NRC disagrees with the comment. Recommendation 2 
of the NTTF, which is the subject of the comment, was focused on 
licensees of nuclear power reactors and was addressed through site-
specific evaluations of the adequacy of the design of the reactors as 
applied to the site-specific seismic and flooding characteristics. By 
contrast, the ESBWR design certification--as any other design 
certification--is not approved for use on

[[Page 61952]]

any specific site. Rather, the ESBWR design specifies ``design 
parameters,'' including maximum flood levels and seismic ground motion 
frequencies and magnitudes, representing the values for which the NRC 
has determined the ESBWR may safely be placed. A nuclear power plant 
applicant intending to use the ESBWR must show that the actual site 
characteristics for the site that the applicant intends to use for the 
ESBWR fall within the ESBWR-specified design parameters. Thus, NTTF 
Recommendation 2 is not relevant to the adequacy of the ESBWR design 
certification. Rather, the NRC regards this NTTF recommendation as an 
issue relevant to the determination whether a referenced design 
certification has been adequately demonstrated to be appropriate at the 
COL applicant's designated site.
    In addition, the NRC does not agree that NTTF Recommendation 2 
demonstrates that the NRC must ``reevaluate the seismic and flooding 
hazards on the ESBWR reactors, the environmental consequences such 
hazards could pose, and what, if any, design measures could be 
implemented'' through a NEPA ``alternatives'' analysis. Recommendation 
2 of the NTTF can best be thought of as a determination to ensure that 
each site's seismic and flooding characteristics are adequately 
justified based upon current information. The recommendation does not 
concern the adequacy of the NRC's substantive regulatory requirements 
governing protection against seismic and flooding events or their 
application to any specific reactor design (such as the ESBWR). Thus, 
even if Recommendation 2 were adopted in full by the Commission and 
fully implemented, those implementing actions would be directed at 
licensees of existing nuclear power plants and applicants for new 
nuclear power plants. The NRC's implementing actions would not be 
directed at the ESBWR design certification. For these reasons, the NRC 
does not agree with the comment that ESBWR's EA must be supplemented to 
address the NTTF Recommendation 2 and implementing actions.
    No change was made to the rule, the DCD, or the EA as a result of 
this comment.
    Comment: The NTTF report makes several significant findings when it 
comes to increasing and improving mitigation measures for new reactor 
designs and recommends a number of specific steps licensees could take 
in this regard. Accordingly, the ESBWR environmental report must be 
supplemented to consider the use of these additional mitigation 
measures to reduce the project's environmental impacts. See 40 CFR 
1502.14(f), 1502.16, 1508.25(b)(3). (P6-5)
    NRC Response: The NRC disagrees with the comment. The NTTF report 
explicitly states that by the ``nature of their passive designs and 
inherent 72-hour coping capability for core, containment, and SFP 
cooling with no operator action required, the ESBWR and AP1000 designs 
have many of the design features and attributes necessary to address 
the Task Force recommendations. The Task Force supports completing 
those design certification rulemaking activities without delay.'' (see 
NTTF Report, pages 71-72). Specifically, the NTTF report does not 
recommend any actions for the ESBWR design in the near term.
    NEPA's obligation to evaluate new information relevant to the 
environmental impact does not attach unless and until the Commission 
determines whether ``new and significant'' information has arisen and 
there is a ``major Federal action'' being undertaken by the NRC for 
which the new information is relevant and material. The Commission has 
stated that ``[a]lthough the Task Force completed its review and 
provided its recommendations to us, the agency continues to evaluate 
the accident and its implications for U.S. facilities and the full 
picture of what happened at Fukushima is still far from clear. In 
short, we do not know today the full implications of the Japan event 
for U.S. facilities. Therefore, any generic NEPA duty--if one were 
appropriate at all--does not accrue now. If, however, new and 
significant information comes to light that requires consideration as 
part of the ongoing preparation of application-specific NEPA documents, 
the agency will assess the significance of that information as 
appropriate.'' CLI-11-05, 74 NRC at 167.
    No change was made to the rule, the DCD, or the EA as a result of 
this comment.
    Comment: Before certifying the ESBWR, the NRC must evaluate the 
relative costs and benefits of adopting all of the NTTF report 
recommendations, and specifically Recommendations 4 and 7, in light of 
the NRC's increased understanding regarding accident risks and the 
strength of its regulatory program to prevent or mitigate them. (P6-6)
    NRC Response: The NRC disagrees with the comment. The NTTF report 
explicitly states that by ``nature of their passive designs and 
inherent 72-hour coping capability for core, containment, and SFP 
cooling with no operator action required, the ESBWR and AP1000 designs 
have many of the design features and attributes necessary to address 
the Task Force recommendations. The Task Force supports completing 
those design certification rulemaking activities without delay.'' Id., 
at 71-72. Specifically, the NTTF report does not recommend any actions, 
to include Recommendations 4 and 7, for the ESBWR design in the near 
term. Any potential need to address these recommendations, by 
addressing ``prestaging of any needed equipment for beyond 72 hours,'' 
and the establishment of inspection, test, analysis, and acceptance 
criteria (ITAACs) ``to confirm effective implementation of minimum and 
extended coping, as described in detailed Recommendation 4.1'' of the 
NTTF report would be placed on COL applicants referencing the ESBWR 
design. Id., at 72.
    No change was made to the rule, the DCD, or the EA as a result of 
this comment.
    Comment: The comment questions the summary conclusions in Section 7 
of the NTTF report regarding Recommendations 4 and 7. Both of these 
recommendations are contrary to the certification process as currently 
followed by the NRC in which an applicant for a COL can incorporate by 
reference a certified reactor design. Directly contrary to this long-
standing process, the process suggested in the NTTF report pushes the 
Fukushima lessons learned onto a COL applicant rather than resolved 
these issues during the design certification process. Each reactor then 
becomes a prototype as case-by-case review of potential design and 
operational changes are made after construction begins. If the phrase 
``completing those design certification rulemaking activities without 
delay'' is an endorsement of the current rulemaking on the ESBWR DCD 
Revision 9 without consideration of the other Fukushima-driven 
recommendations (or the subsequent revision to the DCD), the comment 
questions the depth into which the NTTF analyzed the ESBWR reactor 
design. (P6-7)
    NRC Response: The NRC considers this comment to be outside the 
scope of the ESBWR design certification rulemaking. The comment 
presents the commenter's views on Recommendations 4 and 7 of the NTTF 
Report, but does not address the adequacy of the ESBWR design, the 
rule, or the EA.

[[Page 61953]]

    Nonetheless, the NRC disagrees with the comment. The NTTF 
suggestions that COL applicants or holders address Recommendations 4 
and 7, rather than the design certification applicant during the 
certification process, would not necessitate those COLs to be 
considered ``prototypes.'' The Commission has stated that ``the agency 
continues to evaluate the accident and its implications for U.S, 
facilities and the full picture of what happened at Fukushima is still 
far from clear. In short, we do not know today the full implications of 
the Japan event for U.S. facilities.'' CLI-11-05, 74 NRC at 167. Should 
changes need to be made to the ESBWR design as a result of the 
evaluation of the Fukushima event, the Commission has stated that ``we 
have the authority to ensure that certified designs and combined 
licenses include appropriate Commission-directed changes before 
operation.'' Id. at 163. Further, it is not contrary to the 
certification process to require changes resulting from Fukushima 
lessons learned on COLs. The NRC may, under 10 CFR 52.97(c), place 
conditions upon the COL that the ``Commission deems necessary and 
appropriate.'' Further, the requirements under 10 CFR 52.63(a)(1) 
provide a mechanism for the NRC to modify certified designs. Such 
design changes would be applied to all COL holders referencing this 
design under 10 CFR 52.63(a)(3). As a result, all COL holders 
referencing the certified design would be required to make such 
changes. Moreover, in appropriate (but relatively limited) 
circumstances the NRC could also impose changes as an ``administrative 
exemption'' to the issue finality provisions of 10 CFR 52.63 and the 
ESBWR analogous to what the NRC did in the aircraft impact assessment 
(AIA) final rule, 10 CFR 50.150 (72 FR 56287; October 3, 2007).
    No change was made to the rule, the DCD, or the EA as a result of 
this comment.
Emergency Petition
    NRC Note: The Emergency Petition is comment submissions P1 and P2 
in this ESBWR design certification rulemaking proceeding.
    Comment: The emergency petition is out of process and should be 
dismissed on that basis alone. However, if this petition is not so 
dismissed, the NRC should treat this petition, for aspects related to 
the single issue specifically regarding the ESBWR design certification 
rulemaking, as a public comment on the proposed rule. (P4-1)
    NRC Response: The NRC need not address, in this rulemaking, the 
comment's suggestion that the emergency petition is out of process 
because the Commission considered the merits of it and related filings 
in its Memorandum and Order, CLI-11-05, 74 NRC at 141 (2011) (ADAMS 
Accession No. ML112521106). The Commission determined that the 
Emergency Petition should be denied in the relevant adjudicatory 
proceedings and, on its own motion referred the emergency petition to 
the NRC staff for consideration as comments in the ESBWR rulemaking.
    To the extent that it is relevant to the ESBWR design certification 
rulemaking, the NRC agrees that the Emergency Petition should be 
treated as a public comment on the proposed rule. Comments in the 
Emergency Petition are addressed in this comment response portion of 
this statement of considerations for the final ESBWR DCR.
    No change was made to the rule, the DCD, or the EA as a result of 
this comment.
    Comment: The responses, filed by various industry representatives 
and COL applicants in accordance with an April 19, 2011, Commission 
Order (ADAMS Accession No. ML111101277) and setting forth those 
representatives' and applicants' views on an ``Emergency Petition'' 
(ADAMS Accession No. ML111080855), were based on mischaracterizations 
of the Emergency Petition, incorrect representations regarding the 
NRC's response to the Three Mile Island accident, and incorrect 
interpretations of the law. Therefore, the responses should be rejected 
and the Emergency Petition should be granted. (P5-1)
    NRC Response: On September 9, 2011, the Commission issued a 
Memorandum and Order on the Emergency Petition, CLI-11-05, 74 NRC 141 
(ADAMS Accession No. ML112521106), which referred both the Emergency 
Petition and certain documents filed with the NRC to the NRC staff for 
``consideration as comments'' in the applicable design certification 
rulemaking. CLI-11-05, 74 NRC at 176. Comment submission P5 was one of 
the documents referred by the Commission to the staff for consideration 
as comments. In accordance with the Commission's direction in CLI-11-
05, comment submission P5 has been considered in the ESBWR rulemaking 
in a manner consistent with other comment submissions filed in the 
ESBWR rulemaking. Thus, the NRC reviewed the submission to determine 
the nature of the comments within this comment submission, if it is 
within the scope of the ESBWR rulemaking, and if so, what substantive 
response is appropriate. Based upon that review, the NRC determined 
that comment submission P5 is essentially a procedural reply to 
responses filed by other entities on the Emergency Petition. The NRC 
has determined that the reply does not contain any new substantive 
comments on the adequacy of the ESBWR design that were not already 
presented in the Emergency Petition and, therefore, has concluded that 
no further response is needed. No change was made to the rule, the DCD, 
or the EA as a result of this comment.

III. Regulatory and Policy Issues

    This document addresses the regulatory and policy issues that were 
addressed in the March 2011 proposed rule, the May 2014 supplemental 
proposed rule, and those not addressed in either the proposed rule or 
the supplemental proposed rule. The regulatory and policy issues 
addressed in the March 2011 proposed rule are: (1) Access to safeguards 
information (SGI) and sensitive unclassified non-safeguards information 
(SUNSI), and (2) human factors engineering (HFE) operational program 
elements exclusion from finality. An additional regulatory and policy 
issue addressed in the May 2014 supplemental proposed rule is 
incorporation by reference of public documents and issue resolution 
associated with non-public documents. The NRC provided an opportunity 
for public comment in the supplemental proposed rule on the issue 
resolution associated with non-public documents, but not for 
incorporation by reference of public documents. A number of regulatory 
and policy issues were not included in either the March 2011 proposed 
rule or the May 2014 supplemental proposed rule. These are: (1) How the 
ESBWR design addresses Fukushima NTTF recommendations, (2) changes to 
Tier 2* information, (3) change control for severe accident design 
features, and (4) other changes to the ESBWR rule language and 
difference between the ESBWR rule and other DCRs.
    Each of these issues identified above is discussed below.\1\
---------------------------------------------------------------------------

    \1\ Some of the regulatory and policy issues discussed below 
arose after the close of the public comment period on the March 24, 
2011, proposed rule. The public was afforded an opportunity to 
comment on some of these issues in the May 16, 2014, supplemental 
proposed rule. Section V of the SUPPLEMENTARY INFORMATION section of 
this document describes the NRC's bases for not offering a comment 
opportunity for some of the regulatory and policy issues that arose 
after the close of the public comment period on the proposed rule.

---------------------------------------------------------------------------

[[Page 61954]]

A. How the ESBWR Design Addresses Fukushima NTTF Recommendations

    The application for certification of the ESBWR design was prepared 
and submitted, and the NRC staff's review of the application was 
completed, before the March 11, 2011, Great Tohoku earthquake and 
tsunami and subsequent events at the Fukushima Dai-ichi Nuclear Power 
Plant in Japan. In response to the events at Fukushima, the NRC 
established the NTTF to conduct a systematic and methodical review of 
NRC processes and regulations to: (1) Determine whether the agency 
should make additional improvements to its regulatory system; and (2) 
make recommendations to the Commission for policy directions. On July 
12, 2011, the NTTF issued a 90-day report, SECY-11-0093 (ADAMS 
Accession Number ML11186A950), ``Near Term Report and Recommendations 
for Agency Actions Following the Events in Japan,'' identifying 12 
recommendations. Among other recommendations, the NTTF supported 
completing the ESBWR design certification rulemaking activity without 
delay (see NTTF Report, pages 71-72).
    On September 9, 2011, in SECY-11-0124, ``Recommended Actions to Be 
Taken Without Delay from NTTF Report,'' (ADAMS Accession No. 
ML11245A144) the NRC staff submitted to the Commission for its 
consideration NTTF recommendations that should be partially or entirely 
initiated without delay. In SECY-11-0124, the NRC staff concluded that 
the following subset of actions would provide the greatest potential 
for improving safety in the near term:

(1) Recommendation 2.1: Seismic and Flood Hazard Reevaluations
(2) Recommendation 2.3: Seismic and Flood Walkdowns
(3) Recommendation 4.1: Station Blackout Regulatory Actions
(4) Recommendation 4.2: Equipment Covered under 10 CFR 50.54(hh)(2) 
(subsequently renamed ``Mitigation Strategies for Beyond-Design-Basis 
External Events'' with the issuance of Order EA-12-049)
(5) Recommendation 5.1: Reliable Hardened Vents for Mark I Containments
(6) Recommendation 8: Strengthening and Integration of Emergency 
Operating Procedures, Severe Accidents Management Guidelines, and 
Extensive Damage Mitigation Guidelines
(7) Recommendation 9.3: Emergency Preparedness Regulatory Actions 
(staffing and communications).

    On October 3, 2011, in SECY-11-0137, ``Prioritization of 
Recommended Actions To Be Taken in Response to Fukushima Lessons 
Learned'' (ADAMS Accession No. ML11272A203), the NRC staff identified 
two additional actions that would have the greatest potential for 
improving safety in the near term. The additional actions are: (1) 
Inclusion of Mark II containments in the staff's recommendation for 
reliable hardened vents associated with NTTF Recommendation 5.1 and (2) 
the implementation of SFP instrumentation proposed in Recommendation 
7.1.
    The NRC staff determined that the following two near term 
recommendations are applicable and should be considered for the ESBWR 
design certification: (1) Recommendation 4.2, Mitigation Strategies for 
Beyond-Design-Basis External Events (onsite equipment and connections 
only) and (2) Recommendation 7.1, SFP Instrumentation. The remaining 
Commission-approved near term recommendations are applicable only to 
COLs and existing plants (Recommendations 2.1 and 9.3), only to 
existing plants (Recommendations 2.3 and 5.1), or are planned to be 
addressed through rulemaking (Recommendations 4.1, 4.2, 7.1, 8, and 
9.3).
    On February 17, 2012, in SECY-12-0025, ``Proposed Orders and 
Requests for Information in Response to Lessons Learned from Japan's 
March 11, 2011, Great Tohoku Earthquake and Tsunami,'' (ADAMS Accession 
No. ML12039A103) the NRC staff provided the Commission with proposed 
orders and requests for information to be issued to all power reactor 
licensees and holders of construction permits. In SECY-12-0025, the 
staff indicated its intent to address similar requirements in its 
reviews of pending and future design certification and COL 
applications.
    On March 9, 2012, in the SRM to SECY-12-0025, the Commission 
approved issuing the proposed orders with some modifications. On March 
12, 2012, the NRC issued Order EA-12-049, ``Order Modifying Licenses 
with Regard to Requirements for Mitigation Strategies for Beyond-
Design-Basis External Events''; and Order EA 12-051, ``Order Modifying 
Licenses With Regard to Reliable Spent Fuel Pool Instrumentation'' to 
the appropriate licensees and permit holders (ADAMS Accession Nos. 
ML12054A735 and ML12054A679, respectively).
    The NRC staff provides 6-month updates to the Commission on all 
Fukushima-related activities, including the NTTF recommendations that 
will be addressed in the longer term. The latest update is provided in 
SECY-14-0046, ``Fifth 6-Month Status Update on Response to Lessons 
Learned from Japan's March 11, 2011, Great T[omacr]hoku Earthquake and 
Subsequent Tsunami,'' dated April 17, 2014 (ADAMS Accession No. 
ML14064A523).
    The NRC considered Recommendation 4.2, as modified by SRM-SECY-12-
0025, using the requirements in Order EA-12-049. SECY-12-0025 outlines 
a three-phase approach to developing the strategies. The initial phase 
requires the use of installed equipment and resources to maintain or 
restore core cooling, containment, and SFP cooling without alternating 
current power or loss of normal access to the ultimate heat sink. The 
transition phase requires providing sufficient, portable, onsite 
equipment and consumables to maintain or restore these functions until 
they can be accomplished with resources brought from offsite. The final 
phase requires obtaining sufficient offsite resources to sustain those 
functions indefinitely.
    As discussed in multiple sections of the DCD, and in the FSER, the 
ESBWR is designed such that the reactor core and associated coolant, 
control, and protection systems, including station batteries and other 
necessary support systems, provide sufficient capacity and capability 
to ensure that the core will be cooled and there will be appropriate 
containment integrity and adequate cooling for the spent fuel for 72 
hours in the event of an SBO--loss of all normal and emergency ac 
power.
    The ESBWR design credits the isolation condenser system for the 
first 72 hours of an event in which all ac power sources are lost. 
Beyond the first 72 hours, the isolation condenser system pool and SFP 
need to be refilled. The ESBWR design includes provisions to refill the 
isolation condenser system pool and SFP with onsite equipment without 
reliance on ac power, such as by the diesel-driven fire pump. In 
addition, after the first 72 hours of an event, accident mitigation is 
achieved through the ancillary diesel, which supplies ac power to 
various components such as: PCCS vent fans, motor driven fire pump, 
control room habitability area ventilation system air handling units, 
and emergency lighting. The standby diesels are also needed to support 
FAPCS operations. Both the ancillary and standby diesels supply short-
term and long-term safety loads.
    For the reasons set forth in Section 22.5 of the FSER, the NRC 
found that the applicant has included sufficient nonsafety-related 
equipment in the RTNSS program to ensure that safety

[[Page 61955]]

functions relied upon in the post-72-hour period are successful. 
Emergency procedures are to be developed by the COL applicant to 
support emergencies, which includes the period after 72 hours from the 
onset of the loss of all ac power. Further, the nonsafety-related 
equipment relied upon in the post-72-hour period has been designed in 
accordance with Commission policy (as described in Section 22.5.6.2 of 
the FSER) for use of augmented design standards for protection from 
external hazards and the NRC is engaging with COL applicants to ensure 
they have established appropriate availability controls for this 
equipment. Availability controls will be addressed in connection with a 
COL application referencing the ESBWR standard design.
    The ESBWR design supports a COL applicant refilling the pools with 
offsite equipment, such as local fire pumpers. In the period beyond 
seven days from the onset of the event, the COL applicant will be 
responsible for describing how it will make available offsite sources, 
such as diesel fuel oil for the ancillary and standby diesel generators 
and water makeup to support long term cooling. The COL applicant must 
address the ability of offsite support to sustain these functions 
indefinitely, including procedures, guidance, training and acquisition, 
staging or installing needed equipment. Therefore, the NRC concludes 
that the ESBWR design, as described in the DCD, satisfies the 
underlying purpose of Order EA-12-049 insofar as it includes additional 
equipment to maintain or restore core and spent fuel pool cooling and 
containment function in the event of the loss of all ac power. While 
the ESBWR design includes all of the necessary design features in this 
respect, the COL applicant must address the programmatic aspects of 
Order EA-12-049. The NRC staff has already engaged with COL applicants 
on these arrangements. To the extent a COL applicant proposes to rely 
on additional equipment to perform required functions in the event of a 
loss of all ac power, that equipment is outside the scope of the 
standard ESBWR design and the NRC staff will evaluate it in connection 
with the COL application.
    The NRC considered Recommendation 7.1, as modified by SRM-SECY-12-
0025, using the requirements in Order EA-12-051, which describes the 
key parameters to be used to determine that a level instrument is 
considered reliable. JLD-ISG-2012-03, Revision 0, ``Compliance with 
Order EA-12-051, Reliable Spent Fuel Pool Instrumentation,'' (ADAMS 
Accession No. ML12221A339) endorses with exceptions and clarifications 
the methodologies described in the industry guidance document NEI 12-
02, Revision 1, ``Industry Guidance for Compliance with NRC Order EA-
12-051, To Modify Licenses with Regard to Reliable Spent Fuel Pool 
Instrumentation,'' (ADAMS Accession No. ML122400399) and provides an 
acceptable approach for satisfying the applicable requirements.
    The NRC finds that the ESBWR design has design features that 
satisfy the underlying purpose of Order EA-12-051 for reliable SFP 
level instrumentation, except for two matters. The exceptions are 
whether the safety-related level instrumentation: (1) Are designed to 
allow the connection of an independent power source, and (2) will 
maintain its design accuracy following a power interruption or change 
in power source without recalibration. While the ESBWR design includes 
all of the necessary design features in this respect, the DCD did not 
include any information addressing these two matters. In addition, the 
NRC is currently developing a rulemaking which would address spent fuel 
pool instrumentation for beyond design basis events/accidents. This 
rulemaking may adopt different requirements than what is currently 
considered acceptable to meet the underlying purpose of order EA-12-051 
and its related guidance. For these reasons, the NRC is excluding from 
issue finality and issue resolution these two aspects of the ESBWR 
spent fuel pool instrumentation design features. The exclusions have 
two consequences. First, any combined license applicant referencing the 
ESBWR design certification rule will have to provide information 
demonstrating that the NRC's requirements on these two matters are met. 
Second, the NRC need not address the factors of 10 CFR 52.63 either 
when it reviews the combined license application for adequacy with 
respect to these two matters, or in connection with any amendment of 
the ESBWR design certification rule imposing requirements to govern 
those matters.

B. Incorporation by Reference of Public Documents and Issue Resolution 
Associated With Non-Public Documents

    In Section III, ``Scope and Contents,'' of the proposed ESBWR DCR 
(76 FR 16549; March 24, 2011), the only document for which the NRC 
proposed to obtain approval from the Office of the Federal Register 
(OFR) for incorporation by reference into the ESBWR design 
certification rule was the ESBWR DCD, Revision 9 (DCD Revision 9). Such 
approval would make DCD Revision 9 a legally-binding requirement on any 
referencing combined license applicant and holder by virtue of 
publication in the Federal Register as a final rule. This was based 
upon the assumption that the DCD specified all necessary requirements 
in Tier 1 and Tier 2 (with the exception of non-public documents 
containing proprietary information,\2\ security-related information,\3\ 
and SGI).
---------------------------------------------------------------------------

    \2\ For purposes of this discussion, ``proprietary information'' 
constitutes trade secrets or commercial or financial information 
that are privileged or confidential, as those terms are used under 
the Freedom of Information Act and the NRC's implementing regulation 
at 10 CFR part 9.
    \3\ For purposes of this discussion, ``security-related 
information'' means information subject to non-disclosure under 10 
CFR 2.390(a)(7)(vi).
---------------------------------------------------------------------------

    After the close of the public comment period, the NRC recognized 
that Tier 2, Section 1.6, ``Material Incorporated by Reference and 
General Reference Material,'' of the ESBWR DCD states that a number of 
documents are ``incorporated by reference'' into Tier 2 of the ESBWR 
DCD, and which contain information intended to be requirements. These 
documents were listed in Tables 1.6-1, ``Referenced GE/GEH Reports,'' 
and 1.6-2, ``Referenced non-GE/GEH Topical Reports,'' of the DCD 
Revision 9. Although some of the documents contain information which is 
intended to be requirements (based on the text of the DCD), neither 
Tables 1.6-1 and 1.6-2 of the DCD nor Section III of the proposed ESBWR 
design certification rule clearly stated which of these documents were 
intended as requirements. Documents intended as requirements (and which 
are publicly available) should have been listed in Section III of the 
ESBWR design certification rule as being approved for incorporation by 
reference by the Director of the OFR. Tables 1.6-1 and 1.6-2 also 
included documents that, although ``incorporated by reference'' into 
DCD Revision 9, were not intended to be requirements, but were 
references ``for information only.'' Thus, the ESBWR proposed rule did 
not clearly differentiate between these two different classes of 
documents. Finally, Tables 1.6-1 and 1.6-2 of DCD Revision 9 included 
both publicly-available documents and non-publicly available 
documents,\4\ but for some of the documents which were not publicly 
available, GEH had not created a publicly-available version of that 
document to support the public comment process. The creation of 
publicly-available versions of non-public documents to support the 
public commenting process and transparency has been a long-standing 
practice for

[[Page 61956]]

both design certification rulemakings and licensing actions.
---------------------------------------------------------------------------

    \4\ The non-publicly available documents contain proprietary, 
security-related, and/or safeguards information.
---------------------------------------------------------------------------

    To address the NRC's concerns, for those non-public documents which 
include information intended to be treated as requirements and for 
which publicly-available versions were not previously created, GEH 
created publicly-available versions of those non-public documents. GEH 
also submitted Revision 10 to the DCD (DCD Revision 10), which included 
three tables in Section 1.6 that superseded Tables 1.6-1 and 1.6-2 in 
DCD Revision 9. These three tables--Tables 1.6-1, ``GE/GEH Reports 
Incorporated by Reference,'' 1.6-2, ``Non-GE/GEH Reports Incorporated 
by Reference,'' and 1.6-3, ``Referenced Reports (not Incorporated by 
Reference,''--collectively clarify which documents are intended to be 
requirements and which documents are references only.
    The supplemental proposed rule (79 FR 25715; May 6, 2014): (1) 
Announced the availability of DCD Revision 10; (2) described the 
distinction between those documents intended as requirements versus 
those which were for information only; (3) requested public comments on 
the NRC's intent to treat 50 non-public, referenced documents in DCD 
Revision 10 (listed in Table 2 of the supplemental proposed rule) as 
requirements and matters resolved in subsequent licensing and 
enforcement actions for plants referencing the ESBWR design 
certification; and (4) clarified, but did not request public comments 
on, the NRC's intent to obtain approval for incorporation by reference 
from the Director of the OFR for both DCD Revision 10 and the 20 
publicly-available documents referenced in DCD Revision 10 (listed in 
Table 3 of the supplemental proposed rule), which are intended by the 
NRC to be requirements.
    The 50 non-publicly available documents listed in Table 3 below are 
considered by the NRC to be requirements applicable to any combined 
license applicant or holder of a combined license referencing the ESBWR 
design certification rule, where the language of DCD Revision 10 makes 
clear that any one of those documents is intended to be a requirement. 
In addition, the 50 non-public documents are within the scope of issue 
resolution under Section VI of Appendix E, and are accorded issue 
finality protection under that Section VI and 10 CFR 52.63.

Table 3--50 Non-Public Documents Which the NRC Regards as Requirements, Are Matters Resolved Under Paragraph VI,
      ISSUE RESOLUTION, of the ESBWR Design Certification Rule, and Are Accorded Issue Finality Protection
----------------------------------------------------------------------------------------------------------------
                                                                                                Non-publicly
            Document No.                   Document title           Publicly- available       available ADAMS
                                                                    ADAMS Accession No.        Accession No.
----------------------------------------------------------------------------------------------------------------
NEDE-33391, NEDO-33391.............  GE Hitachi Nuclear Energy,  ML14093A138.............  N/A (Safeguards
                                      ``ESBWR Safeguards                                    information cannot
                                      Assessment Report,'' NEDE-                            be placed in ADAMS)
                                      33391, Class III
                                      (Safeguards, Security-
                                      Related, and
                                      Proprietary), Revision 3,
                                      March 2010, and NEDO-
                                      33391, Class I (Non-
                                      safeguards, Non-security
                                      related, and Non-
                                      proprietary), Revision 3,
                                      March 2014.
NEDC-31959P, NEDO-31959............  GE Nuclear Energy, ``Fuel   ML14093A145.............  ML14093A146
                                      Rod Thermal-Mechanical
                                      Analysis Methodology
                                      (GSTRM),'' NEDC-31959P
                                      (Proprietary), April
                                      1991, and NEDO-31959 (Non-
                                      proprietary), April 1991.
NEDC-32992P-A, NEDO-32992-A........  GE Nuclear Energy, J.S.     ML14093A250.............  ML012610605
                                      Post and A.K. Chung,
                                      ``ODYSY Application for
                                      Stability Licensing
                                      Calculations,'' NEDC-
                                      32992P-A, Class III
                                      (Proprietary), July 2001,
                                      and NEDO-32992-A, Class I
                                      (Non-proprietary), July
                                      2001.
NEDC-33139P-A, NEDO-33139-A........  Global Nuclear Fuel,        ML14094A227.............  ML14094A228
                                      ``Cladding Creep
                                      Collapse,'' NEDC-33139P-
                                      A, Class III
                                      (Proprietary), July 2005,
                                      and NEDO-33139-A, Class I
                                      (Non-proprietary), July
                                      2005.
NEDE-31758P-A, NEDO-31758-A........  GE Nuclear Energy, ``GE     ML14093A142.............  ML14093A143
                                      Marathon Control Rod
                                      Assembly,'' NEDE-31758P-A
                                      (Proprietary), October
                                      1991, and NEDO-31758-A
                                      (Non-proprietary),
                                      October 1991.
NEDC-32084P-A, NEDO-32084-A........  GE Nuclear Energy, ``TASC-  ML100220484.............  ML100220485
                                      03A, A Computer Program
                                      for Transient Analysis of
                                      a Single Channel,'' NEDC-
                                      32084P-A, Revision 2,
                                      Class III (Proprietary),
                                      July 2002, and NEDO-32084-
                                      A, Class 1 (Non-
                                      proprietary), Revision 2,
                                      September 2002.
NEDC-32601 P-A, NEDO-32601-A.......  GE Nuclear Energy,          ML14093A216.............  ML003740145
                                      ``Methodology and
                                      Uncertainties for Safety
                                      Limit MCPR Evaluations,''
                                      NEDC-32601P-A, Class III
                                      (Proprietary), and NEDO-
                                      32601-A, Class I (Non-
                                      proprietary), August 1999.
NEDC-32983P-A, NEDO-32983-A........  GE Nuclear Energy, ``GE     ML072480121.............  ML072480125
                                      Methodology for Reactor
                                      Pressure Vessel Fast
                                      Neutron Flux
                                      Evaluations,'' Licensing
                                      Topical Report NEDC-
                                      32983P-A, Class III
                                      (Proprietary), Revision
                                      2, January 2006, and NEDO-
                                      32983-A, Class I (Non-
                                      proprietary), Revision 2,
                                      January 2006.
NEDC-33075P-A, NEDO-33075-A........  GE Hitachi Nuclear Energy,  ML080310396.............  ML080310402
                                      ``General Electric
                                      Boiling Water Reactor
                                      Detect and Suppress
                                      Solution--Confirmation
                                      Density,'' NEDC-33075P-A,
                                      Class III (Proprietary),
                                      and NEDO-33075-A, Class I
                                      (Non-proprietary),
                                      Revision 6, January 2008.
NEDC-33079P, NEDO-33079............  GE Nuclear Energy, ``ESBWR  ML053460471.............  ML051390233
                                      Test and Analysis Program
                                      Description,'' NEDC-
                                      33079P, Class III
                                      (Proprietary), Revision
                                      1, March 2005, and NEDO-
                                      33079, Class I (Non-
                                      proprietary), Revision 1,
                                      November 2005.

[[Page 61957]]

 
NEDC-33083P-A, NEDO-33083-A........  GE Nuclear Energy, ``TRACG  ML102770606.............  ML102770608
                                      Application for ESBWR,''
                                      NEDC-33083P-A, Revision
                                      1, Class III
                                      (Proprietary), September
                                      2010, and NEDO-33083-A,
                                      Revision 1, Class I (Non-
                                      proprietary), September
                                      2010.
NEDC-33237P-A, NEDO-33237-A........  Global Nuclear Fuel,        ML102770246.............  ML102770244
                                      ``GE14 for ESBWR--
                                      Critical Power
                                      Correlation, Uncertainty,
                                      and OLMCPR Development,''
                                      NEDC-33237P-A, Revision
                                      5, Class III
                                      (Proprietary), and NEDO-
                                      33237-A, Revision 5,
                                      Class I (Non-
                                      proprietary), September
                                      2010.
NEDC-33238P, NEDO-33238............  Global Nuclear Fuel,        ML060050328.............  ML060050330
                                      ``GE14 Pressure Drop
                                      Characteristics,'' NEDC-
                                      33238P, Class III
                                      (Proprietary), and NEDO-
                                      33238, Class I (Non-
                                      proprietary), December
                                      2005.
NEDC-33239P-A, NEDO-33239P-A.......  Global Nuclear Fuel,        ML102800405.............  ML102800408 (part 1)
                                      ``GE14 for ESBWR Nuclear                             ML102800425 (part 2)
                                      Design Report,'' NEDC-
                                      33239P-A, Class III
                                      (Proprietary), and NEDO-
                                      33239-A, Class I (Non-
                                      proprietary), Revision 5,
                                      October 2010.
NEDC-33240P-A, NEDO-33240-A........  Global Nuclear Fuel,        ML102770060.............  ML102770061
                                      ``GE14E Fuel Assembly
                                      Mechanical Design
                                      Report,'' NEDC-33240P-A,
                                      Revision 1, Class III
                                      (Proprietary), and NEDO-
                                      33240-A, Revision 1,
                                      Class I (Non-
                                      proprietary), September
                                      2010.
NEDC-33242P-A, NEDO-33242-A........  Global Nuclear Fuel,        ML102730885.............  ML102730886
                                      ``GE14 for ESBWR Fuel Rod
                                      Thermal-Mechanical Design
                                      Report,'' NEDC-33242P-A,
                                      Revision 2, Class III
                                      (Proprietary), and NEDO-
                                      33242-A, Revision 2,
                                      Class I (Non-
                                      proprietary), September
                                      2010.
NEDC-33326P-A, NEDO-33326-A........  Global Nuclear Fuel,        ML102740191.............  ML102740193 (part 1)
                                      ``GE14E for ESBWR Initial                            ML102740194 (part 2)
                                      Core Nuclear Design
                                      Report,'' NEDC-33326P-A,
                                      Revision 1, Class III
                                      (Proprietary), and NEDO-
                                      33326-A, Revision 1,
                                      Class I (Non-
                                      proprietary), September
                                      2010.
NEDC-33374P-A, NEDO-33374-A........  GE-Hitachi Nuclear Energy,  ML102860687.............  ML102860688
                                      ``Safety Analysis Report
                                      for Fuel Storage Racks
                                      Criticality Analysis for
                                      ESBWR Plants,'' NEDC-
                                      33374P-A, Revision 4,
                                      Class III (Proprietary),
                                      September 2010, and NEDO-
                                      33374-A, Revision 4,
                                      Class I (Non-
                                      proprietary), September
                                      2010.
NEDC-33456P, NEDO-33456............  Global Nuclear Fuel,        ML090920867.............  ML090920868
                                      ``Full-Scale Pressure
                                      Drop Testing for a
                                      Simulated GE14E Fuel
                                      Bundle,'' NEDC-33456P,
                                      Class III (Proprietary),
                                      and NEDO-33456, Class I
                                      (Non-proprietary),
                                      Revision 0, March 2009.
NEDE-10958-PA, NEDO-10958-A........  General Electric Company,   ML102290144.............  ML092820214
                                      ``General Electric
                                      Thermal Analysis Basis
                                      Data, Correlation and
                                      Design Application,''
                                      NEDE-10958-PA, Class III
                                      (Proprietary), and
                                      ``General Electric BWR
                                      Thermal Analysis Basis
                                      (GETAB): Data,
                                      Correlation and Design
                                      Application,'' NEDO-10958-
                                      A, Class I (Non-
                                      proprietary), January
                                      1977.
NEDE-24011-P-A-16, NEDO-24011-A-16.  Global Nuclear Fuel,        ML091340077.............  ML091340081
                                      ``GESTAR II General
                                      Electric Standard
                                      Application for Reactor
                                      Fuel,'' NEDE-24011-P-A-
                                      16, Class III
                                      (Proprietary), and NEDO-
                                      24011-A-16, Class I (Non-
                                      proprietary), Revision
                                      16, October 2007.
NEDE-24011-P-A-US-16, NEDO-24011-A-  Global Nuclear Fuel,        ML091340080.............  ML091340082
 US-16.                               ``GESTAR II General
                                      Electric Standard
                                      Application for Reactor
                                      Fuel, Supplement for
                                      United States,'' NEDE-
                                      24011-P-A-US-16, Class
                                      III (Proprietary), and
                                      NEDO-24011-A-US-16, Class
                                      I (Non-proprietary),
                                      Revision 16, October 2007.
NEDE-30130-P-A, NEDO-30130-A.......  General Electric Company,   ML14104A064.............  ML070400570
                                      ``Steady State Nuclear
                                      Methods,'' NEDE-30130-P-
                                      A, Class III
                                      (Proprietary), April
                                      1985, and NEDO-30130-A,
                                      Class I (Non-
                                      proprietary), May 1985.
NEDE-31152P, NEDO-31152............  Global Nuclear Fuel,        ML071510287.............  ML071510289
                                      ``Global Nuclear Fuels
                                      Fuel Bundle Designs,''
                                      NEDE-31152P, Revision 9,
                                      Class III (Proprietary),
                                      May 2007, and NEDO-33152,
                                      Revision 9, Class I (Non-
                                      proprietary), May 2007.
NEDE-32176P, NEDO-32176............  GE Hitachi Nuclear Energy,  ML080370271.............  ML080370276
                                      J.G.M. Andersen, et al.,
                                      ``TRACG Model
                                      Description,'' NEDE-
                                      32176P, Revision 4, Class
                                      III (Proprietary),
                                      January 2008, and NEDO-
                                      32176, Class I (Non-
                                      proprietary), Revision 4,
                                      January 2008.

[[Page 61958]]

 
NEDE-33083 Supplement 1P-A, NEDO-    GE Hitachi Nuclear Energy,  ML102770552.............  ML102770550
 33083 Supplement 1-A.                B.S. Shiralkar, et al,
                                      ``TRACG Application for
                                      ESBWR Stability
                                      Analysis,'' NEDE-33083,
                                      Supplement 1P-A, Revision
                                      2, Class III
                                      (Proprietary), September
                                      2010, and NEDO-33083,
                                      Supplement 1-A, Revision
                                      2, Class I (Non-
                                      proprietary), September
                                      2010.
NEDE-33083 Supplement 2P-A, NEDO-    GE Hitachi Nuclear Energy,  ML103000353.............  ML103000355
 33083 Supplement 2-A.                ``TRACG Application for
                                      ESBWR Anticipated
                                      Transient Without Scram
                                      Analyses,'' NEDE-33083,
                                      Supplement 2P-A, Revision
                                      2, Class III
                                      (Proprietary), October
                                      2010 and NEDO-33083,
                                      Supplement 2-A, Revision
                                      2, Class I (Non-
                                      proprietary), October
                                      2010.
NEDE-33083 Supplement 3P-A, NEDO-    GE Hitachi Nuclear Energy,  ML102770606.............  ML102770608
 33083 Supplement 3-A.                ``TRACG Application for
                                      ESBWR Transient
                                      Analysis,'' NEDE-33083,
                                      Supplement 3P-A, Revision
                                      1, Class III
                                      (Proprietary), and NEDO-
                                      33083, Supplement 3-A,
                                      Revision 1, Class I (Non-
                                      proprietary), September
                                      2010.
NEDE-33197P-A, NEDO-33197-A........  GE Hitachi Nuclear Energy,  ML102810320.............  ML102810341
                                      ``Gamma Thermometer
                                      System for LPRM
                                      Calibration and Power
                                      Shape Monitoring,'' NEDE-
                                      33197P-A, Revision 3,
                                      Class III (Proprietary),
                                      and NEDO-33197-A,
                                      Revision 3, Class I, (Non-
                                      proprietary), October
                                      2010.
NEDE-33217P, NEDO-33217............  GE Hitachi Nuclear Energy,  ML100480284.............  ML100480285
                                      ``ESBWR Man-Machine
                                      Interface System and
                                      Human Factors Engineering
                                      Implementation Plan,''
                                      NEDE-33217P, Class III
                                      (Proprietary), and NEDO-
                                      33217, Class I (Non-
                                      proprietary), Revision 6,
                                      February 2010.
NEDE-33220P, NEDO-33220............  GE Hitachi Nuclear Energy,  ML100480209.............  ML100480202
                                      ``ESBWR Human Factors
                                      Engineering Allocation of
                                      Function Implementation
                                      Plan,'' NEDE-33220P,
                                      Class III (Proprietary),
                                      and NEDO-33220, Class I
                                      (Non-proprietary),
                                      Revision 4, February 2010.
NEDE-33221P, NEDO-33221............  GE Hitachi Nuclear Energy,  ML100480212.............  ML100480213
                                      ``ESBWR Human Factors
                                      Engineering Task Analysis
                                      Implementation Plan,''
                                      NEDE-33221P, Class III
                                      (Proprietary), and NEDO-
                                      33221, Class I (Non-
                                      proprietary), Revision 4,
                                      February 2010.
NEDE-33226P, NEDO-33226............  GE Hitachi Nuclear Energy,  ML100550837.............  ML100550844
                                      ``ESBWR--Software
                                      Management Program
                                      Manual,'' NEDE-33226P,
                                      Class III (Proprietary),
                                      Revision 5, February
                                      2010, and NEDO-33226,
                                      Class I (Non-
                                      proprietary), Revision 5,
                                      February 2010.
NEDE-33243P-A, NEDO-33243-A........  GE Hitachi Nuclear Energy,  ML102740171.............  ML102740178
                                      ``ESBWR Control Rod
                                      Nuclear Design,'' NEDE-
                                      33243P-A, Revision 2,
                                      Class III (Proprietary),
                                      September 2010, and NEDO-
                                      33243-A, Revision 2,
                                      Class I (Non-
                                      proprietary), September
                                      2010.
NEDE-33244P-A, NEDO-33244-A........  GE Hitachi Nuclear Energy,  ML102770208.............  ML102770209
                                      ``ESBWR Marathon Control
                                      Rod Mechanical Design
                                      Report,'' NEDE-33244P-A,
                                      Class III (Proprietary),
                                      Revision 2, September
                                      2010, and NEDO-33244-A,
                                      Revision 2, Class I (Non-
                                      proprietary), September
                                      2010.
NEDE-33245P, NEDO-33245............  GE Hitachi Nuclear Energy,  ML100550839.............  ML100550847
                                      ``ESBWR--Software Quality
                                      Assurance Program
                                      Manual,'' NEDE-33245P,
                                      Class III (Proprietary),
                                      Revision 5, February
                                      2010, and NEDO-33245,
                                      Class I (Non-
                                      proprietary), Revision 5,
                                      February 2010.
NEDE-33259P-A, NEDO-33259-A........  GE Hitachi Nuclear Energy,  ML102920241.............  ML102920248
                                      ``Reactor Internals Flow
                                      Induced Vibration
                                      Program,'' NEDE-33259P-A,
                                      Class III (Proprietary),
                                      Revision 3, October 2010,
                                      and NEDO-33259-A, Class I
                                      (Non-proprietary),
                                      Revision 3, October 2010.
NEDE-33261P, NEDO-33261............  GE Hitachi Nuclear Energy,  ML082600720.............  ML082600721
                                      ``ESBWR Containment Load
                                      Definition,'' NEDE-
                                      33261P, Class III
                                      (Proprietary), and NEDO-
                                      33261, Class I (Non-
                                      proprietary), Revision 2,
                                      June 2008.
NEDE-33268P, NEDO-33268............  GE Hitachi Nuclear Energy,  ML100480179.............  ML100480180
                                      ``ESBWR Human Factors
                                      Engineering Human-System
                                      Interface Design
                                      Implementation Plan,''
                                      NEDE-33268P, Class III
                                      (Proprietary), and NEDO-
                                      33268, Class I (Non-
                                      proprietary), Revision 5,
                                      February 2010.

[[Page 61959]]

 
NEDE-33276P, NEDO-33276............  GE Hitachi Nuclear Energy,  ML100480182.............  ML100480183
                                      ``ESBWR Human Factors
                                      Engineering Verification
                                      and Validation
                                      Implementation Plan,''
                                      NEDE-33276P, Class III
                                      (Proprietary), and NEDO-
                                      33276, Class I (Non-
                                      proprietary), Revision 4,
                                      February 2010.
NEDE-33295P, NEDO-33295............  GE Hitachi Nuclear Energy,  ML102880103.............  ML102880104
                                      ``ESBWR Cyber Security
                                      Program Plan,'' NEDE-
                                      33295P, Class III
                                      (Proprietary), Revision
                                      2, September 2010, and
                                      NEDO-33295, Class I (Non-
                                      proprietary), Revision 2,
                                      September 2010.
NEDE-33304P, NEDO-33304............  GE Hitachi Nuclear Energy,  ML101450251.............  ML101450253
                                      ``GEH ESBWR Setpoint
                                      Methodology,'' NEDE-
                                      33304P, Class III
                                      (Proprietary), and NEDO-
                                      33304, Class I (Non-
                                      proprietary), Revision 4,
                                      May 2010.
NEDE-33312P, NEDO-33312............  GE Hitachi Nuclear Energy,  ML13344B157.............  ML13344B163
                                      ``ESBWR Steam Dryer
                                      Acoustic Load
                                      Definition,'' NEDE-
                                      33312P, Class III
                                      (Proprietary), Revision
                                      5, December 2013, and
                                      NEDO-33312, Class I (Non-
                                      proprietary), Revision 5,
                                      December 2013.
NEDE-33313P, NEDO-33313............  GE Hitachi Nuclear Energy,  ML13344B158.............  ML13344B164
                                      ``ESBWR Steam Dryer
                                      Structural Evaluation,''
                                      NEDE-33313P, Class III
                                      (Proprietary), Revision
                                      5, December 2013, and
                                      NEDO-33313, Class I (Non-
                                      proprietary), Revision 5,
                                      December 2013.
NEDE-33408P, NEDO-33408............  GE Hitachi Nuclear Energy,  ML13344B159.............  ML13344B176 (part 1)
                                      ``ESBWR Steam Dryer--                                ML13344B175 (part 2)
                                      Plant Based Load
                                      Evaluation Methodology,
                                      PBLE01 Model
                                      Description,'' NEDE-
                                      33408P, Class III
                                      (Proprietary), Revision
                                      5, December 2013, and
                                      NEDO-33408, Class I (Non-
                                      proprietary), Revision 5,
                                      December 2013.
NEDE-33440P, NEDO-33440............  GE Hitachi Nuclear Energy   ML100920316.............  ML100920317 (part 1)
                                      ``ESBWR Safety Analysis--                            ML100920318 (part 2)
                                      Additional Information,''
                                      NEDE-33440P, Class III
                                      (Proprietary), and NEDO-
                                      33440, Class I (Non-
                                      proprietary), Revision 2,
                                      March 2010.
NEDE-33516P-A, NEDO-33516-A........  GE Hitachi Nuclear Energy,  ML102880499.............  ML102880500
                                      ``ESBWR Qualification
                                      Plan Requirements for a
                                      72-Hour Duty Cycle
                                      Battery,'' NEDE-33516P-A,
                                      Revision 2, Class III
                                      (Proprietary), September
                                      2010, and NEDO-33516-A,
                                      Revision 2, Class I (Non-
                                      proprietary), September
                                      2010.
NEDE-33536P, NEDO-33536............  GE Hitachi Nuclear Energy,  ML102780329.............  ML102780330
                                      ``Control Building and
                                      Reactor Building
                                      Environmental Temperature
                                      Analysis for ESBWR,''
                                      NEDE-33536P, Class III
                                      (Security-Related and
                                      Proprietary), Revision 1,
                                      October 2010, and NEDO-
                                      33536, Class I (Non-
                                      security Related and Non-
                                      proprietary), Revision 1,
                                      October 2010.
NEDE-33572P, NEDO-33572............  GE Hitachi Nuclear Energy,  ML102740579.............  ML102740566
                                      ``ESBWR ICS and PCCS
                                      Condenser Combustible Gas
                                      Mitigation and Structural
                                      Evaluation,'' NEDE-
                                      33572P, Class II
                                      (Proprietary), Revision
                                      3, September 2010, and
                                      NEDO-33572, Revision 3,
                                      Class I (Non-
                                      proprietary), September
                                      2010.
Letter w/attachment................  Letter from R.J. Reda (GE)  ML14093A140.............  ML14094A240
                                      to R.C. Jones, Jr. (NRC),
                                      MFN 098-96,
                                      ``Implementation of
                                      Improved Steady-State
                                      Nuclear Methods,'' Class
                                      III (Proprietary), July
                                      2, 1996, and Letter from
                                      J.G. Head (GEH) to NRC
                                      Document Control Desk,
                                      MFN 098-96 Supplement 1,
                                      Class I (Non-
                                      proprietary), March 31,
                                      2014.
----------------------------------------------------------------------------------------------------------------
Table 3 Note: Documents whose document number contains ``NEDC'' or ``NEDE'' are non-public and documents whose
  document number contains ``NEDO'' are public.

C. Changes to Tier 2* Information

    The NRC is making three changes from the proposed rule regarding 
Tier 2* matters under Section VIII, ``Processes for Changes and 
Departures,'' of the ESBWR rule language. These changes are described 
below.
    First, paragraph VIII.B.6.c(1) is changed from ``ASME Boiler and 
Pressure Vessel Code, Section III'' to ``ASME Boiler and Pressure 
Vessel Code, Section III, Subsections NE (Division 1) and CC (Division 
2) for containment vessel design.'' This re-designation of Tier 2* 
information in paragraph VIII.B.6.c.(1) applies only to the ASME BPV 
Code, Section III, Subsections NE (Division 1) and CC (Division 2) for 
the design of ASME BPV Code Class MC (metal containment) and CC 
(concrete containment) pressure-retaining components (e.g., the 
containment vessel). This change does not apply to the design and 
construction of mechanical pressure-boundary components because they 
are required to meet the design and construction requirements in 
Section III for ASME BPV Code Class 1, 2, and 3 mechanical

[[Page 61960]]

pressure-boundary components, which are incorporated by reference into 
10 CFR 50.55a. The regulations in 10 CFR 50.55a include provisions in 
paragraphs 50.55a(c)(3), (d)(2) and (e)(2) for reactor coolant pressure 
boundary, Quality Group B, and Quality Group C (i.e., ASME BPV Code 
Classes 1, 2, and 3 components, respectively. These paragraphs provide 
the necessary regulatory controls on the use of later edition and 
addenda to the ASME BPV Code, Section III through the conditions the 
NRC established on the use of paragraph NCA-1140 of the ASME BPV Code, 
Section III. As a result, these rule requirements adequately control 
the ability of a licensee to use later editions or addenda of the ASME 
BPV Code, Section III such that a Tier 2* designation is not necessary.
    Second, paragraph VIII.B.6.c(3) is changed from ``Motor-operated 
valves'' to ``Power-operated valves.'' This change is necessary to 
correct an error in the proposed rule text. Consistent with Revisions 9 
and 10 of the ESBWR DCD, which were the versions of the DCD available 
for public comment, the only valves that are described in Tier 2* 
information in an ESBWR nuclear power plant are air-operated rather 
than motor-operated.
    Third, the NRC discussed in the supplemental proposed rule its 
proposal to designate the revised ESBWR steam dryer analysis 
methodology as Tier 2* information throughout the life of any license 
referencing the ESBWR DCR. This is a change from Revision 9 of the 
ESBWR DCD, which identified much of this information (in its earlier 
form before the revisions reflected in Revision 10) as Tier 2. 
Therefore, the ESBWR steam dryer analysis methodology was not 
identified as Tier 2* information in the proposed rule.
    In the supplemental proposed rule, the NRC proposed to designate 
the revised ESBWR steam dryer pressure load analysis methodology as 
Tier 2* for two reasons. First, the NRC's experience with other 
applications using this methodology highlights the importance of the 
proper application of the steam dryer pressure load analysis 
methodology. Therefore, it is necessary for the NRC to review any 
changes a referencing applicant or licensee proposes to the methodology 
from that which the NRC previously reviewed and approved. Second, in 
Revision 10 to the ESBWR DCD, GEH revised the designation of this 
methodology to Tier 2* and, therefore, the rule's designation is 
consistent with GEH's designation in the DCD.
    The supplemental proposed rule provided an opportunity for public 
comment on the proposed designation as Tier 2* of certain information 
related to the pressure load analysis methodology supporting the ESBWR 
steam dryer design. The NRC staff did not receive any public comments 
on the proposal to designate information related to the ESBWR steam 
dryer pressure load analysis methodology as Tier 2* information. 
Therefore, the final rule designates the revised ESBWR steam dryer 
pressure load analysis methodology as Tier 2* information throughout 
the life of any license referencing the ESBWR DCR.

D. Change Control for Severe Accident Design Features

    The SUPPLEMENTARY INFORMATION section of the amendment to 10 CFR 
part 52 (72 FR 49392, at 49394; August 28, 2007), states that the 
Commission codified separate criteria in paragraph B.5.c of Section 
VIII of each DCR for determining if a departure from design information 
that resolves these severe accident issues would require a license 
amendment. Originally, the final rule was applied specifically to 
changes to ex-vessel severe accident design features. In the SRM to 
SECY-12-0081, ``Risk-Informed Regulatory Framework for New Reactors,'' 
dated October 22, 2012, the Commission directed the staff to make the 
change process in paragraph B.5.c of Section VIII applicable to severe 
accident design features, both ex-vessel and non-ex-vessel, that are 
described in the plant-specific DCD. This policy was changed after 
issuance of the proposed ESBWR rule. The policy was changed to ensure 
that, for changes to Tier 2 information, the effects on all severe 
accident design features--and not just ex-vessel severe accident design 
features--are considered.
    However, the NRC has not changed the rule language in paragraph 
B.5.c of Section VIII for the ESBWR rulemaking because all of the 
relevant severe accident design features (i.e., those that are non-ex-
vessel) are described in Tier 1 information. Tier 1 information, by 
definition, includes change controls in Section VIII of the rule text 
that meet the underlying purpose of the Commission's direction. 
Therefore, this change was not necessary for the ESBWR design 
certification.

E. Access to Safeguards Information (SGI) and Sensitive Unclassified 
Non-Safeguards Information (SUNSI)

    In the four currently approved design certifications (10 CFR part 
52, appendices A through D), paragraph VI.E sets forth specific 
directions on how to obtain access to proprietary information and SGI 
on the design certification in connection with a license application 
proceeding referencing that DCR. These provisions were developed before 
the events of September 11, 2001. After September 11, 2001, Congress 
changed the statutory requirements governing access to SGI and the NRC 
has revised its rules, procedures, and practices governing control of 
and access to SGI and SUNSI. The NRC has determined that generic 
direction on obtaining access to SGI and SUNSI is no longer appropriate 
for newly approved DCRs. Accordingly, the specific requirements 
governing access to SGI and SUNSI contained in paragraph VI.E of the 
four currently approved DCRs are not included in the DCR for the ESBWR. 
Instead, the NRC will specify the procedures to be used for obtaining 
access at an appropriate time in the COL proceeding referencing the 
ESBWR DCR.

F. Human Factors Engineering (HFE) Operational Program Elements 
Exclusion From Finality

    In the December 6, 1996, SRM (ADAMS Accession No. ML003754873) to 
SECY-96-077, ``Certification of Two Evolutionary Designs,'' dated April 
15, 1996, the Commission set forth a policy that operational programs 
should be excluded from finality except where necessary to find design 
elements acceptable. For HFE programs for the ESBWR standard design, 
the Commission is implementing this policy in a manner different than 
for other existing DCRs. The difference in treatment of HFE for the 
ESBWR design arises from the level of detail of HFE review for the 
ESBWR as compared to earlier certified standard designs. For the 
earlier designs, the NRC staff reviewed the HFE programs at a 
``programmatic'' level of design, while for the ESBWR, the staff 
reviewed the HFE programs at a more detailed ``implementation plan'' 
level of design. In providing this additional detail, GEH addressed 
existing NRC guidelines in NUREG-0711, Revision 2, ``Human Factors 
Engineering Program Review Model,'' which are comprehensive and go 
beyond the operational program information needed as input to the HFE 
design. Therefore, GEH included, in the DCD, details on two HFE 
operational program elements (procedures and training) that are not 
used to determine the adequacy of the HFE design. In keeping with the 
established Commission policy of not approving operational program 
elements through design certification except where necessary to find 
design elements acceptable, the NRC is excluding these two HFE 
operational program elements

[[Page 61961]]

in the ESBWR DCD from the scope of the design approved in the rule. 
This is done explicitly in Section VI, Issue Resolution, of the ESBWR 
rule, by excluding the two HFE operational program elements from the 
issue finality and issue resolution accorded to the design. In 
addition, the procedures and training elements included in the HFE 
program are redundant to what is reviewed as part of the operational 
programs described in Chapter 13, ``Conduct of Operations,'' of the 
SRP. Accordingly, the NRC is revising the HFE regulatory guidance in 
NUREG-0711, Revision 3, ``Human Factors Engineering Program Review 
Model,'' to address this overlap, but the corresponding revision to the 
SRP has not yet been completed. This exclusion is unique to the ESBWR 
design because all other DCDs for the previously certified designs do 
not include operational program descriptions of HFE procedures and 
training and the respective DCRs did not include specific exclusions 
from finality for them.

G. Other Changes to the ESBWR Rule Language and Differences Between the 
ESBWR Rule and Other DCRs

    The language of the ESBWR design certification rule differs from 
the rule language of other DCRs in two substantive areas. First, 
paragraph IX was reserved for future use because the substantive 
requirements in this paragraph (for other DCRs) has since been 
incorporated into 10 CFR part 52 in a 2007 rulemaking (72 FR 49352; 
August 28, 2007) and thus are no longer needed in the four existing DCR 
appendices. The NRC intends to remove these requirements from Section 
IX of the four existing DCR appendices in future amendment(s) separate 
from this rulemaking.
    The second difference involves documents incorporated by reference 
into the ESBWR design certification rule. In the first four DCRs, the 
DCD is the only document identified in Section III of the rule language 
as being approved by the Office of the Federal Register for 
incorporation by reference. However, the ESBWR final rule identifies 
the ESBWR DCD and 20 publicly-available documents referenced in the 
DCD, Tier 2, Section 1.6 as approved for incorporation by reference. 
These 20 documents, which are intended by the NRC and GEH to be 
requirements, are listed in a table in Section III of the ESBWR final 
rule language. By being approved by the Office of the Federal Register 
for incorporation by reference, Revision 10 of the DCD and the 20 
publicly-available documents are considered to be requirements as if 
they had been published in the Federal Register.

IV. Technical Issues

    The NRC issued an FSER for the ESBWR design in March 2011, and 
subsequently published the FSER as NUREG-1966 in April 2014. The NRC 
issued an advanced supplemental SER in April 2014 (ADAMS Accession No. 
ML14043A134) and plans to publish Supplement No. 1 to NUREG-1966, as 
described in Section III of the SUPPLEMENTARY INFORMATION section of 
this document, before this final rule becomes effective. The FSER and 
its supplement provide the basis for issuance of a design certification 
under subpart B to 10 CFR part 52.
    The significant technical issues that were resolved during the 
initial review of the ESBWR design (i.e., the NRC staff's review of 
Revision 9 of the ESBWR DCD and development of an FSER) are: (1) 
Regulatory treatment of nonsafety systems (RTNSS), (2) containment 
performance, (3) control room cooling, (4) feedwater temperature 
operating domain, (5) steam dryer analysis methodology, (6) aircraft 
impact assessment, (7) the use of ASME Code Case N-782, and (8) an 
exemption for the safety parameter display system. These issues were 
discussed in the March 2011 proposed rule. No public comments were 
received on these issues.
    After publishing the proposed rule, the NRC addressed several 
issues that were changed in Revision 10 of the DCD or required a change 
to the FSER. The NRC staff reviewed these changes and developed an 
advanced supplemental SER as described above. The issues that were 
resolved in the advanced supplemental SER are: (1) Steam dryer analysis 
methodology, (2) loss of one or more phases of offsite power, (3) spent 
fuel assembly integrity in spent fuel racks, (4) Turbine Building 
Offgas System design requirements, (5) ASME Code statement in Chapter 1 
of the ESBWR DCD, and (6) clarification of ASME component design 
ITAACs. The NRC also made changes to the advanced supplemental SER 
after the publication of the supplemental proposed rule.
    After publication of the proposed rule, the NRC addressed two 
issues that were not addressed in Revision 10 of the DCD or in the 
advanced supplemental FSER. These issues are: (1) Hurricane-generated 
winds and missiles, and (2) changes to Tier 2* information.
    Each of these issues identified above is discussed below. The 
public was afforded an opportunity to comment on some of these issues 
in the May 6, 2014 supplemental proposed rule. Section V of the 
SUPPLEMENTARY INFORMATION section of this document describes the NRC's 
bases for not offering a supplemental comment opportunity for any of 
the other technical issues that arose after the close of the public 
comment period on the proposed rule.

A. Regulatory Treatment of Nonsafety Systems (RTNSS)

    The ESBWR safety analysis credits passive systems to perform safety 
functions for 72 hours following an initiating event. After 72 hours, 
nonsafety systems, either passive or active, replenish the passive 
systems in order to keep them operating or perform post-accident 
recovery functions directly. The ESBWR design also uses nonsafety-
related active systems to provide defense-in-depth capabilities for key 
safety functions provided by passive systems. The challenge during the 
review was to identify the nonsafety SSCs that should receive enhanced 
regulatory treatment and to identify the appropriate regulatory 
treatment to be applied to these SSCs. Such SSCs are denoted as ``RTNSS 
SSCs'' in the context of the ESBWR design. As a result of the NRC's 
review, the applicant added Appendix 19A to the DCD to identify the 
nonsafety systems that perform these post-72 hour or defense-in-depth 
functions and the basis for their selection. The applicant's selection 
process was based on the guidance in SECY-94-084, ``Policy and 
Technical Issues Associated with the Regulatory Treatment of Non-Safety 
Systems in Passive Plant Designs.''
    To provide reasonable assurance that RTNSS SSCs will be available 
if called upon to function, the applicant established availability 
controls in DCD Tier 2, Appendix 19ACM, and TS in DCD Tier 2, Chapter 
16, when required by 10 CFR 50.36, ``Technical specifications.'' The 
applicant also included all RTNSS SSCs in the reliability assurance 
program described in Chapter 17 of DCD Tier 2 and applied augmented 
design standards as described in DCD Tier 2, Section 19A.8.3. For the 
reasons set forth in Section 22.5 of the FSER, the NRC finds the 
applicant's treatment of the RTNSS SSCs, as described in the DCD, 
acceptable.

B. Containment Performance

    The PCCS maintains the containment within its design pressure and 
temperature limits for DBAs. The system is passive and does not rely 
upon moving components or external power for initiation or operation 
for 72 hours following a loss-of-coolant accident (LOCA). The PCCS and 
its

[[Page 61962]]

design basis are described in detail in Section 6.2.2 of the DCD Tier 
2. The NRC identified a concern regarding the PCCS long-term cooling 
capability for the period from 72 hours to 30 days following a LOCA. To 
address this concern, the applicant proposed additional design features 
credited after 72 hours to reduce the long-term containment pressure. 
The features are the PCCS vent fans and passive autocatalytic hydrogen 
recombiners as described in DCD Tier 2, Section 6.2.1. These SSCs have 
been identified in DCD Appendix 19A as RTNSS SSCs.
    The NRC staff's review of the PCCS design is documented in Section 
6.2.2 of the FSER. The following is a summary of key points of that 
review. The applicant provided calculation results to demonstrate that 
the long-term containment pressure would be acceptable and that the 
design complies with GDC 38. The NRC's independent calculations 
confirmed the applicant's conclusion and the NRC accepts the proposed 
design and licensing basis. The NRC also raised a concern regarding the 
potential accumulation of high concentrations of hydrogen and oxygen in 
the PCCS and Isolation Condenser System, which could lead to combustion 
following a LOCA. The applicant modified the design of the PCCS and 
Isolation Condenser System heat exchangers to withstand potential 
hydrogen detonations. Accordingly, the NRC concludes that the design 
changes to the PCCS and Isolation Condenser System are acceptable and 
meet the applicable requirements.

C. Control Room Cooling

    The ESBWR primarily relies on the mass and structure of the control 
building to maintain acceptable temperatures for human and equipment 
performance for up to 72 hours on loss of normal cooling. The NRC had 
not previously approved this approach for maintaining acceptable 
temperatures in the control building. The applicant proposed acceptance 
criteria for the evaluation of the control building structure's thermal 
performance based on industry and NRC guidelines. The applicant 
incorporated by reference an analysis of the control building 
structure's thermal performance as described in Tier 2, Sections 3H, 
6.4, and 9.4. The applicant also proposed ITAACs to confirm that an 
updated analysis of the as-built structure continues to meet the 
thermal performance acceptance criteria. For the reasons set forth in 
Section 6.4.3 of the FSER, the NRC finds that the applicant's 
acceptance criteria are consistent with the advanced light water 
reactor control room envelope atmosphere temperature limits in NUREG-
1242, ``NRC Review of Electric Power Research Institute's Advanced 
Light Water Reactor Utility Requirements Document,'' and the use of the 
wet bulb globe temperature index in evaluation of heat stress 
conditions as described in NUREG-0700, ``Human-System Interface Design 
Review Guidelines.'' For the reasons set forth in Section 9.4.1 of the 
FSER, the NRC finds the control building structure thermal performance 
analysis and ITAACs acceptable based on the analysis using bounding 
environmental assumptions. Accordingly, the NRC finds that the 
acceptance criteria, control building structure thermal performance 
analysis, and the ITAACs, provide reasonable assurance that acceptable 
temperatures will be maintained in the control building for 72 hours. 
Therefore, the NRC finds that the control building design in regard to 
thermal performance conforms to the guidelines of SRP Section 6.4 and 
complies with the requirements of the GDC 19.

D. Feedwater Temperature Operating Domain

    In operating BWRs, the recirculation pumps are used in combination 
with the control rods to control and maneuver reactor power level 
during normal power operation. The ESBWR design is unique in that the 
core is cooled by natural circulation during normal operation, and 
there are no recirculation pumps. In Chapter 15 of the DCD, GEH 
references licensing topical report (LTR) NEDO-33338, Revision 1, 
``ESBWR Feedwater Temperature Operating Domain Transient and Accident 
Analysis.'' This LTR describes a broadening of the ESBWR operating 
domain, which allows for increased flexibility of operation by 
adjusting the feedwater temperature. This increased flexibility reduces 
the duty (mechanical stress) to the fuel and minimizes the probability 
of pellet-clad interactions and associated fuel failures.
    By adjusting the feedwater temperature, the operator can control 
the reactor power level without control blade motion and with minimum 
impact on the fuel duty. Control blade maneuvering can also be 
performed at lower power levels.
    To control the feedwater temperature, the ESBWR design includes a 
seventh feedwater heater with high-pressure steam. Feedwater 
temperature is controlled by either manipulating the main steam flow to 
the No. 7 feedwater heater to increase feedwater temperature above the 
temperature normally provided by the feedwater heaters with turbine 
extraction steam (normal feedwater temperature) or by directing a 
portion of the feedwater flow around the high-pressure feedwater 
heaters to decrease feedwater temperature below the normal feedwater 
temperature. An increase in feedwater temperature decreases reactor 
power, and a decrease in feedwater temperature increases reactor power. 
As described in Section 15.1.6 of the FSER, the applicant provided 
analyses that demonstrated ample margin to acceptance criteria. For the 
reasons set forth in Section 15.1.6 of the FSER, the NRC concludes that 
the applicant has adequately accounted for the effects of the proposed 
feedwater temperature operating domain extension on the nuclear design. 
Further, the applicant has demonstrated that the fuel design limits 
will not be exceeded during normal or anticipated operational 
transients and that the effects of postulated transients and accidents 
will not impair the capability to cool the core. Based on the 
evaluation documented in Section 15.1.6 of the FSER, the NRC concludes 
that the nuclear design of the fuel assemblies, control systems, and 
reactor core will continue to meet the applicable regulatory 
requirements.

E. Steam Dryer Analysis Methodology

    As a result of RPV steam dryer issues at operating BWRs, the NRC 
issued revised guidance in Regulatory Guide (RG) 1.20, ``Comprehensive 
Vibration Assessment Program for Reactor Internals During 
Preoperational and Initial Startup Testing,'' and SRP Sections 3.9.2, 
``Dynamic Testing and Analysis of Systems, Structures, and 
Components,'' and 3.9.5, ``Reactor Pressure Vessel Internals,'' for the 
evaluation of the structural integrity of steam dryers in BWR nuclear 
power plants. The guidance requested that applicants for BWR nuclear 
power plant design certifications, licenses, or license amendments 
perform analyses to demonstrate that the steam dryer will maintain its 
structural integrity during plant operation when experiencing acoustic 
and hydrodynamic fluctuating pressure loads. This demonstration of RPV 
steam dryer structural integrity consists of three general steps:
    (1) Predict the fluctuating pressure loads on the steam dryer,
    (2) Use these fluctuating pressure loads in a structural analysis 
to demonstrate the adequacy of the steam dryer design, and
    (3) Implement a steam dryer monitoring program for confirming the 
steam dryer design analysis results during the initial plant power 
ascension testing and periodic steam dryer inspections.

[[Page 61963]]

    In its March 2011 FSER, the NRC staff described its review of the 
GEH methodology used to demonstrate the steam dryer structural 
integrity as described in Revision 9 of the ESBWR DCD and four 
referenced topical reports on which the NRC staff had issued separate 
SERs. The NRC staff concluded that the methodology was technically 
sound and provided a conservative analytical approach for definition of 
flow-induced acoustic pressure loading on the steam dryer, and that the 
design provided assurance of the structural integrity of the steam 
dryer and demonstrated conformance with GDCs 1, ``Quality Standards and 
Records,'' 2 ``Design Bases for Protection Against Natural Phenomena,'' 
and 4, ``Environmental and Dynamic Effects Design Bases.'' The NRC 
received no public comments on the proposed rule with respect to the 
steam dryer analysis methodology.
    Following the publication of the proposed rule, the NRC staff 
identified safety issues applicable to the ESBWR steam dryer structural 
analysis based on information obtained during the NRC's review of a 
license amendment request for a power uprate at an operating BWR 
nuclear power plant. Consequently, the NRC staff communicated to GEH in 
a letter dated January 19, 2012 (ADAMS Accession No. ML120170304), that 
it was concerned that the bases for its FSER on the ESBWR DCD and its 
SERs on several applicable GEH topical reports were no longer valid. 
Specifically, errors were identified in the benchmarking GEH used as a 
basis for determining fluctuating pressure loading on the steam dryer 
and errors were identified in a number of GEH's modeling parameters. 
The NRC staff subsequently issued requests for additional information 
(RAIs) and held multiple public meetings and non-public meetings (in 
which the NRC staff and GEH discussed GEH proprietary information) to 
clarify and discuss the safety issues with the ESBWR steam dryer 
analysis methodology. The NRC staff also conducted an audit of the GEH 
steam dryer analysis methodology at the GEH facility in Wilmington, 
North Carolina, in March 2012, and a vendor inspection, at that 
facility, of the quality assurance program for GEH engineering methods 
in April 2012.
    To document the resolution of those issues, GEH revised the ESBWR 
DCD by removing references to its LTRs that addressed the ESBWR steam 
dryer structural evaluation and to reference new engineering reports 
that describe the updated ESBWR steam dryer analysis methodology. The 
following four LTRs were removed by GEH (public and proprietary 
versions cited):

 NEDE-33313 and NEDE-33313P, ``ESBWR Steam Dryer Structural 
Evaluation,'' all revisions
 NEDE-33312 and NEDE-33312P, ``ESBWR Steam Dryer Acoustic Load 
Definition,'' all revisions
 NEDC-33408 and NEDC-33408P, ``ESBWR Steam Dryer--Plant Based 
Load Evaluation Methodology,'' all revisions
 NEDC-33408, Supplement 1, and NEDC-33408P, Supplement 1, 
``ESBWR Steam Dryer--Plant Based Load Evaluation Methodology Supplement 
1,'' all revisions

    To replace the information formerly provided by the four LTRs, GEH 
revised the ESBWR DCD to reference three new engineering reports 
(public and proprietary versions cited):

 NEDO-33312 and NEDE-33312P, Rev. 5, December 2013, ``ESBWR 
Steam Dryer Acoustic Load Definition''
 NEDO-33408 and NEDE-33408P, Rev. 5, December 2013, ``ESBWR 
Steam Dryer--Plant Based Load Evaluation Methodology--PBLE01 Model 
Description''
 NEDO-33313 and NEDE-33313P, Rev. 5, December 2013, ``ESBWR 
Steam Dryer Structural Evaluation''

    GEH revised the following DCD sections to correct errors and 
provide additional information related to the design and evaluation of 
the structural integrity of the ESBWR steam dryer:

 Tier 1, Chapter 2, Section 2.1, ``Nuclear Steam Supply''
 Tier 1, Chapter 2, Section 2.1.1, ``Reactor Pressure Vessel 
and Internals''
 Tier 2, Chapter 1, Tables 1.6-1, 1.9-21, and 1D-1
 Tier 2, Chapter 3, Section 3.9.2, ``Dynamic Testing and 
Analysis of Systems, Components and Equipment''
 Tier 2, Chapter 3, Section 3.9.5, ``Reactor Pressure Vessel 
Internals''
 Tier 2, Chapter 3, Section 3.9.9, ``COL Information''
 Tier 2, Chapter 3, Section 3.9.10, ``References''
 Tier 2, Chapter 3, Appendix 3L, ``Reactor Internals Flow 
Induced Vibration Program''

    The revisions to these documents enhance the detailed design and 
evaluation process related to the structural integrity of the ESBWR 
steam dryer in several ways. For example, the source of data used to 
benchmark the analysis methodology was modified in Revision 10 to the 
ESBWR DCD to a different operating nuclear power plant for which the 
NRC recently authorized an extended power uprate. In addition, the 
details of the design methodology were made more restrictive in several 
respects, including limiting the analysis methods for fillet welds and 
using more conservative data and assumptions. The changes also 
designate additional information as Tier 2* and clarify regulatory 
process steps for completing the detailed design and startup testing of 
the ESBWR steam dryer, including COL information items to be satisfied 
by a COL applicant, ITAACs to be met by a COL licensee, and model 
license conditions that may be proposed by a COL applicant.
    The NRC staff reviewed the revised ESBWR DCD sections, new GEH 
engineering reports, and RAI responses and prepared an advanced 
supplemental SER to replace Section 3.9.5, ``Reactor Pressure Vessel 
Internals,'' of the original FSER. To maintain the description of the 
regulatory evaluation of all ESBWR reactor vessel internals in the same 
location, the advanced supplemental SER replaced the entire Section 
3.9.5 in the original FSER, although only the ESBWR steam dryer 
discussion has been modified in the advanced supplemental SER in any 
significant respect. The advanced supplemental SER documents the NRC 
staff conclusion that Revision 10 to the ESBWR DCD and the referenced 
engineering reports provide sufficient information to support the 
adequacy of the design basis for the ESBWR reactor vessel internals. 
The advanced supplemental SER also documents the NRC staff conclusion 
that the design process for the ESBWR reactor vessel internals is 
acceptable and meets the requirements of 10 CFR part 50, appendix A, 
GDC 1, 2, 4, and 10; 10 CFR 50.55a; and 10 CFR part 52. Finally, the 
advanced supplemental SER documents the NRC staff conclusion that the 
ESBWR design documentation for the reactor vessel internals in Revision 
10 to the ESBWR DCD is acceptable and provides the bases for the NRC 
staff conclusion that GEH's application for the ESBWR design 
certification meets the requirements of 10 CFR part 52, subpart B, that 
are applicable and technically relevant to the ESBWR standard plant 
design. The NRC adopts the above conclusions and finds, based on the 
application materials discussed in the FSER as modified by the advanced 
supplemental SER, that the ESBWR steam dryer design meets all 
applicable NRC requirements and may be incorporated by reference in a 
COL application.

[[Page 61964]]

    The changes to the ESBWR steam dryer description in the DCD and 
supporting documentation may be regarded as significant changes which 
do not represent a ``logical outgrowth'' of the proposed rule and would 
therefore require an opportunity for public comment. To preclude any 
procedural challenges to the ESBWR final design certification rule in 
this area, the NRC staff published a supplemental proposed rule to 
provide an opportunity for public comment on these changes. The 
proposed rule and the supplemental proposed rule both provided an 
opportunity for public comment on the GEH evaluation methodology 
supporting the ESBWR steam dryer design. The NRC did not receive any 
comments on the proposed rule or the supplemental proposed rule related 
to the ESBWR steam dryer analysis methodology.
    The NRC staff briefed the Advisory Committee for Reactor Safeguards 
(ACRS) Subcommittee on the ESBWR Design Certification on March 5, 2014, 
and the ACRS Full Committee on April 10, 2014, on its detailed review 
of the ESBWR steam dryer analysis methodology, including the 
significant improvements to the GEH Plant-Based Load Evaluation 
(PBLE01) methodology for the ESBWR steam dryer to resolve the technical 
issues with the reliability of the methodology. During the ACRS 
Subcommittee briefing, the Committee suggested that the NRC staff 
change the advanced supplemental SER to clarify the description of the 
steam dryer analysis methodology. Following the Full Committee meeting, 
the ACRS provided a letter to the Commission on April 17, 2014, that 
found that the ESBWR steam dryer design is adequate, and the associated 
structural analysis and planned startup test program are acceptable. In 
its letter, the ACRS noted that, ``the process agreed to by the staff 
and GEH provides a good basis for satisfactory operation of the ESBWR 
steam dryer. In light of this reevaluation, there is reasonable 
assurance that the ESBWR design can be constructed and operated without 
undue risk to the health and safety of the public.''
    In preparing the supplemental FSER referenced in this final rule 
(Supplement No. 1 to NUREG-1966), the NRC staff modified the advanced 
supplemental SER referenced in the supplemental proposed rule to 
reflect the changes suggested during the March 5, 2014, ACRS 
subcommittee meeting. These changes include: (1) Clarifying an 
inconsistency in referring to steam flow rates, (2) clarifying the 
acceptable methods for the analysis of the stress in the fillet welds 
in the ESBWR steam dryer caused by acoustic and hydrodynamic 
fluctuating pressure loads, and for the three allowable methods 
proposed by GEH to analyze the stress in fillet welds in the ESBWR 
steam dryer, clarifying the description of (a) the test problem used by 
GEH to demonstrate the adequacy of those methods, (b) the limitations 
in the specific GEH engineering report for application of those 
methods, and (c) the results of the test problem in demonstrating the 
acceptability of each of the three fillet weld analysis methods. In 
addition, the supplemental FSER includes a new section that provides 
the conclusion of the review by the ACRS of the ESBWR steam dryer 
analysis methodology. The NRC's regulatory basis for the acceptance of 
the ESBWR steam dryer analysis methodology remains the same in the 
supplemental FSER as provided in the advanced supplemental SER 
referenced in the supplemental proposed rule. In addition, the NRC 
staff corrected a variety of typographical, grammatical, and format 
errors in the advanced supplemental SER. The NRC staff also added 
appendices to the supplemental SER, each of which correspond to and 
augment the appendices in the FSER.

F. Aircraft Impact Assessment (AIA)

    Under 10 CFR 50.150, which became effective on July 13, 2009, 
designers of new nuclear power reactors are required to perform an 
assessment of the effects on the designed facility of the impact of a 
large, commercial aircraft. An applicant for a new DCR is required to 
submit a description of the design features and functional capabilities 
identified as a result of the assessment (key design features) in its 
DCD together with a description of how the identified design features 
and functional capabilities show that the acceptance criteria in 10 CFR 
50.150(a)(1) are met.
    To address the requirements of 10 CFR 50.150, GEH completed an 
assessment of the effects on the designed facility of the impact of a 
large, commercial aircraft. GEH also added Appendix 19D to DCD Tier 2 
to describe the design features and functional capabilities of the 
ESBWR identified as a result of the assessment that ensure the reactor 
core remains cooled and the SFP integrity is maintained. These design 
features and their functional capabilities are summarized as follows:
     The isolation condenser system provides core cooling.
     The emergency core cooling system provides core cooling.
     The main steam isolation system maintains high pressure 
for core cooling with the isolation condenser system.
     The CRD system inserts control rods to shut down the 
reactor. This enables core cooling with the systems described above.
     The digital control and instrumentation system actuates 
the CRD system to shut down the reactor and enable core cooling and 
initiates the automatic depressurization system and gravity-driven 
cooling system for core cooling at low pressure.
     The reinforced concrete containment vessel protects key 
design features located inside the vessel from structural and fire 
damage.
     The location and design of the reactor building structure, 
including exterior walls, interior walls, intervening structures inside 
the building and barriers on large openings in the exterior walls 
protect the reinforced concrete containment vessel from impact.
     The location and design of the turbine building structure 
protect the adjacent wall of the reactor building from impact.
     The location and design of the fuel building structure 
protect the adjacent wall of the reactor building from impact.
     The location and design of fire barriers inside the 
reactor building protect credited core cooling equipment from fire 
damage.
     The location (below grade) and design of SFP structure 
protect the SFP from impact.
    The acceptance criteria in 10 CFR 50.150(a)(1) are: 1) the reactor 
core will remain cooled or the containment will remain intact; and 2) 
spent fuel pool cooling or spent fuel pool integrity is maintained. For 
the reasons set forth in Section 19.2.7 of the FSER, the NRC finds that 
the applicant has performed an aircraft impact assessment using an NRC-
endorsed methodology that is reasonably formulated to identify design 
features and functional capabilities to show, with reduced use of 
operator action, that the acceptance criteria in 10 CFR 50.150(a)(1) 
are met. For the same reasons, the NRC finds that the applicant 
adequately described the key design features and functional 
capabilities credited to meet 10 CFR 50.150, including descriptions of 
how the key design features and functional capabilities show that the 
acceptance criteria in 10 CFR 50.150(a)(1) are met. Therefore, the NRC 
finds that the applicant meets the applicable requirements of 10 CFR 
50.150(b).

[[Page 61965]]

G. ASME Code Case N-782

    Under 10 CFR 50.55a(a)(3), GEH requested NRC approval for the use 
of ASME Code Case N-782, ``Use of Code Editions, Addenda, and Cases 
Section III, Division 1,'' as a proposed alternative to the rules of 
Section III, Subsection NCA-1140 regarding applied Code Editions and 
Addenda required by 10 CFR 50.55a(c), (d), and (e). ASME Code Case N-
782 provides that the Code Edition and Addenda endorsed in a certified 
design or licensed by the regulatory authority may be used for systems 
and components subject to ASME Code, Section III requirements. These 
alternative requirements are in lieu of the requirements that base the 
Edition and Addenda solely on the date of an application for a 
construction permit and were issued to address new reactors licensed 
under 10 CFR part 52. Reference to ASME Code Case N-782 will be 
included in component and system design specifications and design 
reports to permit certification of these specifications and reports to 
the Code Edition and Addenda cited in the DCD. For the reasons set 
forth in Section 5.2.1.1.3 of the FSER, the NRC finds the use of ASME 
Code Case N-782 as a proposed alternative to the requirements of 
Section III, Subsection NCA-1140 under 10 CFR 50.55a(a)(3) acceptable 
for the ESBWR.

H. Exemption for the Safety Parameter Display System

    The NRC is approving an exemption from 10 CFR 50.34(f)(2)(iv) as it 
relates to the safety parameter display system. This provision requires 
an applicant to provide a plant safety parameter display console that 
will display to operators a minimum set of parameters defining the 
safety status of the plant, and is capable of displaying a full range 
of important plant parameters and data trends on demand and indicating 
when process limits are being approached or exceeded. The ESBWR design 
integrates the safety parameter display system into the design of the 
nonsafety-related distribution control and information system, rather 
than using a stand-alone console. For the reasons set forth in Section 
18.8.3.2 of the FSER, the NRC finds that the special circumstances 
described in 10 CFR 50.12(a)(2)(ii) exist in that application of 10 CFR 
50.34(f)(2)(iv) is not necessary to serve the underlying purpose of 
that rule in the context of the ESBWR design because the applicant has 
provided an acceptable alternative that accomplishes the purpose of the 
regulation. For the ESBWR, this purpose is accomplished by the plant 
alarm and display systems. In addition, the NRC finds that the proposed 
exemption is authorized by law, will not present an undue risk to 
public health and safety, and is consistent with the common defense and 
security.

I. Hurricane-Generated Winds and Missiles

    Nuclear power plants must be designed to withstand the effects of 
natural phenomena, including those that could result in the most severe 
wind events (tornadoes and hurricanes). The design bases for plant 
structures, systems, and components must reflect consideration of the 
most severe of the natural phenomena that have been historically 
reported for the site and surrounding area, with sufficient margin to 
account for the limited accuracy, quantity, and period of time in which 
the historical data have been accumulated. Initially, the U.S. Atomic 
Energy Commission, the predecessor to the NRC, considered tornadoes to 
be the bounding extreme wind events and issued RG 1.76, ``Design-Basis 
Tornado for Nuclear Power Plants,'' in April 1974, which reflected this 
technical position. RG 1.76 describes a design-basis tornado that a 
nuclear power plant should be designed to withstand without undue risk 
to the health and safety of the public. The design-basis tornado wind 
speeds were chosen so that the probability that a tornado exceeding the 
design-basis would occur was on the order of 10-\7\ per year 
per nuclear power plant.
    In March 2007, the NRC issued Revision 1 of RG 1.76. Revision 1 of 
RG 1.76 relies on the Enhanced Fujita Scale, which was implemented by 
the National Weather Service in February 2007. The Enhanced Fujita 
Scale is a revised assessment relating tornado damage to wind speed, 
which resulted in a decrease in design-basis tornado wind speed 
criteria in Revision 1 of RG 1.76, although the probability that a 
tornado would exceed this reduced wind speed remained on the order of 
10-\7\ per year per nuclear power plant. Because design-
basis tornado wind speeds were decreased as a result of the analysis 
performed to update RG 1.76, it could no longer be assumed that the 
revised tornado design-basis wind speeds would bound design-basis 
hurricane wind speeds in all areas of the U.S. This prompted the NRC to 
research extreme wind gusts during hurricanes and their relationship to 
design-basis hurricane wind speeds, which resulted in the NRC 
developing a new regulatory guide, RG 1.221, ``Design-Basis Hurricane 
and Hurricane Missiles for Nuclear Power Plants.''
    RG 1.221 evaluates missile velocities associated with several types 
of missiles considered for different hurricane wind speeds. The 
hurricane missile analyses presented in RG 1.221 are based on missile 
aerodynamic and initial condition assumptions that are similar to those 
used for the analyses of tornado-borne missile velocities adopted for 
Revision 1 to RG 1.76. However, the assumed hurricane wind field 
differs from the assumed tornado wind field in that the hurricane wind 
field does not change spatially during the missile's flight time, but 
does vary with height above the ground. Because the size of the 
hurricane zone with the highest winds is large relative to the size of 
the missile trajectory, the hurricane missile is subjected to the 
highest wind speeds throughout its trajectory. In contrast, the tornado 
wind field is smaller, so the tornado missile is subject to the 
strongest winds only at the beginning of its flight. This results in 
the same missile having a higher maximum velocity in a hurricane wind 
field than in a tornado wind field with the same maximum (3-second 
gust) wind speed.
    RG 1.221 was issued in final form in October 2011 (76 FR 63541). 
Thus, formal NRC adoption of RG 1.221 occurred after the June 7, 2011, 
close of the public comment period for the proposed ESBWR DCR, and well 
after completion of the NRC's review of the ESBWR DCD and the FSER for 
the ESBWR design in March 2011.
    Tornado loads on SSCs are addressed in Section 3.3.2 of the ESBWR 
DCD. However, Section 3.3.2 of the ESBWR DCD does not explicitly state 
whether the loads that would be experienced during a hurricane would be 
bounded under the load analysis for tornadoes. Tornado-generated 
missiles are addressed in Section 3.5.1.4 of the ESBWR DCD. Section 
3.5.1.4 of the ESBWR DCD states that ``tornado generated missiles are 
determined to be the limiting natural phenomena hazard in the design of 
all structures required for safe shutdown of the nuclear power plant. 
Because tornado missiles are used in the design basis, they envelop 
missiles generated by less intense phenomena such as extreme winds.'' 
The DCD also provides the design-basis tornado and missile spectrum in 
Tier 1, Table 5.1-1 and Tier 2, Table 2.0-1, and states its conformance 
with certain positions in RGs 1.13, 1.27, 1.76, and 1.117.
    Thus, the ESBWR applicant has not addressed, and the NRC has not 
specifically determined, whether the

[[Page 61966]]

ESBWR design is in conformance with GDCs 2 and 4 for hurricane wind and 
missile loads that are not bounded by the total tornado loads analyzed 
in the DCD. For these reasons, the NRC is only making a final safety 
determination on the acceptability of the ESBWR design with respect to 
loads on the applicable SSCs from hurricane winds and hurricane-
generated missiles that are bounded by other loads analyzed in the DCD.
    Accordingly, the NRC is excluding two issues from issue finality 
and issue resolution in the ESBWR DCD. First, with respect to the scope 
of the design in Section 3.3.2 of the ESBWR DCD, the NRC is excluding 
from finality the narrow issue of loads on applicable SSCs from 
hurricanes, but only to the extent that such loads are not bounded by 
other loads analyzed in the ESBWR DCD. Second, with respect to the 
scope of the design in Section 3.5.1.4 of the ESBWR DCD, the NRC is 
excluding from finality the narrow issue of loads on applicable SSCs 
from hurricane-generated missiles, but only to the extent that such 
loads are not bounded by other loads analyzed in the ESBWR DCD. This is 
accomplished in paragraph A.2.g of Section IV, ``Additional 
Requirements and Restrictions,'' and paragraph B.1 of Section VI, 
``Issue Resolution,'' of the new appendix E to 10 CFR part 52, by 
excluding loads from hurricane winds and hurricane-generated missiles 
on the applicable SSCs from the finality accorded to the ESBWR design 
if they are not bounded as described. Under the exclusion, a COL 
applicant referencing the ESBWR DCR must demonstrate that loads from 
site-specific hurricane winds and hurricane-generated missiles are 
bounded by the total tornado load as analyzed in the ESBWR DCD. If the 
total tornado load analyses are not bounding, the COL applicant has 
several ways of addressing the exclusion, for example, demonstrating 
that the design can withstand the hurricane wind loads and hurricane-
generated missile loads.
    The NRC's narrow exclusion with respect to issue finality, as 
reflected in the ESBWR DCR language, does not require any change to the 
ESBWR design, the ESBWR DCD, or the NRC's EA supporting the ESBWR 
rulemaking. Nor are any changes required to the associated analyses for 
total tornado loads as described in the ESBWR DCD.

J. Loss of One or More Phases of Offsite Power

    Bulletin 2012-01, ``Design Vulnerability in Electric Power 
System,'' as applied to passive plant designs such as the ESBWR, 
addresses the need for electric power system designs to be able to 
detect the loss of one or more of the three phases of an offsite power 
circuit connected to the plant electrical systems and provide an alarm 
in the control room. Bulletin 2012-01 was issued after the proposed 
rule was issued and the public comment period closed. In its response 
to Bulletin 2012-01, GEH provided additional details on the monitoring 
and alarm functions for all three phases of the offsite power circuits 
and included applicable information in Revision 10 to the DCD. GEH also 
added new ITAACs to ensure implementation of these design features by a 
COL holder. The NRC staff reviewed the ESBWR design features that can 
detect and provide an alarm for the loss of one or more of the three 
phases of an offsite power circuit. For the reasons set forth in 
Section 8.2.3, ``Staff Evaluation,'' of the supplemental FSER, the NRC 
concludes that no design vulnerability identified in Bulletin 2012-01 
exists in the ESBWR electric power system.

K. Spent Fuel Assembly Integrity in Spent Fuel Racks

    Prior to publishing the proposed rule, the NRC performed its review 
of the integrity of spent fuel racks based on SRP Section 9.1.2, ``New 
and Spent Fuel Storage.'' This section states that ``Designing the 
storage pool and fuel storage racks to meet seismic Category I 
requirements provides reasonable assurance that earthquakes will not 
cause a substantial coolant loss, a reduction in margin to criticality, 
or damage to the fuel assemblies.'' This section supports the NRC's 
requirements in GDC 2, which requires that nuclear power plant SSCs 
important to safety be designed to withstand the effects of natural 
phenomena, such as an earthquake without loss of capability to perform 
their safety functions. The ESBWR FSER concluded that the design of the 
SFP, the buffer pool, and the fuel storage racks complied with the 
requirements of GDC 2 and met the guidance of SRP Section 9.1.2.
    After publication of the proposed rule, the NRC recognized that 
Appendix D, ``Guidance on Spent Fuel Racks,'' to SRP Section 3.8.4, 
``Other Seismic Category I Structures,'' states that, ``It should be 
demonstrated that the consequent loads on the fuel assembly do not lead 
to damage of the fuel.'' In other words, though the spent fuel rack may 
have remained intact during a seismic event, because there are gaps 
between the rack and the fuel assemblies, the applicant should 
demonstrate that the spent fuel assemblies in the rack have not 
sustained damage during that seismic event. During the NRC staff's 
review of the ESBWR design and prior to its publication of its FSER, 
the NRC staff did not specifically review the design of the spent fuel 
in the spent fuel racks against this guidance, but only against that of 
SRP Section 9.1.2 as described above.
    To confirm the structural integrity of the fuel in the spent fuel 
racks, the NRC staff conducted an audit on August 5 and September 8, 
2011. The audit summary is available under ADAMS Accession No. 
ML112860614. GEH subsequently submitted additional information (ADAMS 
Accession No. ML11269A093) to address whether the consequent loads on 
the fuel assembly that result from the design-basis seismic event would 
lead to fuel damage. For the reasons set forth in Section 3.8.4 of the 
supplemental FSER, the NRC finds that the fuel assemblies maintain 
structural integrity when subject to the design-basis seismic loads, 
the fuel assemblies in the fuel storage racks are structurally adequate 
to withstand the design-basis seismic loads, and the fuel assemblies 
are in compliance with GDC 2.

L. Turbine Building Offgas System Design Requirements

    Regulatory Guide (RG) 1.143, ``Design Guidance for Radioactive 
Waste Management Systems, Structures, and Components Installed in 
Light-Water-Cooled Nuclear Power Plants,'' provides guidance on 
classifying and designing radioactive waste management systems (RWMSs). 
The Offgas System (OGS), which is part of the Gaseous Waste Management 
System, is classified as a Category RW-IIa (High Hazard) RWMS in 
accordance with RG 1.143. Following publication of the proposed rule, 
the NRC staff identified that while it had evaluated the OGS against 
the guidelines of RG 1.143, the NRC staff had not evaluated the 
structure housing the OGS (i.e., the turbine building), against the 
guidelines of RG 1.143. Subsequently, the NRC staff reviewed the 
information included in various sections of the ESBWR DCD regarding 
protection of the OGS. For the reasons set forth in Section 3.8.4.3 of 
the supplemental FSER, the NRC finds that the turbine building 
structure provides adequate protection for the OGS components to meet 
the design criteria in RG 1.143 for Category RW-IIa.
    Because the NRC staff's evaluation of the turbine building 
structure came after completion of the FSER, issuance of the final SDA, 
and publication of the proposed rule, the NRC decided to

[[Page 61967]]

document the NRC staff's review on this issue in the supplemental FSER. 
The evaluation was performed using information already included in 
Revision 9 of the ESBWR DCD and that information did not change in 
Revision 10 of the DCD. Further, the NRC determined that no changes 
were required to the ESBWR DCD, the proposed rule text, or the EA 
supporting this rulemaking.

M. ASME BPV Code Statement in Chapter 1 of the ESBWR DCD

    In Revision 10 to the ESBWR DCD, Tier 1, Section 1.1.1, 
``Definitions,'' the applicant added a definition of ``ASME Code'' to 
its Tier 1 definitions. This addition addressed compliance with the 
ASME BPV Code and the use of alternatives to the ASME BPV Code 
requirements as permitted in 10 CFR 50.55a(a)(3). For the ESBWR DCR, 
several ITAACs in the ESBWR Tier 1 are required to verify that ASME BPV 
Code, Section III construction requirements have been met. During 
actual construction of a nuclear power plant, it is inevitable that 
departures from the ASME BPV Code construction requirements will be 
needed. These departures occur for various reasons such as 
unavailability of material, hardship in implementing fabrication 
sequences required by the Code, and the availability of newer and more 
effective construction techniques. As such, the regulations in 10 CFR 
50.55a, ``Codes and standards,'' provide for the use of alternatives to 
Section III construction requirements to overcome such hardships and 
allow a degree of flexibility in constructing nuclear power plants 
without compromising safety requirements. Pursuant to 10 CFR 
50.55a(a)(3), proposed alternatives to Section III requirements may be 
used when authorized by the NRC. Before using these alternatives, the 
applicant or licensee must demonstrate that: (1) the proposed 
alternative would provide an acceptable level of quality and safety, or 
(2) compliance with the specified requirements of 10 CFR 50.55a would 
result in hardship or unusual difficulty without a compensating 
increase in the level of quality and safety.
    During the construction of two nuclear power plants licensed under 
10 CFR part 52 (Vogtle Electric Generating Plant, Units 3 and 4, and 
V.C. Summer Nuclear Station, Units 2 and 3), the question arose whether 
changes to ASME BPV Code requirements, such as the use of alternatives 
in accordance with 10 CFR 50.55a(a)(3), are permitted without the need 
to submit an exemption from the regulations pursuant to 10 CFR 50.12, 
``Specific exemptions.'' The NRC staff found that this issue was 
previously discussed in the SUPPLEMENTARY INFORMATION section of a 
final rule dated August 28, 2007, amending the regulations to address 
10 CFR part 52 requirements (72 FR 49352). Therein, the NRC stated in 
Section VI, ``Section-by-Section Analysis,'' for Section 52.7, 
``Specific Exemptions,'' (at 72 FR 49438) that, ``Sec.  52.7 does not 
supersede the applicability of more specific dispensation provisions in 
other parts of Chapter I. For example, a holder of a COL would not 
require a separate part 52 exemption in order to obtain approval of an 
alternative to a provision of an applicable ASME Code provision that is 
otherwise required under 10 CFR 50.55a; the licensee need only satisfy 
the criteria in Sec.  50.55a(a)(3) . . .'' The 2007 10 CFR part 52 
final rule SUPPLEMENTARY INFORMATION clarified that using alternatives 
to ASME Code requirements authorized in accordance with 10 CFR 50.55a 
is sufficient and does not require a COL holder to submit an exemption 
when changes involve a departure from only ASME Code requirements.
    To clarify the use of alternatives when verifying compliance with 
ASME BPV Code ITAACs, GEH proposed to clarify in its Tier 1 definitions 
in Revision 10 to the ESBWR DCD, Section 1.1.1, ``Definitions,'' that 
``ASME Code'' means ASME BPV Code requirements or any alternative 
authorized by the NRC pursuant to 10 CFR 50.55a(a)(3). This change does 
not affect previous NRC safety findings in the FSER or change the 
status of how the ESBWR standard design complies with ASME BPV Code 
requirements. For the reasons set forth in Section 14.3 of the 
supplemental FSER, the NRC finds that these changes to the definition 
of ASME Code are acceptable.

N. Clarification of ASME Component Design ITAACs

    Following the publication of the proposed rule, the NRC staff 
reviewed ITAACs for inspectability and consistency across several 
design certifications. This review identified the potential issue that 
the ITAACs related to verification of component design, as written in 
Revision 9 of the ESBWR DCD, might be viewed as requiring design 
verification of as-designed ASME BPV Code components, rather than as-
built ASME BPV Code components, as originally intended. Verifying 
interim ASME BPV Code design reports at the design stage would result 
in an unnecessary regulatory burden with no benefit to safety. In 
Revision 10 of the ESBWR DCD, the ASME BPV Code component ITAACs were 
revised to clarify that the activities needed to satisfy the ITAACs are 
performed at the as-built stage. For the reasons set forth in Section 
14.3.3 of the supplemental FSER, the NRC concludes that this 
clarification promotes efficient ITAAC closure and reduces potential 
confusion while having no effect on previous NRC safety findings.

O. Corrections, Editorial, and Conforming Changes

    GEH made corrections and editorial changes in Revision 10 of the 
DCD. The NRC corrected typographical errors, made other editorial 
changes, and added units of measurements to the advanced supplemental 
SER. The NRC also revised the advanced supplemental SER after 
publication of the supplemental proposed rule to include conforming 
changes such as adding appendices that augment the appendices in the 
FSER.

V. Rulemaking Procedure

A. Exclusions From Issue Finality and Issue Resolution for Spent Fuel 
Pool Instrumentation

    As described in Section III of the SUPPLEMENTARY INFORMATION 
section of this document related to how the ESBWR design addresses 
Fukushima NTTF recommendations, the NRC is changing the ESBWR DCR 
language to exclude from finality the safety-related SFP level 
instruments: (1) Being designed to allow the connection of an 
independent power source, and (2) maintaining its design accuracy 
following a power interruption or change in power source without 
recalibration. There was no change to the ESBWR design, as described in 
the DCD, the NRC's EA supporting the ESBWR rulemaking (and in 
particular, the SAMDA analysis), or the ESBWR FSER. In addition, the 
final rule is more conservative than the proposed rule because it is 
more limiting both as to what is certified and to the scope of issue 
finality. The NRC is not aware of any entity other than the applicant, 
GEH, who would be adversely affected by this change. With respect to 
the exclusions, GEH voluntarily declined to submit additional 
information that would avoid the need for exclusions from issue 
finality and issue resolution on this matter. The NRC did not receive 
any public comments in the area of spent fuel pool instrumentation 
(which otherwise would suggest public interest in this matter). For 
these reasons, the NRC staff concluded that a supplemental opportunity 
for public comment was not warranted for these

[[Page 61968]]

exclusions from issue finality and issue resolution.

B. Incorporation by Reference of Public Documents

    The change to the ESBWR DCR language related to approval for 
incorporation by reference by the Office of the Federal Register of 20 
publicly-available documents is described in Section III of the 
SUPPLEMENTARY INFORMATION section of this document. The supplemental 
proposed rule discussed the changes to the ESBWR DCR language but 
deferred the discussion of why a public comment opportunity was not 
provided to the final rule. The NRC did not offer a supplemental 
opportunity for public comment on this matter for the following 
reasons. First, the text of the DCD--when discussing each of the 20 
publicly-available documents--makes clear that these are intended to be 
requirements. Thus, a member of the public could have discerned and 
commented on the failure of Tables 1.6-1 and 1.6-2 of the Revision 9 of 
the DCD to differentiate between documents intended to be requirements 
(given the information presented throughout DCD Revision 9) and 
documents which were intended only to be references (i.e., ``for 
information only''). The public could also have commented on the 
discrepancy between the language of Revision 9 of the DCD (which 
regards these documents as being incorporated by reference into the 
DCD) and the failure of the proposed ESBWR design certification rule to 
list the publicly-available referenced documents as being approved by 
the Office of the Federal Register for incorporation by reference. 
Finally, the NRC did not receive any comments on the proposed rule with 
respect to Tables 1.6-1 and 1.6-2 in Revision 9 of the DCD, or the 
incorporation by reference language in Section III of proposed Appendix 
E to part 52 (which otherwise would suggest public interest in this 
matter). For these reasons, the NRC staff concluded that a supplemental 
opportunity for public comment was not warranted with respect to the 
status of the 20 documents as requirements and their incorporation by 
reference into the ESBWR design certification rule.

C. Changes to Tier 2* Information

    The final rule includes three changes from the proposed rule 
regarding Tier 2* matters under Section VIII of the ESBWR rule language 
as described in Section III of the SUPPLEMENTARY INFORMATION section of 
this document. Because one of those changes was related to the steam 
dryer, and for the same reasons as the steam dryer analysis methodology 
being offered a supplemental opportunity for public comment, the 
related Tier 2* change was included in the supplemental proposed rule 
and no public comments were received on this topic. The other two Tier 
2* changes--related to the specific subsections of ASME BPV Code and a 
correction to the type of valves used in the ESBWR design--were 
included for consistency with the ESBWR design as described in the DCD. 
First, paragraph VIII.B.6.c.(1) is changed from ``ASME Boiler and 
Pressure Vessel Code, Section III'' to ``ASME Boiler and Pressure 
Vessel Code, Section III, Subsections NE (Division 1) and CC (Division 
2) for containment vessel design.'' The NRC determined that no changes 
were required to the ESBWR design or the DCD; rather, the change to the 
rule text is needed to make the rule consistent with Revisions 9 and 10 
of the ESBWR DCD. Further, the change represents a restriction as 
compared to the proposed rule language. That is, the proposed rule 
would allow the larger scope of Tier 2* information with respect to 
ASME BPV Code, Section III to revert to Tier 2 after full power, 
whereas the change to the final rule does not allow containment vessel 
design information subject to Subsection NE., Division 1, and 
Subsection CC, Division 2, to revert to Tier 2 after the plant first 
achieves full power following the finding required by 10 CFR 52.103(g). 
Therefore, the NRC concludes that a supplemental opportunity for public 
comment on these changes to the rule is not warranted.
    Second, paragraph VIII.B.6.c.(3) is changed from ``Motor-operated 
valves'' to ``Power-operated valves.'' The NRC determined that no 
changes were required to the ESBWR design or the DCD; rather, the 
change to the rule text is needed to make the rule consistent with 
Revisions 9 and 10 of the ESBWR DCD. Further, the change to the rule 
text is corrective in nature and does not represent a substantive 
change to the nature of Tier 2* matters. Therefore, the NRC concludes 
that a supplemental opportunity for public comment on these changes to 
the rule is not warranted.

D. Other Changes to the ESBWR Rule Language and Difference From Other 
DCRs

    The ESBWR final rule language differs from the proposed rule 
language in several areas that are administrative or clarifying and do 
not involve any substantive change. Those differences, and the 
rationale for the differences, are as follows. Paragraph III.A, which 
describes the document being incorporated by reference and how to 
examine or obtain copies of that document, was revised to conform to 
other recently issued DCRs and to the Office of the Federal Register's 
guidance. Paragraphs III.D and V.A were revised to include the NUREG 
number for the FSER; the NUREG was not available when the NRC published 
the ESBWR proposed rule. Paragraphs IV.A.3, VI.E, and X.A.1 were 
administratively revised to remove acronyms for SUNSI and SGI but 
retain the terms that these acronyms represent for consistency with 
other DCRs. For paragraph VI.E, footnoted text was moved into the body 
of the regulation where these terms were noted. Paragraph V.B.1 was 
revised to clarify that, similar to the regulations that apply to the 
ESBWR design in Paragraph V.A, the regulations that the ESBWR design is 
exempt from are those codified as of the date the final rule is signed 
by the Secretary of the Commission. Because these changes are 
administrative in nature, the NRC concluded that a supplemental 
opportunity for public comment was not warranted for these matters.
    ESBWR final rule language differs from the rule language of other 
DCRs in several areas that are not otherwise explained in the preceding 
paragraph. Those differences, and the rationale for the differences, 
are as follows. Paragraph II.B was administratively revised to include 
the term ``generic TS,'' similar to that of ``generic DCD'' in 
Paragraph II.A, as it is used in appendix E. Paragraph II.C was revised 
to clarify the actual content of a plant-specific DCD. Paragraph 
IV.A.2.a was revised to provide flexibility to COL applicants by 
updating the process by which a COL applicant can reference information 
in the generic DCD--either by including that information or 
incorporating it by reference; current DCRs are silent as to how to 
include this information. Paragraphs IV.A.2.d and VI.B.7 were revised 
to conform to other NRC regulations regarding site characteristics for 
a COL, postulated site parameters for a certified design, and the 
interface requirements. Finally, paragraph IX was reserved for future 
use because the substantive requirements in this paragraph (for other 
DCRs) has since been incorporated into 10 CFR part 52 in a 2007 
rulemaking (72 FR 49352; August 28, 2007) and thus are no longer needed 
in the four existing DCR appendices. The NRC intends to remove these 
requirements from Section IX of the four existing DCR appendices in

[[Page 61969]]

future amendment(s) separate from this rulemaking. Because these are 
administrative in nature, the NRC concluded that a supplemental 
opportunity for public comment was not warranted for these matters.

E. Exclusions From Issue Finality and Issue Resolution for Hurricane-
Generated Winds and Missiles

    As described in Section IV of the SUPPLEMENTARY INFORMATION section 
of this document, the final rule contains exclusions from issue 
finality and issue resolution related to hurricane-generated winds and 
missiles. The ESBWR design, as described in the DCD, the NRC's EA 
supporting the ESBWR rulemaking (and in particular, the SAMDA 
analysis), and the ESBWR FSER did not change. In addition, the change 
to the final rule is more conservative than the proposed rule because 
it is more limiting as to what is certified and the scope of issue 
finality. The NRC is not aware of any entity other than the applicant, 
GEH, who would be adversely affected by this change. With respect to 
the exclusions, GEH voluntarily declined to submit additional 
information which would avoid the need for exclusions from issue 
finality and issue resolution on this matter. The NRC did not receive 
any public comments on hurricane winds or hurricane missiles (which 
otherwise would suggest public interest in this matter). For these 
reasons, the NRC staff concluded that a supplemental opportunity for 
public comment was not warranted for these exclusions from issue 
finality and issue resolution.

F. Loss of One or More Phases of Offsite Power

    The changes that GEH made to the DCD and the NRC staff conclusions 
in its supplemental FSER to clarify how the ESBWR design addresses the 
loss of one or more phases of offsite power in order to demonstrate 
compliance with GDC 17, ``Electric Power Systems,'' are described in 
Section IV of the SUPPLEMENTARY INFORMATION section of this document. 
These changes did not require a change to the rule text or to the EA 
supporting this rulemaking. The NRC did not receive any public comments 
on the proposed rule with respect to the adequacy of the offsite power 
system (which would otherwise suggest public interest in this matter). 
For these reasons, the NRC staff concluded that a supplemental 
opportunity for public comment was not warranted for this matter.

G. Spent Fuel Assembly Integrity in Spent Fuel Racks

    The discussion in the supplemental FSER related to spent fuel 
assembly integrity in spent fuel racks is described in Section IV of 
the SUPPLEMENTARY INFORMATION section of this document. The NRC staff 
determined that the additional information provided by GEH did not 
require a change to the design of the fuel or the spent fuel racks as 
described in Revision 9 of the ESBWR DCD or new design commitments in 
the DCD. No changes were required to the ESBWR DCD, the rule text, or 
the EA supporting this rulemaking. The NRC did not receive any public 
comments on the proposed rule with respect to spent fuel pool assembly 
integrity (which otherwise would suggest public interest in this 
matter). For these reasons, the NRC staff concluded that a supplemental 
opportunity for public comment was not warranted for this matter, 
including the supplemental FSER.

H. Turbine Building Offgas System Design Requirements

    The NRC staff's evaluation of the turbine building structure 
relative to the Turbine Building Offgas System design requirements, as 
documented in a supplemental FSER, is described in Section IV of the 
SUPPLEMENTARY INFORMATION section of this document. The staff's 
evaluation, which was not documented in the March 2011 FSER, was 
performed using information in Revision 9 of the ESBWR DCD that did not 
change in Revision 10 of the DCD. Further, there were no changes 
required to the ESBWR DCD, the rule text, or the EA supporting this 
rulemaking. The NRC did not receive any public comments on the proposed 
rule with respect to the Turbine Building Offgas System (which 
otherwise would suggest public interest in this matter). For these 
reasons, the NRC staff concluded that a supplemental opportunity for 
public comment was not warranted for this matter.

I. ASME BPV Code Statement in Chapter 1 of the ESBWR DCD

    The technical clarification to the DCD and supplemental FSER 
related to the ASME BPV Code statement in Chapter 1 of the ESBWR DCD is 
described in Section IV of the SUPPLEMENTARY INFORMATION section of 
this document. This clarification does not affect previous NRC safety 
findings in the FSER, change the ESBWR's compliance with Code 
requirements, or require changes to the rule text for this rulemaking. 
For these reasons, the NRC staff concluded that a supplemental 
opportunity for public comment was not warranted for this matter.

J. Clarification of ASME Component Design ITAACs

    The technical clarifications that GEH made to the DCD and the 
staff's conclusions in its supplemental FSER regarding the ASME 
component design ITAACs are described in Section IV of the 
SUPPLEMENTARY INFORMATION section of this document. This clarification 
does not affect previous NRC safety findings in the FSER, nor does it 
require changes to the rule text for this rulemaking. For these 
reasons, the NRC staff concluded that a supplemental opportunity for 
public comment was not warranted for this matter.

K. Changes to the Supplemental FSER After Publication of the 
Supplemental Proposed Rule

    The advanced supplemental SER was issued on April 17, 2014 (ADAMS 
Accession No. ML14043A134). After the supplemental proposed rule was 
issued, and to reflect the changes suggested during the March 5, 2014, 
ACRS subcommittee meeting, the NRC revised the advanced supplemental 
SER and prepared it as a supplement to the FSER. In this revision the 
NRC clarified the discussion of the ESBWR steam dryer analysis 
methodology regarding Methods 1, 2, and 3 in Section 3.9.5.3.3.5.2.3. 
In addition, the supplemental FSER includes a new section that provides 
the conclusion of the review by the ACRS of the ESBWR steam dryer 
analysis methodology. The NRC staff's regulatory basis for the 
acceptance of the ESBWR steam dryer analysis methodology remains the 
same in the supplemental FSER as provided in the advanced supplemental 
SER referenced in the supplemental proposed rule. For this reason, the 
NRC staff concluded that a supplemental opportunity for public comment 
was not warranted for this matter. The supplemental FSER (ADAMS 
Accession No. ML14155A333) will be published as Supplement No. 1 to 
NUREG 1966. NUREG-1966 was published in April 2014 (ADAMS Accession No. 
ML14100A304).

L. Corrections, Editorial, and Conforming Changes

    GEH made editorial changes in Revision 10 of the DCD. The NRC 
corrected typographical errors, made other editorial changes, and added 
units of measurements to the advanced supplemental SER. The NRC staff 
also revised the advanced supplemental SER after publication of the 
supplemental

[[Page 61970]]

proposed rule to include conforming changes such as adding appendices 
that augment the appendices in the FSER. Because these changes are 
administrative in nature, the NRC staff concluded that a supplemental 
opportunity for public comment was not warranted for these matters.

VI. Planned Withdrawal of the ESBWR SDA

    In its application (ADAMS Accession No. ML052450245), GEH requested 
the NRC provide its design approval for the ESBWR design. The SDA for 
the ESBWR design was issued in March 2011 (ADAMS Accession No. 
ML110540310) after the completion of the FSER. In a letter dated June 
3, 2014 (ADAMS Accession No. ML14154A094), GEH requested that the NRC 
retire the SDA at the time of issuance of the final ESBWR DCR. In 
accordance with GEH's request, the NRC plans to issue a Federal 
Register notice announcing the withdrawal of the ESBWR SDA after the 
effective date of the final ESBWR design certification rule.

VII. Section-by-Section Analysis

    The following discussion sets forth the purpose and key aspects of 
each section and paragraph of the final ESBWR DCR. All section and 
paragraph references are to the provisions in appendix E to 10 CFR part 
52 unless otherwise noted. The NRC has modeled the ESBWR DCR on the 
existing DCRs, with certain modifications where necessary to account 
for differences in the ESBWR design documentation, design features, and 
EA (including SAMDAs). As a result, the DCRs are standardized to the 
extent practical.

A. Introduction (Section I)

    The purpose of Section I of appendix E to 10 CFR part 52 (this 
appendix) is to identify the standard plant design that would be 
approved by this DCR and the applicant for certification of the 
standard design. Identification of the design certification applicant 
is necessary to implement this appendix for two reasons. First, the 
implementation of 10 CFR 52.63(c) depends on whether an applicant for a 
COL contracts with the design certification applicant to provide the 
generic DCD and supporting design information. If the COL applicant 
does not use the design certification applicant to provide the design 
information and instead uses an alternate nuclear plant vendor, then 
the COL applicant must meet the requirements in 10 CFR 52.73. The COL 
applicant must demonstrate that the alternate supplier is qualified to 
provide the standard plant design information. Second, paragraph X.A.1 
requires the design certification applicant to maintain the generic DCD 
throughout the time this appendix may be referenced. Thus, it is 
necessary to identify the entity to which the requirement in paragraph 
X.A.1 applies.

B. Definitions (Section II)

    During development of the first two DCRs, the NRC decided that 
there would be both generic (master) DCDs maintained by the NRC and the 
design certification applicant, as well as individual plant-specific 
DCDs maintained by each applicant and licensee that reference this 
appendix. This distinction is necessary in order to specify the 
relevant plant-specific requirements to applicants and licensees 
referencing the appendix. In order to facilitate the maintenance of the 
master DCDs, the NRC requires that each application for a standard 
design certification be updated to include an electronic copy of the 
final version of the DCD. The final version is required to incorporate 
all amendments to the DCD submitted since the original application, as 
well as any changes directed by the NRC as a result of its review of 
the original DCD or as a result of public comments. This final version 
is the master DCD incorporated by reference in the DCR. The master DCD 
would be revised as needed to include generic changes to the version of 
the DCD approved in this design certification rulemaking. These changes 
would occur as the result of generic rulemaking by the Commission, 
under the change criteria in Section VIII.
    The NRC also requires each applicant and licensee referencing this 
appendix to submit and maintain a plant-specific DCD as part of the COL 
FSAR. This plant-specific DCD must either include or incorporate by 
reference the information in the generic DCD. The plant-specific DCD 
would be updated as necessary to reflect the generic changes to the DCD 
that the Commission may adopt through rulemaking, plant-specific 
departures from the generic DCD that the Commission imposed on the 
licensee by order, and any plant-specific departures that the licensee 
chooses to make in accordance with the relevant processes in Section 
VIII. Thus, the plant-specific DCD functions like an updated FSAR 
because it would provide the most complete and accurate information on 
a plant's design-basis for that part of the plant within the scope of 
this appendix. Therefore, this appendix defines both a generic DCD and 
a plant-specific DCD.
    Also, the NRC is treating the TS in Chapter 16 of the generic DCD 
as a special category of information and designating them as generic TS 
in order to facilitate the special treatment of this information under 
this appendix. A COL applicant must submit plant-specific TS that 
consist of the generic TS, which may be modified under paragraph 
VIII.C, and the remaining plant-specific information needed to complete 
the TS. The FSAR that is required by 10 CFR 52.79 will consist of the 
plant-specific DCD, the site-specific portion of the FSAR, and the 
plant-specific TS.
    The terms Tier 1, Tier 2, Tier 2*, and COL action items (license 
information) are defined in this appendix because these concepts were 
not envisioned when 10 CFR part 52 was developed. The design 
certification applicants and the NRC used these terms in implementing 
the two-tiered rule structure that was proposed by representatives of 
the nuclear industry after issuance of 10 CFR part 52. Therefore, 
appropriate definitions for these additional terms are included in this 
appendix. The nuclear industry representatives requested a two-tiered 
structure for the DCRs to achieve issue preclusion for a greater amount 
of information than was originally planned for the DCRs, while 
retaining flexibility for design implementation. The Commission 
approved the use of a two-tiered rule structure in its SRM, dated 
February 14, 1991, on SECY-90-377, ``Requirements for Design 
Certification under 10 CFR Part 52,'' dated November 8, 1990. This 
document and others are available in the Regulatory History of Design 
Certification (see Section VII of this document).
    The Tier 1 portion of the design-related information contained in 
the DCD is certified by this appendix and, therefore, subject to the 
special backfit provisions in paragraph VIII.A. An applicant who 
references this appendix is required to include or incorporate by 
reference and comply with Tier 1, under paragraphs III.B and IV.A.1. 
This information consists of an introduction to Tier 1, the system 
based and non-system based design descriptions and corresponding 
ITAACs, significant interface requirements, and significant site 
parameters for the design (refer to Section C.I.1.8 of RG 1.206 for 
guidance on significant interface requirements and site parameters). 
The design descriptions, interface requirements, and site parameters in 
Tier 1 were derived from Tier 2, but may be more general than the Tier 
2 information. The NRC staff's evaluation of the Tier 1 information is 
provided in Section 14.3

[[Page 61971]]

of the FSER. Changes to or departures from the Tier 1 information must 
comply with Section VIII.A.
    The Tier 1 design descriptions serve as requirements for the 
lifetime of a facility license referencing the design certification. 
The ITAACs verify that the as-built facility conforms to the approved 
design and applicable regulations. Under 10 CFR 52.103(g), the 
Commission must find that the acceptance criteria in the ITAACs are met 
before authorizing operation. After the Commission has made the finding 
required by 10 CFR 52.103(g), the ITAACs do not constitute regulatory 
requirements for licensees or for renewal of the COL. However, 
subsequent modifications to the facility within the scope of the design 
certification must comply with the design descriptions in the plant-
specific DCD unless changes are made under the change process in 
Section VIII. The Tier 1 interface requirements are the most 
significant of the interface requirements for systems that are wholly 
or partially outside the scope of the standard design. Tier 1 interface 
requirements must be met by the site-specific design features of a 
facility that references this appendix. An application that references 
this appendix must demonstrate that the site characteristics at the 
proposed site fall within the site parameters (both Tier 1 and Tier 2) 
(refer to paragraph V.D of this document).
    Tier 2 is the portion of the design-related information contained 
in the DCD that is approved by this appendix but not certified. Tier 2 
information is subject to the backfit provisions in paragraph VIII.B. 
Tier 2 includes the information required by 10 CFR 52.47(a) and 
52.47(c) (with the exception of generic TS and conceptual design 
information) and the supporting information on inspections, tests, and 
analyses that will be performed to demonstrate that the acceptance 
criteria in the ITAACs have been met. As with Tier 1, paragraphs III.B 
and IV.A.1 require an applicant who references this appendix to include 
or incorporate by reference Tier 2 and to comply with Tier 2, except 
for the COL action items, including the availability controls in 
Appendix 19ACM of the generic DCD. The definition of Tier 2 makes clear 
that Tier 2 information has been determined by the NRC, by virtue of 
its inclusion in this appendix and its designation as Tier 2 
information, to be an approved sufficient method for meeting Tier 1 
requirements. However, there may be other acceptable ways of complying 
with Tier 1 requirements. The appropriate criteria for departing from 
Tier 2 information are specified in paragraph VIII.B. Departures from 
Tier 2 information do not negate the requirement in paragraph III.B to 
incorporate by reference Tier 2 information.
    A definition of ``combined license action items'' (COL 
information), which is part of the Tier 2 information, has been added 
to clarify that COL applicants who reference this appendix are required 
to address COL action items in their license application. However, the 
COL action items are not the only acceptable set of information. An 
applicant may depart from or omit COL action items, provided that the 
departure or omission is identified and justified in the FSAR. After 
issuance of a construction permit or COL, these items are not 
requirements for the licensee unless they are restated in the FSAR. For 
additional discussion, see Section V.D of this document.
    The availability controls, which are set forth in Appendix 19ACM of 
the generic DCD, were added to the information that is part of Tier 2 
to clarify that the availability controls are not operational 
requirements for the purposes of paragraph VIII.C. Rather, the 
availability controls are associated with specific design features. The 
availability controls may be changed if the associated design feature 
is changed under paragraph VIII.B. For additional discussion, see 
Section V.C of this document.
    Certain Tier 2 information has been designated in the generic DCD 
with brackets and italicized text as ``Tier 2*'' information and, as 
discussed in greater detail in the section-by-section analysis for 
Section H, a plant-specific departure from Tier 2* information requires 
prior NRC approval. However, the Tier 2* designation expires for some 
of this information when the facility first achieves full power after 
the finding required by 10 CFR 52.103(g). The process for changing Tier 
2* information and the time at which its status as Tier 2* expires is 
set forth in paragraph VIII.B.6. Some Tier 2* requirements concerning 
special preoperational tests are designated to be performed only for 
the first plant or first three plants referencing the ESBWR DCR. The 
Tier 2* designation for these selected tests will expire after the 
first plant or first three plants complete the specified tests. 
However, a COL action item requires that subsequent plants also perform 
the tests or justify that the results of the first-plant-only or first-
three-plants-only tests are applicable to the subsequent plant.
    The regulations at 10 CFR 50.59 set forth thresholds for permitting 
changes to a plant as described in the FSAR without NRC approval. 
Inasmuch as 10 CFR 50.59 is the primary change mechanism for operating 
nuclear plants, the NRC has determined that future plants referencing 
the ESBWR DCR should use thresholds as close to 10 CFR 50.59, as is 
practicable and appropriate for new reactors. Because of some 
differences in how the change control requirements are structured in 
the DCRs, certain definitions contained in 10 CFR 50.59 are not 
applicable to 10 CFR part 52 and are not being included in this rule. 
The NRC is including a definition for a ``departure from a method of 
evaluation'' (paragraph II.G), which is appropriate to include in this 
rulemaking so that the eight criteria in paragraph VIII.B.5.b will be 
implemented for new reactors as intended.

C. Scope and Contents (Section III)

    The purpose of Section III is to describe and define the scope and 
contents of this design certification and to set forth how 
documentation discrepancies or inconsistencies are to be resolved. 
Paragraph III.A is the required statement of the OFR for approval of 
the incorporation by reference of Tier 1, Tier 2, and the generic TS in 
Revision 10 of the ESBWR DCD, as well as the 20 documents listed in 
Table 1 of paragraph III.A. Paragraph III.B requires COL applicants and 
licensees to comply with the requirements of this appendix. The legal 
effect of incorporation by reference is that the incorporated material 
has the same legal status as if it were published in the Code of 
Federal Regulations. This material, like any other properly-issued 
regulation, has the force and effect of law. Tier 1 and Tier 2 
information, as well as the generic TS, have been combined into a 
single document called the generic DCD, in order to effectively control 
this information and facilitate its incorporation by reference into the 
rule. The generic DCD was prepared to meet the technical information 
contents of application requirements for design certifications under 10 
CFR 52.47(a) and the requirements of the OFR for incorporation by 
reference under 1 CFR part 51. One of the requirements of the OFR for 
incorporation by reference is that the design certification applicant 
must make the documents incorporated by reference available upon 
request after the final rule becomes effective. Therefore, paragraph 
III.A identifies a GEH representative to be contacted in order to 
obtain a copy of the DCD and the 20 documents incorporated by reference 
into the ESBWR design certification rule.

[[Page 61972]]

    Paragraphs III.A and III.B also identify the availability controls 
in Appendix 19ACM of the generic DCD as part of the Tier 2 information. 
During its review of the ESBWR design, the NRC determined that residual 
uncertainties associated with passive safety system performance 
increased the importance of nonsafety-related active systems in 
providing defense-in-depth functions that back-up the passive systems. 
As a result, GEH developed administrative controls to provide a high 
level of confidence that active systems having a significant safety 
role are available when challenged. GEH named these additional controls 
``availability controls.'' The NRC included this characterization in 
Section III to ensure that these availability controls are binding on 
applicants and licensees that reference this appendix and will be 
enforceable by the NRC. The NRC's evaluation of the availability 
controls is provided in Chapter 22 of the FSER.
    The generic DCD (master copy) and the 20 publicly-available 
documents listed in Table 1 of paragraph III.A are electronically 
accessible under the ADAMS Accession Nos. provided in paragraph III.A 
and at the OFR. Copies of these documents are also available at the 
NRC's PDR and from GEH as described in paragraph III.A. Questions 
concerning the accuracy of information in an application that 
references this appendix will be resolved by checking the master copy 
of the generic DCD or its referenced documents in ADAMS. If the design 
certification applicant makes a generic change (rulemaking) to the DCD 
under 10 CFR 52.63 and the change process provided in Section VIII, 
then at the completion of the rulemaking the NRC would request approval 
of the Director, OFR, for the revised master DCD. The NRC is requiring 
that the design certification applicant maintain an up-to-date copy of 
the master DCD that includes any generic changes it has made under 
paragraph X.A.1 because it is likely that most applicants intending to 
reference the standard design would obtain the generic DCD from the 
design certification applicant. Plant-specific changes to and 
departures from the generic DCD will be maintained by the applicant or 
licensee that references this appendix in a plant-specific DCD under 
paragraph X.A.2.
    In addition to requiring compliance with this appendix, paragraph 
III.B clarifies that the conceptual design information and GEH's 
evaluation of SAMDAs are not considered to be part of this appendix. 
The conceptual design information is for those portions of the plant 
that are outside the scope of the standard design and are contained in 
Tier 2 information. As provided by 10 CFR 52.47(a)(24), these 
conceptual designs are not part of this appendix and, therefore, are 
not applicable to an application that references this appendix. 
Therefore, the applicant is not required to conform to the conceptual 
design information that was provided by the design certification 
applicant. The conceptual design information, which consists of site-
specific design features, was required to facilitate the design 
certification review. Conceptual design information is neither Tier 1 
nor Tier 2. Section 1.8.2 of Tier 2 identifies the location of the 
conceptual design information. GEH's evaluation of various design 
alternatives to prevent and mitigate severe accidents does not 
constitute design requirements. The NRC's assessment of this 
information is discussed in Section IX of this document.
    Paragraphs III.C and III.D set forth the way potential conflicts 
are to be resolved. Paragraph III.C establishes the Tier 1 description 
in the DCD as controlling in the event of an inconsistency between the 
Tier 1 and Tier 2 information in the DCD. Paragraph III.D establishes 
the generic DCD as the controlling document in the event of an 
inconsistency between the DCD and the FSER (including Supplement No. 1) 
for the certified standard design.
    Paragraph III.E makes it clear that design activities that are 
wholly outside the scope of this design certification may be performed 
using actual site characteristics, provided the design activities do 
not affect Tier 1 or Tier 2, or conflict with the interface 
requirements in the DCD. This provision applies to site-specific 
portions of the plant, such as the administration building. Because 
this statement is not a definition, this provision has been located in 
Section III.

D. Additional Requirements and Restrictions (Section IV)

    Section IV sets forth additional requirements and restrictions 
imposed upon an applicant who references this appendix. Paragraph IV.A 
sets forth the information requirements for these applicants. This 
paragraph distinguishes between information and/or documents which must 
actually be included in the application or the DCD, versus those which 
may be incorporated by reference (i.e., referenced in the application 
as if the information or documents were included in the application). 
Any incorporation by reference in the application should be clear and 
should specify the title, date, edition, or version of a document, the 
page number(s), and table(s) containing the relevant information to be 
incorporated.
    Paragraph IV.A.1 requires an applicant who references this appendix 
to incorporate by reference this appendix in its application. The legal 
effect of such an incorporation by reference into the application is 
that this appendix is legally binding on the applicant or licensee. 
Paragraph IV.A.2.a requires that a plant-specific DCD be included in 
the initial application to ensure that the applicant commits to 
complying with the DCD. This paragraph also requires the plant-specific 
DCD to either include or incorporate by reference the generic DCD 
information. Further, this paragraph also requires the plant-specific 
DCD to use the same format as the generic DCD and reflect the 
applicant's proposed exemptions and departures from the generic DCD as 
of the time of submission of the application. The plant-specific DCD 
will be part of the plant's FSAR, along with information for the 
portions of the plant outside the scope of the referenced design. 
Paragraph IV.A.2.a also requires that the initial application include 
the reports on departures and exemptions as of the time of submission 
of the application.
    Paragraph IV.A.2.b requires that an application referencing this 
appendix include the reports required by paragraph X.B for exemptions 
and departures proposed by the applicant as of the date of submission 
of its application. Paragraph IV.A.2.c requires submission of plant-
specific TS for the plant that consists of the generic TS from Chapter 
16 of the DCD, with any changes made under paragraph VIII.C, and the TS 
for the site-specific portions of the plant that are either partially 
or wholly outside the scope of this design certification. The applicant 
must also provide the plant-specific information designated in the 
generic TS, such as bracketed values (refer to guidance provided in 
Interim Staff Guidance (ISG) DC/COL-ISG-8, ``Necessary Content of 
Plant-Specific Technical Specifications,'' ADAMS Accession No. 
ML083310259).
    Paragraph IV.A.2.d requires the applicant referencing this appendix 
to provide information demonstrating that the proposed site 
characteristics fall within the site parameters for this appendix and 
that the plant-specific interface requirements have been met as 
required by 10 CFR 52.79(d). If the proposed site has a characteristic 
that does not fall within one or more of the site parameters in the 
DCD, then the proposed site is unacceptable for this

[[Page 61973]]

design unless the applicant seeks an exemption under Section VIII and 
provides adequate justification for locating the certified design on 
the proposed site. Paragraph IV.A.2.e requires submission of 
information addressing COL action items, identified in the generic DCD 
as COL information in the application. The COL information identifies 
matters that need to be addressed by an applicant who references this 
appendix, as required by subpart C of 10 CFR part 52. An applicant may 
differ from or omit these items, provided that the difference or 
omission is identified and justified in its application. Based on the 
applicant's difference or omission, the NRC may impose additional 
licensing requirement(s) on the COL applicant as appropriate. Paragraph 
IV.A.2.f requires that the application include the information 
specified by 10 CFR 52.47(a) that is not within the scope of this rule, 
such as generic issues that must be addressed or operational issues not 
addressed by a design certification, in whole or in part, by an 
applicant that references this appendix. Paragraph IV.A.2.g requires 
that the application include information demonstrating that hurricane 
loads on those SSCs described in Section 3.3.2 of the generic DCD are 
either bounded by the total tornado loads analyzed in Section 3.3.2 of 
the generic DCD or will meet applicable NRC requirements with 
consideration of hurricane loads in excess of the total tornado loads. 
Paragraph IV.A.2.g further requires that hurricane-generated missile 
loads on those SSCs described in Section 3.5.2 of the generic DCD are 
either bounded by tornado-generated missile loads analyzed in Section 
3.5.1.4 of the generic DCD or will meet applicable NRC requirements 
with consideration of hurricane-generated missile loads in excess of 
the tornado-generated missile loads. Paragraph IV.A.2.h requires that 
the application include information demonstrating that SFP level 
instrumentation is designed to allow the connection of an independent 
power source and that the instrumentation will maintain its design 
accuracy following a power interruption or change in power source 
without recalibration. Paragraph IV.A.3 requires the applicant to 
physically include, not simply reference, the SUNSI (including 
proprietary information and security-related information) and SGI 
referenced in the DCD, or its equivalent, to ensure that the applicant 
has actual notice of these requirements.
    Paragraph IV.A.4 indicates requirements that must be met in cases 
where the COL applicant is not using the entity that was the original 
applicant for the design certification (or amendment) to supply the 
design for the applicant's use. Paragraph IV.A.4 requires that a COL 
applicant referencing this appendix include, as part of its 
application, a demonstration that an entity other than GEH Nuclear 
Energy is qualified to supply the ESBWR certified design unless GEH 
Nuclear Energy supplies the design for the applicant's use. This 
includes the non-public versions (or their equivalents) of the 
documents listed in Table 3 under section III.B of the SUPPLEMENTARY 
INFORMATION section of this document. In cases where a COL applicant is 
not using GEH Nuclear Energy to supply the ESBWR certified design, the 
required information would be used to support any NRC finding under 10 
CFR 52.73(a) that an entity other than the one originally sponsoring 
the design certification or design certification amendment is qualified 
to supply the certified design.
    Paragraph IV.B reserves to the Commission the right to determine in 
what manner this appendix may be referenced by an applicant for a 
construction permit or operating license under 10 CFR part 50. This 
determination may occur in the context of a subsequent rulemaking 
modifying 10 CFR part 52 or this DCR, or on a case-by-case basis in the 
context of a specific application for a 10 CFR part 50 construction 
permit or operating license. This provision is necessary because the 
previous DCRs were not implemented in the manner that was originally 
envisioned at the time that 10 CFR part 52 was promulgated. The NRC's 
concern is with the way ITAACs were developed and the lack of 
experience with design certifications in license proceedings. 
Therefore, it is appropriate that the Commission retain some discretion 
regarding the way this appendix could be referenced in a 10 CFR part 50 
licensing proceeding.

E. Applicable Regulations (Section V)

    The purpose of Section V is to specify the regulations that were 
applicable and in effect at the time this design certification was 
approved (i.e., as of the date specified in paragraph V.A, which would 
be the date that this appendix is approved by the Commission and signed 
by the Secretary of the Commission). These regulations consist of the 
technically relevant regulations identified in paragraph V.A, except 
for the regulations in paragraph V.B that are not applicable to this 
certified design.
    In paragraph V.B, the NRC identifies the regulations that do not 
apply to the ESBWR design. The Commission has determined that the ESBWR 
design should be exempt from portions of 10 CFR 50.34 as described in 
the FSER (NUREG-1966) and/or summarized below:
    Paragraph (f)(2)(iv) of 10 CFR 50.34--Contents of Construction 
Permit and Operating License Applications: Technical Information.
    This paragraph requires an applicant to provide a plant safety 
parameter display console that will display to operators a minimum set 
of parameters defining the safety status of the plant, capable of 
displaying a full range of important plant parameters and data trends 
on demand, and capable of indicating when process limits are being 
approached or exceeded. The ESBWR design integrates the safety 
parameter display system into the design of the nonsafety-related 
distribution control and information system, rather than uses a stand-
alone console. The safety parameter display system is described in 
Section 7.1.5 of the DCD.
    The NRC has also determined that the ESBWR design is approved to 
use the following alternative. Under 10 CFR 50.55a(a)(3), GEH requested 
NRC approval for the use of ASME Code Case N-782 as a proposed 
alternative to the rules of Section III, Subsection NCA-1140, regarding 
applied Code Editions and Addenda required by 10 CFR 50.55a(c), (d), 
and (e). ASME Code Case N-782 provides that the Code Edition and 
Addenda endorsed in a certified design or licensed by the regulatory 
authority may be used for systems and components constructed to ASME 
Code, Section III requirements. These alternative requirements are in 
lieu of the requirements that base the Edition and Addenda on the 
construction permit date. Reference to ASME Code Case N-782 will be 
included in component and system design specifications and design 
reports to permit certification of these specifications and reports to 
the Code Edition and Addenda cited in the DCD. The NRC's bases for 
approving the use of ASME Code Case N-782 as a proposed alternative to 
the requirements of ASME Section III Subsection NCA-1140 under 10 CFR 
50.55a(a)(3) for ESBWR are described in Section 5.2.1.1.3 of the FSER.

F. Issue Resolution (Section VI)

    The purpose of Section VI is to identify the scope of issues that 
are resolved by the NRC in this rulemaking and, therefore, are 
``matters resolved'' within the meaning and intent of 10 CFR 
52.63(a)(5). The section is divided into five parts: Paragraph A 
identifies

[[Page 61974]]

the NRC's safety findings in adopting this appendix, paragraph B 
identifies the scope and nature of issues which are resolved by this 
rulemaking, paragraph C identifies issues that are not resolved by this 
rulemaking, paragraph D identifies the backfit restrictions applicable 
to the Commission with respect to this appendix, and paragraph E 
identifies the availability of secondary references.
    Paragraph VI.A describes the nature of the Commission's findings in 
general terms and makes the findings required by 10 CFR 52.54 for the 
Commission's approval of this DCR. Furthermore, paragraph VI.A 
explicitly states the Commission's determination that this design 
provides adequate protection of the public health and safety.
    Paragraph VI.B sets forth the scope of issues that may not be 
challenged as a matter of right in subsequent proceedings. The 
introductory phrase of paragraph VI.B clarifies that issue resolution 
as described in the remainder of the paragraph extends to the 
delineated NRC proceedings referencing this appendix. The remainder of 
paragraph VI.B describes the categories of information for which there 
is issue resolution. Specifically, paragraph VI.B.1 provides that all 
nuclear safety issues arising from the Atomic Energy Act of 1954, as 
amended, that are associated with the information in the NRC staff's 
FSER (NUREG-1966 and Supplement No. 1), the Tier 1 and Tier 2 
information (including the availability controls in Appendix 19ACM of 
the generic DCD), the 20 documents referenced in Table 1 of paragraph 
III.A, and the rulemaking record for this appendix are resolved within 
the meaning of 10 CFR 52.63(a)(5). These resolved issues include the 
information referenced in the DCD that are requirements (i.e., 
``secondary references''), as well as all issues arising from SUNSI 
(including proprietary information and security-related information) 
and SGI that are intended to be requirements. However, paragraph VI.B.1 
expressly excludes from issue resolution: The HFE procedure development 
and training program development identified in Sections 18.9 and 18.10 
of the generic DCD; hurricane loads on those SSCs described in Section 
3.3.2 of the generic DCD that are not bounded by the total tornado 
loads analyzed in Section 3.3.2 of the generic DCD; hurricane-generated 
missile loads on those SSCs described in Section 3.5.2 of the generic 
DCD that are not bounded by tornado-generated missile loads analyzed in 
Section 3.5.1.4 of the generic DCD; or that SFP level instrumentation 
is designed to allow the connection of an independent power source, and 
that the instrumentation will maintain its design accuracy following a 
power interruption or change in power source without recalibration.
    Paragraph VI.B.2 provides for issue preclusion of SUNSI (including 
proprietary information and security-related information) and SGI, 
consisting of the fifty (50) non-publicly available documents listed in 
Tables 1.6-1 and 1.6-2 of Tier 2 of the ESBWR DCD, Revision 10.
    Paragraphs VI.B.3, VI.B.4, VI.B.5, and VI.B.6 clarify that approved 
changes to and departures from the DCD, which are accomplished in 
compliance with the relevant procedures and criteria in Section VIII, 
continue to be matters resolved in connection with this rulemaking. 
Paragraphs VI.B.4, VI.B.5, and VI.B.6, which characterize the scope of 
issue resolution in three situations, use the phrase ``but only for 
that plant.'' Paragraph VI.B.4 describes how issues associated with a 
DCR are resolved when an exemption has been granted for a plant 
referencing the DCR. Paragraph VI.B.5 describes how issues are resolved 
when a plant referencing the DCR obtains a license amendment for a 
departure from Tier 2 information. Paragraph VI.B.6 describes how 
issues are resolved when the applicant or licensee departs from the 
Tier 2 information on the basis of paragraph VIII.B.5, which will waive 
the requirement for NRC approval. In all three situations, after a 
matter (e.g., an exemption in the case of paragraph VI.B.4) is 
addressed for a specific plant referencing a DCR, the adequacy of that 
matter for that plant is resolved and will constitute part of the 
licensing basis for that plant. Therefore, that matter will not 
ordinarily be subject to challenge in any subsequent proceeding or 
action for that plant (e.g., an enforcement action) listed in the 
introductory portion of paragraph IV.B. By contrast, there will be no 
legally binding issue resolution on that subject matter for any other 
plant, or in a subsequent rulemaking amending the applicable DCR. 
However, the NRC's consideration of the safety, regulatory or policy 
issues necessary to the determination of the exemption or license 
amendment may, in appropriate circumstances, be relied upon as part of 
the basis for NRC action in other licensing proceedings or rulemaking.
    Paragraph VI.B.7 provides that, for those plants located on sites 
whose site characteristics fall within the site parameters assumed in 
the GEH evaluation of SAMDAs, all issues with respect to SAMDAs arising 
under the NEPA, associated with the information in the EA for this 
design and the information regarding SAMDAs in NEDO-33306, Revision 4, 
``ESBWR Severe Accident Mitigation Design Alternatives'' are also 
resolved within the meaning and intent of 10 CFR 52.63(a)(5). If a 
deviation from a site parameter is granted, the deviation applicant has 
the initial burden of demonstrating that the original SAMDA analysis 
still applies to the actual site characteristics; however, if the 
deviation is approved, requests for litigation at the COL stage must 
meet the requirements of 10 CFR 2.309 and present sufficient 
information to create a genuine controversy in order to obtain a 
hearing on the site parameter deviation.
    Paragraph VI.C reserves the right of the Commission to impose 
operational requirements on applicants that reference this appendix. 
This provision reflects the fact that only some operational 
requirements, including portions of the generic TS in Chapter 16 of the 
DCD, and no operational programs, such as operational quality assurance 
(QA), were completely or comprehensively reviewed by the NRC in this 
design certification rulemaking proceeding. Therefore, the special 
backfit and finality provisions of 10 CFR 52.63 apply only to those 
operational requirements that either the NRC completely reviewed and 
approved, or formed the basis for an NRC safety finding of the adequacy 
of the ESBWR, as documented in the NRC's FSER and Supplement No. 1 for 
the ESBWR. This is consistent with the currently approved design 
certifications in 10 CFR part 52, appendices A through D. Although 
information on operational matters is included in the DCDs of each of 
these currently approved designs, for the most part these design 
certifications do not provide approval for operational information, and 
none provide approval for operational ``programs'' (e.g., emergency 
preparedness programs, operational QA programs). Most operational 
information in the DCD simply serves as ``contextual information'' 
(i.e., information necessary to understand the design of certain SSCs 
and how they would be used in the overall context of the facility). The 
NRC did not use contextual information to support the NRC's safety 
conclusions and such information does not constitute the underlying 
safety bases for the adequacy of those SSCs. Thus, contextual 
operational information on any particular topic does not constitute one 
of the ``matters resolved'' under paragraph VI.B.
    The NRC notes that operational requirements may be imposed on

[[Page 61975]]

licensees referencing this design certification through the inclusion 
of license conditions in the license, or inclusion of a description of 
the operational requirement in the plant-specific FSAR.\5\ The NRC's 
choice of the regulatory vehicle for imposing the operational 
requirements will depend upon, among other things: (1) Whether the 
development and/or implementation of these requirements must occur 
prior to either the issuance of the COL or the Commission finding under 
10 CFR 52.103(g), and (2) the nature of the change controls that are 
appropriate given the regulatory, safety, and security significance of 
each operational requirement.
---------------------------------------------------------------------------

    \5\ Certain activities, ordinarily conducted following fuel load 
and therefore considered ``operational requirements,'' but which may 
be relied upon to support a Commission finding under 10 CFR 
52.103(g), may themselves be the subject of ITAAC to ensure their 
implementation prior to the 10 CFR 52.103(g) finding.
---------------------------------------------------------------------------

    Paragraph VI.C allows the NRC to impose future operational 
requirements (distinct from design matters) on applicants who reference 
this design certification. Also, license conditions for portions of the 
plant within the scope of this design certification (e.g., start-up and 
power ascension testing) are not restricted by 10 CFR 52.63. The 
requirement to perform these testing programs is contained in Tier 1 
information. However, ITAACs cannot be specified for these subjects 
because the matters to be addressed in these license conditions cannot 
be verified prior to fuel load and operation, when the ITAACs are 
satisfied. Therefore, another regulatory vehicle is necessary to ensure 
that licensees comply with the matters contained in the license 
conditions. License conditions for these areas cannot be developed now 
because this requires the type of detailed design information that will 
be developed during a COL review. In the absence of detailed design 
information to evaluate the need for and develop specific post-fuel 
load verifications for these matters, the Commission is reserving in 
this rule the right to impose, at the time of COL issuance, license 
conditions addressing post-fuel load verification activities for 
portions of the plant within the scope of this design certification.
    Paragraph VI.D reiterates the restrictions (contained in Section 
VIII) placed upon the Commission when ordering generic or plant-
specific modifications, changes or additions to SSCs, design features, 
design criteria, and ITAACs (paragraph VI.D.3 addresses ITAACs) within 
the scope of the certified design.
    Paragraph VI.E provides that the NRC will specify at an appropriate 
time the procedures for interested persons to obtain access to SUNSI 
(including proprietary information and security-related information) 
and SGI information for the ESBWR DCR. Access to such information would 
be for the sole purpose of requesting or participating in certain 
specified hearings, such as: (1) The hearing required by 10 CFR 52.85 
where the underlying application references this appendix; (2) any 
hearing provided under 10 CFR 52.103 where the underlying COL 
references this appendix; and (3) any other hearing relating to this 
appendix in which interested persons have the right to request an 
adjudicatory hearing.
    For proceedings where the notice of hearing was published before 
the effective date of the final rule, the Commission's order governing 
access to SUNSI and SGI shall be used to govern access to such 
information within the scope of the rulemaking. For proceedings in 
which the notice of hearing or opportunity for hearing is published 
after the effective date of the final rule, paragraph VI.E applies and 
governs access to SUNSI and SGI. For these proceedings, as stated in 
paragraph VI.E, the NRC will specify the access procedures at an 
appropriate time.
    For both a hearing required by 10 CFR 52.85 where the underlying 
application references this appendix, and in any hearing on ITAACs 
completion under 10 CFR 52.103, the NRC expects to follow its current 
practice of establishing the procedures by order at the time that the 
notice of hearing is published in the Federal Register. See, for 
example, Florida Power and Light Co., Combined License Application for 
the Turkey Point Units 6 & 7, Notice of Hearing, Opportunity To 
Petition for Leave To Intervene and Associated Order Imposing 
Procedures for Access to SUNSI and Safeguards Information for 
Contention Preparation (75 FR 34777; June 18, 2010); Notice of Receipt 
of Application for License; Notice of Consideration of Issuance of 
License; Notice of Hearing and Commission Order and Order Imposing 
Procedures for Access to SUNSI and Safeguards Information for 
Contention Preparation; In the Matter of AREVA Enrichment Services, LLC 
(Eagle Rock Enrichment Facility) (74 FR 38052; July 30, 2009).

G. Duration of This Appendix (Section VII)

    The purpose of Section VII is, in part, to specify the period 
during which this design certification may be referenced by an 
applicant for a COL, under 10 CFR 52.55. This section also states that 
the design certification remains valid for an applicant or licensee 
that references the design certification until the application is 
withdrawn or the license expires. Therefore, if an application 
references this design certification during the 15-year period, then 
the design certification will be effective until the application is 
withdrawn or the license issued on that application expires. Also, the 
design certification will be effective for the referencing licensee if 
the license is renewed. The NRC intends this appendix to remain valid 
for the life of the plant that references the design certification to 
achieve the benefits of standardization and licensing stability. This 
means that changes to, or plant-specific departures from, information 
in the plant-specific DCD must be made under the change processes in 
Section VIII for the life of the plant.

H. Processes for Changes and Departures (Section VIII)

    The purpose of Section VIII is to set forth the processes for 
generic changes to, or plant-specific departures (including exemptions) 
from, the DCD. The Commission adopted this restrictive change process 
in order to achieve a more stable licensing process for applicants and 
licensees that reference DCRs. Section VIII is divided into three 
paragraphs, which correspond to Tier 1, Tier 2, and operational 
requirements. The language of Section VIII distinguishes between 
generic changes to the DCD versus plant-specific departures from the 
DCD. Generic changes must be accomplished by rulemaking because the 
intended subject of the change is this DCR itself, as is contemplated 
by 10 CFR 52.63(a)(1). Consistent with 10 CFR 52.63(a)(3), any generic 
rulemaking changes are applicable to all plants, absent circumstances 
which render the change [``modification'' in the language of 10 CFR 
52.63(a)(3)] ``technically irrelevant.'' By contrast, plant-specific 
departures could be either a Commission-issued order to one or more 
applicants or licensees; or an applicant or licensee-initiated 
departure applicable only to that applicant's or licensee's plant(s), 
similar to a 10 CFR 50.59 departure or an exemption. Because these 
plant-specific departures will result in a DCD that is unique for that 
plant, Section X requires an applicant or licensee to maintain a plant-
specific DCD. For purposes of brevity, the following discussion refers 
to both generic changes and plant-specific departures as ``change 
processes.''

[[Page 61976]]

    Section VIII refers to an exemption from one or more requirements 
of this appendix and the criteria for granting an exemption. The NRC 
cautions that when the exemption involves an underlying substantive 
requirement (applicable regulation), then the applicant or licensee 
requesting the exemption must also show that an exemption from the 
underlying applicable requirement meets the criteria of 10 CFR 52.7.

Tier 1 Information

    The change processes for Tier 1 information are covered in 
paragraph VIII.A. Generic changes to Tier 1 are accomplished by 
rulemakings that amend the generic DCD and are governed by the 
standards in 10 CFR 52.63(a)(1) and 10 CFR 52.63(a)(2). No matter who 
proposes it, a generic change under 10 CFR 52.63(a)(1) will not be made 
to a certified design while it is in effect unless the change: (1) Is 
necessary for compliance with Commission regulations applicable and in 
effect at the time the certification was issued; (2) is necessary to 
provide adequate protection of the public health and safety or common 
defense and security; (3) reduces unnecessary regulatory burden and 
maintains protection to public health and safety and common defense and 
security; (4) provides the detailed design information necessary to 
resolve selected design acceptance criteria; (5) corrects material 
errors in the certification information; (6) substantially increases 
overall safety, reliability, or security of a facility and the costs of 
the change are justified; or (7) contributes to increased 
standardization of the certification information. The rulemakings must 
provide for notice and opportunity for public comment on the proposed 
change, as required by 10 CFR 52.63(a)(2). 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.
    Departures from Tier 1 may occur in two ways: (1) The Commission 
may order a licensee to depart from Tier 1, as provided in paragraph 
VIII.A.3; or (2) an applicant or licensee may request an exemption from 
Tier 1, as provided in paragraph VIII.A.4. If the Commission seeks to 
order a licensee to depart from Tier 1, paragraph VIII.A.3 requires 
that the Commission find both that the departure is necessary for 
adequate protection or for compliance and that special circumstances 
are present. Paragraph VIII.A.4 provides that exemptions from Tier 1 
requested by an applicant or licensee are governed by the requirements 
of 10 CFR 52.63(b)(1) and 52.98(f), which provide an opportunity for a 
hearing. In addition, the Commission will not grant requests for 
exemptions that may result in a significant decrease in the level of 
safety otherwise provided by the design.

Tier 2 Information

    The change processes for the three different categories of Tier 2 
information, namely, Tier 2, Tier 2*, and Tier 2* with a time of 
expiration, are set forth in paragraph VIII.B. The change process for 
Tier 2 has the same elements as the Tier 1 change process, but some of 
the standards for plant-specific orders and exemptions are different.
    The process for generic Tier 2 changes (including changes to Tier 
2* and Tier 2* with a time of expiration) tracks the process for 
generic Tier 1 changes. As set forth in paragraph VIII.B.1, generic 
Tier 2 changes are accomplished by rulemaking amending the generic DCD 
and are governed by the standards in 10 CFR 52.63(a)(1). No matter who 
proposes it, a generic change under 10 CFR 52.63(a)(1) will not be made 
to a certified design while it is in effect unless the change: (1) Is 
necessary for compliance with NRC regulations applicable and in effect 
at the time the certification was issued; (2) is necessary to provide 
adequate protection of the public health and safety or common defense 
and security; (3) reduces unnecessary regulatory burden and maintains 
protection to public health and safety and common defense and security; 
(4) provides the detailed design information necessary to resolve 
selected design acceptance criteria; (5) corrects material errors in 
the certification information; (6) substantially increases overall 
safety, reliability, or security of a facility and the costs of the 
change are justified; or (7) contributes to increased standardization 
of the certification information. If a generic change is made to Tier 
2* information, then the category and expiration, if necessary, of the 
new information will also be determined in the rulemaking and the 
appropriate change process for that new information would apply.
    Departures from Tier 2 may occur in five ways: (1) The Commission 
may order a plant-specific departure, as set forth in paragraph 
VIII.B.3; (2) an applicant or licensee may request an exemption from a 
Tier 2 requirement as set forth in paragraph VIII.B.4; (3) a licensee 
may make a departure without prior NRC approval under paragraph 
VIII.B.5; (4) the licensee may request NRC approval for proposed 
departures which do not meet the requirements in paragraph VIII.B.5 as 
provided in paragraph VIII.B.5.d; and (5) the licensee may request NRC 
approval for a departure from Tier 2* information under paragraph 
VIII.B.6.
    Similar to Commission-ordered Tier 1 departures and generic Tier 2 
changes, Commission-ordered Tier 2 departures cannot be imposed except 
when necessary either to bring the certification into compliance with 
the NRC's regulations applicable and in effect at the time of approval 
of the design certification or to ensure adequate protection of the 
public health and safety or common defense and security, as set forth 
in paragraph VIII.B.3. However, the special circumstances for the 
Commission-ordered Tier 2 departures do not have to outweigh any 
decrease in safety that may result from the reduction in 
standardization caused by the plant-specific order, as required by 10 
CFR 52.63(a)(4). The Commission determined that it was not necessary to 
impose an additional limitation similar to that imposed on Tier 1 
departures by 10 CFR 52.63(a)(4) and (b)(1). This type of additional 
limitation for standardization would unnecessarily restrict the 
flexibility of applicants and licensees with respect to Tier 2 
information.
    An applicant or licensee may request an exemption from Tier 2 
information as set forth in paragraph VIII.B.4. The applicant or 
licensee must demonstrate that the exemption complies with one of the 
special circumstances in 10 CFR 50.12(a). In addition, the Commission 
will not grant requests for exemptions that may result in a significant 
decrease in the level of safety otherwise provided by the design. 
However, the special circumstances for the exemption do not have to 
outweigh any decrease in safety that may result from the reduction in 
standardization caused by the exemption. If the exemption is requested 
by an applicant for a license, the exemption is subject to litigation 
in the same manner as other issues in the license hearing, consistent 
with 10 CFR 52.63(b)(1). If the exemption is requested by a licensee, 
then the exemption is subject to litigation in the same manner as a 
license amendment.
    Paragraph VIII.B.5 allows an applicant or licensee to depart from 
Tier 2 information, without prior NRC approval, if the proposed 
departure does not involve a change to, or departure from, Tier 1 or 
Tier 2* information, TS, or does not require a license amendment under 
paragraphs VIII.B.5.b or VIII.B.5.c. The TS referred to in

[[Page 61977]]

VIII.B.5.a of this paragraph are the TS in Chapter 16 of the generic 
DCD, including bases, for departures made prior to issuance of the COL. 
After issuance of the COL, the plant-specific TS are controlling under 
paragraph VIII.B.5. The bases for the plant-specific TS will be 
controlled by the bases control program, which is specified in the 
plant-specific TS administrative controls section. The requirement for 
a license amendment in paragraph VIII.B.5.b will be similar to the 
requirement in 10 CFR 50.59 and apply to all information in Tier 2 
except for the information that resolves the severe accident issues.
    The NRC concludes that the resolution of ex-vessel severe accident 
design features should be preserved and maintained in the same fashion 
as all other safety issues that were resolved during the design 
certification review (refer to SRM on SECY-90-377, ``Requirements for 
Design Certification Under 10 CFR Part 52,'' dated February 15, 1991, 
ADAMS Accession No. ML003707892). However, because of the increased 
uncertainty in ex-vessel severe accident issue resolutions, the NRC has 
adopted separate criteria in paragraph VIII.B.5.c for determining if a 
departure from information that resolves ex-vessel severe accident 
design features would require a license amendment. For purposes of 
applying the special criteria in paragraph VIII.B.5.c, ex-vessel severe 
accident resolutions are limited to design features where the intended 
function of the design feature is relied upon to resolve postulated 
accidents when the reactor core has melted and exited the reactor 
vessel, and the containment is being challenged. These design features 
are identified in Sections 19.2.3, 19.3.2, 19.3.3, 19.3.4, and 
Appendices 19A and 19B of the DCD, with other issues, and are described 
in other sections of the DCD. Therefore, the location of design 
information in the DCD is not important to the application of this 
special procedure for ex-vessel severe accident design features. 
However, the special procedure in paragraph VIII.B.5.c does not apply 
to design features that resolve so-called ``beyond design-basis 
accidents'' or other low-probability events. The important aspect of 
this special procedure is that it is limited to ex-vessel severe 
accident design features, as defined above. Some design features may 
have intended functions to meet ``design basis'' requirements and to 
resolve ``severe accidents.'' If these design features are reviewed 
under paragraph VIII.B.5, then the appropriate criteria from either 
paragraphs VIII.B.5.b or VIII.B.5.c are selected depending upon the 
function being changed.
    An applicant or licensee that plans to depart from Tier 2 
information, under paragraph VIII.B.5, is required to prepare an 
evaluation that provides the bases for the determination that the 
proposed change does not require a license amendment or involve a 
change to Tier 1 or Tier 2* information, or a change to the TS, as 
explained above. In order to achieve the NRC's goals for design 
certification, the evaluation needs to consider all of the matters that 
were resolved in the DCD, such as generic issue resolutions that are 
relevant to the proposed departure. The benefits of the early 
resolution of safety issues would be lost if departures from the DCD 
were made that violated these resolutions without appropriate review.
    The evaluation of the relevant matters needs to consider the 
proposed departure over the full range of power operation from startup 
to shutdown, as it relates to anticipated operational occurrences, 
transients, DBAs, and severe accidents. The evaluation must also 
include a review of all relevant secondary references from the DCD 
because Tier 2 information, which is intended to be treated as a 
requirement, is contained in the secondary references. The evaluation 
should consider Tables 14.3-1a through 14.3-1c and 19.2-3 of the 
generic DCD to ensure that the proposed change does not impact Tier 1 
information. These tables contain cross-references from the safety 
analyses and probabilistic risk assessment (PRA) in Tier 2 to the 
important parameters that were included in Tier 1.
    Paragraph VIII.B.5.d addresses information described in the DCD to 
address aircraft impacts, in accordance with 10 CFR 52.47(a)(28). Under 
10 CFR 52.47(a)(28), applicants are required to include the information 
required by 10 CFR 50.150(b) in their DCD. Under 10 CFR 50.150(b), 
applications for standard design certifications are required to 
include:
    1. A description of the design features and functional capabilities 
identified as a result of the AIA required by 10 CFR 50.150(a)(1); and
    2. A description of how such design features and functional 
capabilities meet the assessment requirements in 10 CFR 50.150(a)(1).
    An applicant or licensee who changes this information is required 
to consider the effect of the changed design feature or functional 
capability on the original AIA required by 10 CFR 50.150(a). The 
applicant or licensee is also required to describe in the plant-
specific DCD how the modified design features and functional 
capabilities continue to meet the assessment requirements in 10 CFR 
50.150(a)(1). Submittal of this updated information is governed by the 
reporting requirements in Section X.B.
    In an adjudicatory proceeding (e.g., for issuance of a COL), a 
person who believes that an applicant or licensee has not complied with 
paragraph VIII.B.5 when departing from Tier 2 information is permitted 
to petition to admit such a contention into the proceeding under 
paragraph VIII.B.5.f. This provision was included because an incorrect 
departure from the requirements of this appendix essentially places the 
departure outside of the scope of the Commission's safety finding in 
the design certification rulemaking. Therefore, it follows that 
properly founded contentions alleging such incorrectly implemented 
departures cannot be considered ``resolved'' by this rulemaking. As set 
forth in paragraph VIII.B.5.f, the petition must comply with the 
requirements of 10 CFR 2.309 and show that the departure does not 
comply with paragraph VIII.B.5. Other persons may file a response to 
the petition under 10 CFR 2.309. If, on the basis of the petition and 
any responses, the presiding officer in the proceeding determines that 
the required showing has been made, the matter shall be certified to 
the Commission for its final determination. In the absence of a 
proceeding, petitions alleging nonconformance with paragraph VIII.B.5 
requirements applicable to Tier 2 departures will be treated as 
petitions for enforcement action under 10 CFR 2.206.
    Paragraph VIII.B.6 provides a process for departing from Tier 2* 
information. The creation of and restrictions on changing Tier 2* 
information resulted from the development of the Tier 1 information for 
the Advanced Boiling Water Reactor design certification (appendix A to 
10 CFR part 52) and the System 80+ design certification (appendix B to 
10 CFR part 52). During this development process, these applicants 
requested that the amount of information in Tier 1 be minimized to 
provide additional flexibility for an applicant or licensee who 
references these appendices. Also, many codes, standards, and design 
processes that were not specified in Tier 1 as acceptable for meeting 
ITAACs were specified in Tier 2. The result of these departures is that 
certain significant information exists only in Tier 2 and the 
Commission does not want this significant information to be changed 
without prior NRC approval. This Tier 2* information is identified in 
the

[[Page 61978]]

generic DCD with italicized text and brackets (see Table 1D-1 in 
Appendix 1D of the ESBWR DCD).
    Although the Tier 2* designation was originally intended to last 
for the lifetime of the facility, like Tier 1 information, the NRC 
determined that some of the Tier 2* information could expire when the 
plant first achieves full (100 percent) power, after the finding 
required by 10 CFR 52.103(g), while other Tier 2* information must 
remain in effect throughout the life of the facility. The factors 
determining whether Tier 2* information could expire after full power 
is first achieved (first full power) were whether the Tier 1 
information would govern these areas after first full power and the 
NRC's determination that prior approval was required before 
implementation of the change due to the significance of the 
information. Therefore, certain Tier 2* information listed in paragraph 
VIII.B.6.c ceases to retain its Tier 2* designation after full power 
operation is first achieved following the Commission finding under 10 
CFR 52.103(g). Thereafter, that information is deemed to be Tier 2 
information that is subject to the departure requirements in paragraph 
VIII.B.5. By contrast, the Tier 2* information identified in paragraph 
VIII.B.6.b retains its Tier 2* designation throughout the duration of 
the license, including any period of license renewal.
    Certain preoperational tests in paragraph VIII.B.6.c are designated 
to be performed only for the first plant that references this appendix. 
GEH's basis for performing these ``first-plant-only'' preoperational 
tests is provided in Section 14.2.8 of the DCD. The NRC found GEH's 
basis for performing these tests and its justification for only 
performing the tests on the first plant acceptable. The NRC's decision 
was based on the need to verify that plant-specific manufacturing and/
or construction variations do not adversely impact the predicted 
performance of certain passive safety systems, while recognizing that 
these special tests will result in significant thermal transients being 
applied to critical plant components. The NRC concludes that the range 
of manufacturing or construction variations that could adversely affect 
the relevant passive safety systems would be adequately disclosed after 
performing the designated tests on the first plant. The Tier 2* 
designation for these tests will expire after the first plant completes 
these tests, as indicated in paragraph VIII.B.6.c.
    If Tier 2* information is changed in a generic rulemaking, the 
designation of the new information (Tier 1, 2*, or 2) will also be 
determined in the rulemaking and the appropriate process for future 
changes will apply. If a plant-specific departure is made from Tier 2* 
information, then the new designation will apply only to that plant. If 
an applicant who references this design certification makes a departure 
from Tier 2* information, the new information will be subject to 
litigation in the same manner as other plant-specific issues in the 
licensing hearing. If a licensee makes a departure from Tier 2* 
information, it will be treated as a license amendment under 10 CFR 
50.90 and the finality will be determined under paragraph VI.B.5. Any 
requests for departures from Tier 2* information that affects Tier 1 
must also comply with the requirements in paragraph VIII.A.

Operational Requirements

    The change process for TS and other operational requirements in the 
DCD is set forth in paragraph VIII.C. This change process has elements 
similar to the Tier 1 and Tier 2 change processes in paragraphs VIII.A 
and VIII.B, but with significantly different change standards. Because 
of the different finality status for TS and other operational 
requirements (refer to paragraph V.F of this document), the Commission 
designated a special category of information, consisting of the TS and 
other operational requirements, with its own change process in proposed 
paragraph VIII.C. The key to using the change processes proposed in 
Section VIII is to determine if the proposed change or departure 
requires a change to a design feature described in the generic DCD. If 
a design change is required, then the appropriate change process in 
paragraph VIII.A or VIII.B applies. However, if a proposed change to 
the TS or other operational requirements does not require a change to a 
design feature in the generic DCD, then paragraph VIII.C applies. The 
language in paragraph VIII.C also distinguishes between generic 
(Chapter 16 of the DCD) and plant-specific TS to account for the 
different treatment and finality accorded TS before and after a license 
is issued.
    The process in paragraph VIII.C.1 for making generic changes to the 
generic TS in Chapter 16 of the DCD or other operational requirements 
in the generic DCD is accomplished by rulemaking and governed by the 
backfit standards in 10 CFR 50.109. The determination of whether the 
generic TS and other operational requirements were completely reviewed 
and approved in the design certification rulemaking is based upon the 
extent to which the NRC reached a safety conclusion in the FSER on this 
matter. If it cannot be determined, in the absence of a specific 
statement, that the TS or operational requirement was comprehensively 
reviewed and finalized in the design certification rulemaking, then 
there is no backfit restriction under 10 CFR 50.109 because no prior 
position, consistent with paragraph VI.B, was taken on this safety 
matter. Generic changes made under paragraph VIII.C.1 are applicable to 
all applicants or licensees (refer to paragraph VIII.C.2), unless the 
change is irrelevant because of a plant-specific departure.
    Some generic TS and availability controls contain values in 
brackets [ ]. The brackets are placeholders indicating that the NRC's 
review is not complete and represent a requirement that the applicant 
for a COL referencing the ESBWR DCR must replace the values in brackets 
with final plant-specific values (refer to guidance provided in Interim 
Staff Guidance DC/COL-ISG-8, ``Necessary Content of Plant-Specific 
Technical Specifications''). The values in brackets are neither part of 
the DCR nor are they binding. Therefore, the replacement of bracketed 
values with final plant-specific values does not require an exemption 
from the generic TS or availability controls.
    Plant-specific departures may occur by either a Commission order 
under paragraph VIII.C.3 or an applicant's exemption request under 
paragraph VIII.C.4. The basis for determining if the TS or operational 
requirement was completely reviewed and approved for these processes is 
the same as for paragraph VIII.C.1 above. If the TS or operational 
requirement is comprehensively reviewed and finalized in the design 
certification rulemaking, then the Commission must demonstrate that 
special circumstances are present before ordering a plant-specific 
departure. If not, there is no restriction on plant-specific changes to 
the TS or operational requirements, prior to the issuance of a license, 
provided a design change is not required. Although the generic TS were 
reviewed and approved by the NRC staff in support of the design 
certification review, the Commission intends to consider the lessons 
learned from subsequent operating experience during its licensing 
review of the plant-specific TS. The process for petitioning to 
intervene on a TS or operational requirement contained in paragraph 
VIII.C.5 is similar to other issues in a licensing hearing, except that 
the petitioner must also demonstrate why special circumstances are 
present pursuant to 10 CFR 2.335.

[[Page 61979]]

    Finally, the generic TS will have no further effect on the plant-
specific TS after the issuance of a license that references this 
appendix. The bases for the generic TS will be controlled by the change 
process in paragraph VIII.C. After a license is issued, the bases will 
be controlled by the bases change provision set forth in the 
administrative controls section of the plant-specific TS.

I. [RESERVED] (Section IX)

    This section is reserved for future use. As discussed in Section IV 
of the SUPPLEMENTARY INFORMATION section of this document, the matters 
discussed in this section of earlier design certification rules--
inspections, tests, analyses, and acceptance criteria--are now 
addressed in the substantive provisions of 10 CFR part 52. Accordingly, 
there is no need to repeat these regulatory provisions in the ESBWR 
design certification rule.

J. Records and Reporting (Section X)

    The purpose of Section X is to set forth the requirements that will 
apply to maintaining records of changes to and departures from the 
generic DCD, which are to be reflected in the plant-specific DCD. 
Section X also sets forth the requirements for submitting reports 
(including updates to the plant-specific DCD) to the NRC. This section 
of the appendix is similar to the requirements for records and reports 
in 10 CFR part 50, except for minor differences in information 
collection and reporting requirements.
    Paragraph X.A.1 requires that a generic DCD and the SUNSI 
(including proprietary information and security-related information) 
and SGI referenced in the generic DCD be maintained by the applicant 
for this rule. The generic DCD concept was developed, in part, to meet 
the OFR requirements for incorporation by reference, including public 
availability of documents incorporated by reference. However, the SUNSI 
(including proprietary information and security-related information) 
and SGI could not be included in the generic DCD because they are not 
publicly available. Nonetheless, the SUNSI (including proprietary 
information and security-related information) and SGI was reviewed by 
the NRC and, as stated in paragraph VI.B.2, the NRC considers the 
information to be resolved within the meaning of 10 CFR 52.63(a)(5). 
Because this information is not in the generic DCD, this information, 
or its equivalent, is required to be provided by an applicant for a 
license referencing this DCR. Paragraph X.A.1 requires the design 
certification applicant to maintain the SUNSI (including proprietary 
information and security-related information) and SGI, which it 
developed and used to support its design certification application. 
This ensures that the referencing applicant has direct access to this 
information from the design certification applicant, if it has 
contracted with the applicant to provide the SUNSI (including 
proprietary information and security-related information) and SGI to 
support its license application. The NRC may also inspect this 
information if it was not submitted to the NRC (e.g., the AIA required 
by 10 CFR 50.150). Only the generic DCD and 20 publicly-available 
documents referenced in the DCD are identified and incorporated by 
reference into this rule. The generic DCD and the NRC-approved version 
of the SUNSI (including proprietary information and security-related 
information) and SGI must be maintained by the applicant (GEH) for the 
period of time that this appendix may be referenced.
    Paragraphs X.A.2 and X.A.3 place recordkeeping requirements on the 
applicant or licensee who references this design certification so that 
its plant-specific DCD accurately reflects both generic changes to the 
generic DCD and plant-specific departures made under Section VIII. The 
term ``plant-specific'' is used in paragraph X.A.2 and other sections 
of this appendix to distinguish between the generic DCD that is 
incorporated by reference into this appendix and the plant-specific DCD 
that the applicant is required to submit under paragraph IV.A. The 
requirement to maintain changes to the generic DCD is explicitly stated 
to ensure that these changes are not only reflected in the generic DCD, 
which will be maintained by the applicant for design certification, but 
also in the plant-specific DCD. Therefore, records of generic changes 
to the DCD will be required to be maintained by both entities to ensure 
that both entities have up-to-date DCDs.
    Paragraph X.A.4.a requires the applicant to maintain a copy of the 
AIA performed to comply with the requirements of 10 CFR 50.150(a) for 
the term of the certification (including any period of renewal). This 
provision, which is consistent with 10 CFR 50.150(c)(3), will 
facilitate any NRC inspections of the assessment that the NRC decides 
to conduct. Similarly, paragraph X.A.4.b requires an applicant or 
licensee who references this appendix to maintain a copy of the AIA 
performed to comply with the requirements of 10 CFR 50.150(a) 
throughout the pendency of the application and for the term of the 
license (including any period of renewal). This provision is consistent 
with 10 CFR 50.150(c)(4). For all applicants and licensees, the 
supporting documentation retained onsite should describe the 
methodology used in performing the assessment, including the 
identification of potential design features and functional capabilities 
to show that the acceptance criteria in 10 CFR 50.150(a)(1) will be 
met.
    Paragraph X.A does not place recordkeeping requirements on site-
specific information that is outside the scope of this rule. As 
discussed in paragraph V.D of this document, the FSAR required by 10 
CFR 52.79 will contain the plant-specific DCD and the site-specific 
information for a facility that references this rule. The phrase 
``site-specific portion of the final safety analysis report'' in 
paragraph X.B.3.c refers to the information that is contained in the 
FSAR for a facility (required by 10 CFR 52.79) but is not part of the 
plant-specific DCD (required by paragraph IV.A). Therefore, this rule 
does not require that duplicate documentation be maintained by an 
applicant or licensee that references this rule because the plant-
specific DCD is part of the FSAR for the facility.
    Paragraph X.B.1 requires applicants or licensees that reference 
this rule to submit reports, which describe departures from the DCD and 
include a summary of the written evaluations. The requirement for the 
written evaluations is set forth in paragraph X.A.1. The frequency of 
the report submittals is set forth in paragraph X.B.3. The requirement 
for submitting a summary of the evaluations is similar to the 
requirement in 10 CFR 50.59(d)(2).
    Paragraph X.B.2 requires applicants or licensees that reference 
this rule to submit updates to the DCD, which include both generic 
changes and plant-specific departures. The frequency for submitting 
updates is set forth in paragraph X.B.3. The requirements in paragraph 
X.B.3 for submitting the reports and updates will vary according to 
certain time periods during a facility's lifetime. If a potential 
applicant for a COL who references this rule decides to depart from the 
generic DCD prior to submission of the application, then paragraph 
X.B.3.a will require that the updated DCD be submitted as part of the 
initial application for a license. Under paragraph X.B.3.b, the 
applicant may submit any subsequent updates to its plant-specific DCD 
along with its amendments to the application provided that the 
submittals are made at least once per year. Because amendments to an 
application are typically made more frequently than

[[Page 61980]]

once a year, this should not be an excessive burden on the applicant.
    Paragraph X.B.3.b also requires semi-annual submission of the 
reports required by paragraph X.B.1 throughout the period of 
application review and construction. The NRC will use the information 
in the reports to help plan the NRC's inspection and oversight during 
this phase when the licensee is conducting detailed design, procurement 
of components and equipment, construction, and preoperational testing. 
In addition, the NRC will use the information in making its finding on 
ITAACs under 10 CFR 52.103(g), as well as any finding on interim 
operation under Section 189.a(1)(B)(iii) of the AEA. Once a facility 
begins operation (for a COL under 10 CFR part 52, after the Commission 
has made a finding under 10 CFR 52.103(g)), the frequency of reporting 
will be governed by the requirements in paragraph X.B.3.c.

VIII. Agreement State Compatibility

    Under the ``Policy Statement on Adequacy and Compatibility of 
Agreement States Programs,'' approved by the Commission on June 20, 
1997, and published in the Federal Register (62 FR 46517; September 3, 
1997), this rule is classified as compatibility ``NRC.'' Compatibility 
is not required for Category ``NRC'' regulations. The NRC program 
elements in this category are those that relate directly to areas of 
regulation reserved to the NRC by the AEA or the provisions of Title 10 
of the Code of Federal Regulations, and although an Agreement State may 
not adopt program elements reserved to the NRC, it may wish to inform 
its licensees of certain requirements by a mechanism that is consistent 
with a particular State's administrative procedure laws, but does not 
confer regulatory authority on the State.

IX. 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:
    SECY-11-0006, ``Proposed Rule--ESBWR       ML102220172
     Design Certification''.
    Staff Requirements Memorandum for SECY-11- ML110670047
     0006, ``Proposed Rule--ESBWR Design
     Certification''.
    General Electric Company Application for   ML052450245
     Final Design Approval and Design
     Certification of ESBWR Standard Plant
     Design.
    ESBWR Design Control Document, Revision 9  ML103440266
    ESBWR Final Safety Evaluation Report       ML14100A304
     (NUREG-1966).
    ESBWR FSER Final Chapters................  ML103470210
    Final Design Approval for the Economic     ML110540310
     Simplified Boiling Water Reactor.
    ESBWR Draft Environmental Assessment.....  ML102220247
    ESBWR Proposed Rule Federal Register       ML110610353
     Notice, 76 FR 16549, March 24, 2011.
Public Comments on the March 2011 Proposed
 Rule:
    Comment (1) from Farouk D. Baxter on       ML102350160
     Environmental Impact Statement for Two
     AP1000 Units at Levy County Site.
    Comment submission S1 from Paul C.         ML110880057
     Daugherty.
    Comment submission S2 from Farouk D.       ML110880315
     Baxter.
    Comment submission S3 from Patricia T.     ML11158A088
     Birnie, Chair, General Electric
     Stockholders' Alliance.
    Comment submission S4 from anonymous.....  ML11187A303
    Comment submission P1, Emergency Petition  ML111040472
     To Suspend All Pending Reactor Licensing
     Decisions and Related Rulemaking
     Decisions Pending Investigation of
     Lessons Learned From Fukushima Daiichi
     Nuclear Power Station Accident (initial).
    Comment submission P2, Emergency Petition  ML111080855
     To Suspend All Pending Reactor Licensing
     Decisions and Related Rulemaking
     Decisions Pending Investigation of
     Lessons Learned From Fukushima Daiichi
     Nuclear Power Station Accident (amended).
    Comment submission P3, Declaration of Dr.  ML111100618
     Arjun Makhijani in Support of Emergency
     Petition To Suspend All Pending Reactor
     Licensing Decisions and Relating
     Rulemaking Decisions Pending
     Investigation of Lessons Learned From
     Fukushima Daiichi Nuclear Power Station
     Accident.
    Comment submission P4, Comment of Jerald   ML11124A103
     Head on Behalf of GE-Hitachi Nuclear
     Energy Opposing Petition To Suspend All
     Pending Reactor Licensing Decisions and
     Related Rulemaking Decisions Pending
     Investigation of Lessons Learned From
     Fukushima Daiichi Nuclear Power Station
     Accident.
    Comment submission P5, Petitioners' Reply  ML111260637
     to Responses to Emergency Petition To
     Suspend All Pending Reactor Licensing
     Decisions and Related Rulemaking
     Decisions Pending Investigation of
     Lessons Learned From Fukushima Daiichi
     Nuclear Power Station Accident.
    Comment submission P6, Comments of Terry   ML112430118
     J. Lodge on PR 52, NEPA Requirement To
     Address Safety and Environmental
     Implications of the Fukushima Task Force
     Report From ESBWR, Fermi 3 Intervenors.
    Public Comments Compilation--Final Rule--  ML113130141
     ESBWR Design Certification (RIN 3150-
     AI85).
Supplemental Safety Evaluation for the ESBWR
 Design Certification:
    Advanced Supplemental Safety Evaluation    ML14043A134
     Report for the Economic Simplified
     Boiling-Water Reactor Standard Plant
     Design.
    Supplemental Safety Evaluation Report for  ML14155A333
     the Economic Simplified Boiling-Water
     Reactor Standard Plant Design.
Supplemental Proposed Rule Documents:
    ESBWR Design Control Document, Rev. 10...  ML14104A929
    ESBWR Supplemental Proposed Rule Federal   ML14043A508
     Register Notice, 79 FR 25715, May 6,
     2014.
Final Rule Documents:
    SECY-14-0081, ``Final Rule--ESBWR Design   ML111730346
     Certification''.
    Staff Requirements Memorandum for SECY-14- ML14259A545
     0081, ``Final Rule--ESBWR Design
     Certification''.
    ESBWR Final Environmental Assessment.....  ML111730382
Other Documents Relevant to the ESBWR
 Rulemaking:
    NEDO-33306, Revision 4, ``ESBWR Severe     ML102990433
     Accident Mitigation Design
     Alternatives''.
    NEDO-33312, Rev. 5, ``ESBWR Steam Dryer    ML13344B157
     Acoustic Load Definition''.

[[Page 61981]]

 
    NEDO-33313, Rev. 5, ``ESBWR Steam Dryer    ML13344B158
     Structural Evaluation''.
    NEDO-33338, Revision 1, ``ESBWR Feedwater  ML091380173
     Temperature Operating Domain Transient
     and Accident Analysis''.
    NEDO-33408P, Revision 5, ``ESBWR Steam     ML13344B159
     Dryer--Plant-Based Load Evaluation
     Methodology, PBLE01 Model Description''.
    Commission Memorandum and Order (CLI-11-   ML112521106
     05), September 9, 2011 (available on the
     NRC Web site in Volume 74 at https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr0750/ nuregs/staff/sr0750/).
    Commission Order, ``Scheduling Order of    ML111101277
     the Secretary Regarding Petitions To
     Suspend Adjudicatory, Licensing and
     Rulemaking Activities (PR 52 re ESBWR
     Design Certification)''.
    Order EA-12-049, ``Order Modifying         ML12054A735
     Licenses With Regard to Requirements for
     Mitigation Strategies for Beyond-Design-
     Basis External Events''.
    Order EA 12-051, ``Order Modifying         ML12054A679
     Licenses With Regard to Reliable Spent
     Fuel Pool Instrumentation''.
    Staff Requirements Memorandum for SECY-90- ML003707892
     377, ``Requirements for Design
     Certification Under 10 CFR Part 52''.
    SECY-94-084, ``Policy and Technical        ML003708068
     Issues Associated With the Regulatory
     Treatment of Non-Safety Systems in
     Passive Plant Designs''.
    Staff Requirements Memorandum for SECY-96- ML003754873
     077, ``Certification of Two Evolutionary
     Designs''.
    SECY-96-077, ``Certification of Two        ML003708129
     Evolutionary Designs''.
    Staff Requirements Memorandum for SECY-11- ML112310021
     0093, ``Near-Team Report and
     Recommendations for Agency Actions
     Following the Events in Japan''.
    SECY-11-0093, ``Enclosure: The Near-Term   ML111861807
     Task Force Review of Insights From the
     Fukushima Dai-ichi Accident''.
    Staff Requirements Memorandum for SECY-11- ML112920034
     0117, ``Proposed Charter for the Longer-
     Term Review of Lessons Learned From the
     March 11, 2011, Japanese Earthquake and
     Tsunami''.
    SECY-11-0117, ``Proposed Charter for the   ML11231A723
     Longer-Term Review of Lessons Learned
     From the March 11, 2011, Japanese
     Earthquake and Tsunami''.
    SECY-11-0124, ``Recommended Actions To Be  ML11245A127
     Taken Without Delay From The Near-Term
     Task Force Report''.
    SECY-11-0137, ``Prioritization of          ML11269A204
     Recommended Actions To Be Taken in
     Response to Fukushima Lessons Learned''.
    Staff Requirements Memorandum for SECY-12- ML120690347
     0025, ``Proposed Orders and Requests for
     Information in Response to Lessons
     Learned From Japan's March 11, 2011,
     Great Tohoku Earthquake and Tsunami''.
    SECY-12-0025, ``Proposed Orders and        ML12039A103
     Requests for Information in Response to
     Lessons Learned From Japan's March 11,
     2011, Great Tohoku Earthquake and
     Tsunami''.
    SECY-14-0046, ``Fifth 6-Month Status       ML14064A523
     Update on Response to Lessons Learned
     From Japan's March 11, 2011, Great
     T[omacr]hoku Earthquake and Subsequent
     Tsunami''.
    Regulatory Guide 1.13, ``Spent Fuel        ML070310035
     Storage Facility Design Basis''.
    Regulatory Guide 1.20, ``Comprehensive     ML070260376
     Vibration Assessment Program for Reactor
     Internals During Preoperational and
     Initial Startup Testing''.
    Regulatory Guide 1.27, ``Ultimate Heat     ML003739996
     Sink for Nuclear Power Plants (for
     Comment)''.
    Regulatory Guide 1.76, ``Design-Basis      ML070360253
     Tornado and Tornado Missiles for Nuclear
     Power Plants''.
    Regulatory Guide 1.117, ``Tornado Design   ML003739346
     Classification''.
    Regulatory Guide 1.143, ``Design Guidance  ML003740200
     for Radioactive Waste Management
     Systems, Structures, and Components
     Installed in Light-Water-Cooled Nuclear
     Power Plants''.
    Regulatory Guide 1.206, Section C.I.1,     ML070630005
     ``Standard Format and Content of
     Combined License Applications--
     Introduction and General Description of
     the Plant''.
    Regulatory Guide 1.221, ``Design-Basis     ML110940303
     Hurricane and Hurricane Missiles for
     Nuclear Power Plants''.
    NUREG-0700, Revision 2, ``Human-Systems    ML021700337
     Interface Design Review Guidelines''      ML021700342
     (three volumes).                          ML021700371
    NUREG-0711, Revision 2, ``Human Factors    ML040770540
     Engineering Program Review Model''.
    NUREG-0711, Revision 3, ``Human Factors    ML12324A013
     Engineering Program Review Model''.
    NUREG-0800, Section 3.8.4, Revision 2,     ML070550054
     ``Other Seismic Category I Structures,''
     Appendix D, ``Guidance on Spent Fuel
     Pool Racks''.
    NUREG-0800, Section 3.9.2, Revision 3,     ML070230008
     ``Dynamic Testing and Analysis of
     Systems, Structures, and Components''.
    NUREG-0800, Section 3.9.5, Revision 3,     ML070230009
     ``Reactor.
    Pressure Vessel Internals''..............
    NUREG-0800, SRP Section 6.4, Revision 3,   ML070550069
     ``Control Room Habitability System''.
    NUREG-0800, SRP Section 9.1.2, Revision    ML070550057
     4, ``New and Spent Fuel Storage''.
    NUREG-0800, SRP Section 13.4, Revision 3,  ML070470463
     ``Operational Programs''.
    NUREG-0800, SRP Section 13.5.2.1,          ML070100635
     Revision 2, ``Operating and Emergency
     Operating Procedures''.
    NUREG-0800, SRP Section 18, Revision 2,    ML070670253
     ``Human Factors Engineering''.
    NUREG-1242, ``NRC Review of Electric       ML100610048
     Power Research Institute's Advanced       ML100430013
     Light Water Reactor Utility Requirements  ML063620331
     Document, Evolutionary Plant Designs''    ML070600372
     (five volumes).                           ML070600373
    NRC Bulletin 2012-01: Design               ML12074A115
     Vulnerability in Electric Power System.
    Interim Staff Guidance DC/COL-ISG-8,       ML083310259
     ``Necessary Content of Plant-Specific
     Technical Specifications''.
    JLD-ISG-2012-03 Revision 0, ``Compliance   ML12221A339
     With Order EA-12-051, Reliable Spent
     Fuel Pool Instrumentation,''.
    NEI 12-02, Revision 1, ``Industry          ML122400399
     Guidance for Compliance With NRC Order
     EA-12-051, To Modify Licenses With
     Regard to Reliable Spent Fuel Pool
     Instrumentation''.
    ``Clarifications Requested by NRC Staff    ML11269A093
     on Economic Simplified Boiling Water
     Reactor Fuel Design''.
    Audit Report, ``ESBWR Fuel Seismic Audit   ML112860614
     Summary''.

[[Page 61982]]

 
    Notice of Violation, ``ESBWR AIA           ML102740292
     Inspection Report Inspection, NRC
     Inspection Report No. 0520000/10/2010-
     201 and Notice of Violation''.
    Reply to Notice of Violation, NRC          ML103010047
     Inspection Report 052000010-10-201.
    GE-Hitachi Nuclear Energy Americas, LLC,   ML103400150
     Reply to Notice of Violation, NRC IR
     052000010-10-201.
    ACRS Memorandum--Final Rule--ESBWR Design  ML113120076
     Certification (RIN 3150-AI85).
    ACRS Memorandum--ESBWR Design              ML11340A043
     Certification Rulemaking and
     Supplemental Final Safety Evaluation
     Report.
    ACRS Memorandum--Supplemental Final        ML14107A263
     Safety Evaluation Report on the General
     Electric-Hitachi Nuclear Energy (GEH)
     Application for Certification of the
     Economic Simplified Boiling Water
     Reactor (ESBWR) Design.
    ACRS Memorandum--Final Rule--ESBWR Design  ML14196A207
     Certification (RIN 3150-AI85).
    Regulatory History of Design               ML003761550
     Certification \6\.
------------------------------------------------------------------------

X. Voluntary Consensus Standards
---------------------------------------------------------------------------

    \6\ The regulatory history of the NRC's design certification 
reviews is a package of documents that is available in NRC's PDR and 
Electronic Reading Room. This history spans the period during which 
the NRC simultaneously developed the regulatory standards for 
reviewing these designs and the form and content of the rules that 
certified the designs.
---------------------------------------------------------------------------

    The National Technology Transfer and Advancement Act of 1995 (Act), 
Pub. L. 104-113, 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 final rule, the NRC is approving the 
ESBWR standard plant design for use in nuclear power plant licensing 
under 10 CFR part 50 or part 52. Design certifications are not generic 
rulemakings establishing a generally applicable standard with which all 
10 CFR parts 50 and 52 nuclear power plant licensees or applicants for 
SDAs, design certifications, or manufacturing licenses must comply. 
Design certifications are NRC approvals of specific nuclear power plant 
designs by rulemaking. Furthermore, design certifications are initiated 
by an applicant for rulemaking, rather than by the NRC. For these 
reasons, the NRC concludes that the Act does not apply to this final 
rule.

XI. Finding of No Significant Environmental Impact: Availability

    The NRC has determined under NEPA, and the NRC's regulations in 
subpart A, ``National Environmental Policy Act; Regulations 
Implementing Section 102(2),'' of 10 CFR part 51, ``Environmental 
Protection Regulations for Domestic Licensing and Related Regulatory 
Functions,'' that this DCR is not a major Federal action significantly 
affecting the quality of the human environment and, therefore, an 
environmental impact statement (EIS) is not required. The NRC's generic 
determination in this regard is reflected in 10 CFR 51.32(b)(1). The 
basis for the NRC's categorical exclusion in this regard, as discussed 
in the 2007 final rule amending 10 CFR parts 51 and 52 (August 28, 
2007; 72 FR 49352-49566), is based upon the following considerations. A 
DCR does not authorize the siting, construction, or operation of a 
facility referencing any particular design; it only codifies the ESBWR 
design in a rule. The NRC will evaluate the environmental impacts and 
issue an EIS as appropriate under NEPA as part of the application for 
the construction and operation of a facility referencing any particular 
DCR.
    In addition, consistent with 10 CFR 51.30(d) and 10 CFR 51.32(b), 
the NRC has prepared a final EA (ADAMS Accession No. ML111730382) for 
the ESBWR design addressing various design alternatives to prevent and 
mitigate severe accidents. The EA is based, in part, upon the NRC's 
review of GEH's evaluation of various design alternatives to prevent 
and mitigate severe accidents in NEDO-33306, Revision 4, ``ESBWR Severe 
Accident Mitigation Design Alternatives.'' Based upon review of GEH's 
evaluation, the Commission concludes that: (1) GEH identified a 
reasonably complete set of potential design alternatives to prevent and 
mitigate severe accidents for the ESBWR design; (2) none of the 
potential design alternatives are justified on the basis of cost-
benefit considerations; and (3) it is unlikely that other design 
changes would be identified and justified during the term of the design 
certification on the basis of cost-benefit considerations because the 
estimated core damage frequencies for the ESBWR are very low on an 
absolute scale. These issues are considered resolved for the ESBWR 
design.
    The NRC requested comments on the draft EA but the comments 
received did not include anything to suggest that: (i) A rule 
certifying the ESBWR standard design would be a major Federal action, 
or (ii) the SAMDA evaluation omitted a design alternative that should 
have been considered or incorrectly considered the costs and benefits 
of the alternatives it did consider. Therefore, no change to the EA was 
warranted. All environmental issues concerning SAMDAs associated with 
the information in the final EA and NEDO-33306 are considered resolved 
for facility applications referencing the ESBWR design if the site 
characteristics at the site proposed in the facility application fall 
within the site parameters specified in NEDO-33306.
    The final EA, upon which the Commission's finding of no significant 
impact is based, and the ESBWR DCD are available for examination and 
copying at the NRC's PDR, One White Flint North, Room O-1 F21, 11555 
Rockville Pike, Rockville, Maryland 20852.

XII. Paperwork Reduction Act

    This rule contains new or amended information collection 
requirements that are subject to the Paperwork Reduction Act of 1995 
(44 U.S.C. 3501, et seq.). These requirements were approved by the 
Office of Management and Budget (OMB), control number 3150-0151. The 
burden to the public for these information collections is estimated to 
average 15 hours per response.
    Send comments on any aspect of these information collections, 
including suggestions for reducing the burden, to the Records and FOIA/
Privacy Services Branch (T-5 F52), U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, or by Internet electronic mail to 
INFOCOLLECTS.RESOURCE@NRC.GOV; and to the Desk Officer, Office of 
Information and Regulatory Affairs, NEOB-10202, (3150-0151), Office of 
Management and Budget, Washington, DC 20503.

Public Protection Notification

    The NRC may not conduct or sponsor, and a person is not required to 
respond

[[Page 61983]]

to, a request for information or an information collection requirement 
unless the requesting document displays a currently valid OMB control 
number.

XIII. Regulatory Analysis

    The NRC has not prepared a regulatory analysis for this final rule. 
The NRC prepares regulatory analyses for rulemakings that establish 
generic regulatory requirements applicable to all licensees. Design 
certifications are not generic rulemakings in the sense that design 
certifications do not establish standards or requirements with which 
all licensees must comply. Rather, design certifications are NRC 
approvals of specific nuclear power plant designs by rulemaking, which 
then may be voluntarily referenced by applicants for COLs. Furthermore, 
design certification rulemakings are initiated by an applicant for a 
design certification, rather than the NRC. Preparation of a regulatory 
analysis in this circumstance would not be useful because the design to 
be certified is proposed by the applicant rather than the NRC. For 
these reasons, the NRC concludes that preparation of a regulatory 
analysis is neither required nor appropriate.

XIV. Regulatory Flexibility Certification

    Under the Regulatory Flexibility Act (5 U.S.C. 605(b)), the NRC 
certifies that this rule does not have a significant economic impact on 
a substantial number of small entities. This final rule provides for 
certification of a nuclear power plant design. Neither the design 
certification applicant, nor prospective nuclear power plant licensees 
who reference this DCR, fall within the scope of the definition of 
``small entities'' set forth in the Regulatory Flexibility Act or the 
size standards established by the NRC (10 CFR 2.810). Thus, this rule 
does not fall within the purview of the Regulatory Flexibility Act.

XV. Backfitting and Issue Finality

    The NRC has determined that this final rule does not constitute a 
backfit as defined in the backfit rule (10 CFR 50.109) and that it is 
not inconsistent with any applicable issue finality provision in 10 CFR 
part 52.
    This initial DCR does not constitute backfitting as defined in the 
backfit rule (10 CFR 50.109) because there are no operating licenses 
under 10 CFR part 50 referencing this DCR.
    This initial DCR is not inconsistent with any applicable issue 
finality provision in 10 CFR part 52 because it does not impose new or 
changed requirements on existing DCRs in appendices A through D to 10 
CFR part 52, and no COLs or manufacturing licenses issued by the NRC at 
this time reference a final ESBWR DCR. Although there are several COL 
applications referencing the application for the ESBWR DCR, there is no 
issue finality protection accorded to such a COL applicant under either 
10 CFR 52.63 or 10 CFR 52.83.
    For these reasons, neither a backfit analysis nor a discussion 
addressing the issue finality provisions in 10 CFR part 52 was prepared 
for this rule.

XVI. Congressional Review Act

    In accordance with the Congressional Review Act of 1996 (5 U.S.C. 
801-808), the NRC has determined that this action is not a major rule 
and has verified this determination with the Office of Information and 
Regulatory Affairs of the Office of Management and Budget.

XVII. 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 
31883).

XVIII. Availability of Guidance

    The NRC will not be issuing guidance for this rulemaking. The NRC 
has previously published relevant guidance in RG 1.206, ``Combined 
License Applications for Nuclear Power Plants (LWR Edition).'' This RG 
provides guidance for preparing an application for a COL under 10 CFR 
part 52, including guidance related to referencing a design 
certification in that application. Each DCR is similar in its content 
and structure. Therefore, the existing guidance in RG 1.206 is adequate 
to support this DCR.

List of Subjects in 10 CFR Part 52

    Administrative practice and procedure, Antitrust, Backfitting, 
Combined license, Early site permit, Emergency planning, Fees, 
Incorporation by reference, Inspection, Limited work authorization, 
Nuclear power plants and reactors, Probabilistic risk assessment, 
Prototype, Reactor siting criteria, Redress of site, Reporting and 
recordkeeping requirements, Standard design, Standard design 
certification.

    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 adopting 
the following amendments to 10 CFR part 52.

PART 52--LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER 
PLANTS

0
1. The authority citation for 10 CFR part 52 continues to read as 
follows:

    Authority:  Atomic Energy Act secs. 103, 104, 147, 149, 161, 
181, 182, 183, 185, 186, 189, 223, 234 (42 U.S.C. 2133, 2201, 2167, 
2169, 2232, 2233, 2235, 2236, 2239, 2282); Energy Reorganization Act 
secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); 
Government Paperwork Elimination Act sec. 1704 (44 U.S.C. 3504 
note); Energy Policy Act of 2005, Pub. L. 109-58, 119 Stat. 594 
(2005).


0
2. In Sec.  52.11, paragraph (b) is revised to read as follows:


Sec.  52.11  Information collection requirements: OMB approval.

* * * * *
    (b) The approved information collection requirements contained in 
this part appear in Sec. Sec.  52.7, 52.15, 52.16, 52.17, 52.29, 52.35, 
52.39, 52.45, 52.46, 52.47, 52.57, 52.63, 52.75, 52.77, 52.79, 52.80, 
52.93, 52.99, 52.110, 52.135, 52.136, 52.137, 52.155, 52.156, 52.157, 
52.158, 52.171, 52.177, and appendices A, B, C, D, E, and N of this 
part.

0
3. A new Appendix E to 10 CFR part 52 is added to read as follows:

Appendix E to Part 52--Design Certification Rule for the ESBWR Design

I. Introduction

    Appendix E constitutes the standard design certification for the 
Economic Simplified Boiling-Water Reactor (ESBWR) design, in 
accordance with 10 CFR part 52, subpart B. The applicant for 
certification of the ESBWR design is GE-Hitachi Nuclear Energy.

II. Definitions

    A. Generic design control document (generic DCD) means the 
document containing the Tier 1 and Tier 2 information and generic 
technical specifications that is incorporated by reference into this 
appendix.
    B. Generic technical specifications (generic TS) means the 
information required by 10 CFR 50.36 and 50.36a for the portion of 
the plant that is within the scope of this appendix.
    C. Plant-specific DCD means that portion of the combined license 
(COL) final safety analysis report (FSAR) that sets forth both the 
generic DCD information and any plant-specific changes to generic 
DCD information.
    D. Tier 1 means the portion of the design-related information 
contained in the generic DCD that is approved and certified by this 
appendix (Tier 1 information). The design descriptions, interface 
requirements, and site parameters are derived from Tier 2 
information. Tier 1 information includes:
    1. Definitions and general provisions;

[[Page 61984]]

    2. Design descriptions;
    3. Inspections, tests, analyses, and acceptance criteria 
(ITAACs);
    4. Significant site parameters; and
    5. Significant interface requirements.
    E. Tier 2 means the portion of the design-related information 
contained in the generic DCD that is approved but not certified by 
this appendix (Tier 2 information). Compliance with Tier 2 is 
required, but generic changes to and plant-specific departures from 
Tier 2 are governed by Section VIII of this appendix. Compliance 
with Tier 2 provides a sufficient, but not the only acceptable, 
method for complying with Tier 1. Compliance methods differing from 
Tier 2 must satisfy the change process in Section VIII of this 
appendix. Regardless of these differences, an applicant or licensee 
must meet the requirement in paragraph III.B of this appendix to 
reference Tier 2 when referencing Tier 1. Tier 2 information 
includes:
    1. Information required by Sec. Sec.  52.47(a) and 52.47(c), 
with the exception of generic TS and conceptual design information;
    2. Supporting information on the inspections, tests, and 
analyses that will be performed to demonstrate that the acceptance 
criteria in the ITAACs have been met;
    3. COL action items (COL license information), which identify 
certain matters that must be addressed in the site-specific portion 
of the FSAR by an applicant who references this appendix. These 
items constitute information requirements but are not the only 
acceptable set of information in the FSAR. An applicant may depart 
from or omit these items, provided that the departure or omission is 
identified and justified in the FSAR. After issuance of a 
construction permit or COL, these items are not requirements for the 
licensee unless such items are restated in the FSAR; and
    4. The availability controls in Appendix 19ACM of the DCD.
    F. Tier 2* means the portion of the Tier 2 information, 
designated as such in the generic DCD, which is subject to the 
change process in paragraph VIII.B.6 of this appendix. This 
designation expires for some Tier 2* information under paragraph 
VIII.B.6 of this appendix.
    G. Departure from a method of evaluation described in the plant-
specific DCD used in establishing the design bases or in the safety 
analyses means:
    1. Changing any of the elements of the method described in the 
plant-specific DCD unless the results of the analysis are 
conservative or essentially the same; or
    2. Changing from a method described in the plant-specific DCD to 
another method unless that method has been approved by the NRC for 
the intended application.
    H. All other terms in this appendix have the meaning set out in 
10 CFR 50.2, 10 CFR 52.1, or Section 11 of the Atomic Energy Act of 
1954, as amended, as applicable.

III. Scope and Contents

    A. Incorporation by reference approval. The documents in Table 1 
are approved for incorporation by reference by the Director of the 
Office of the Federal Register under 5 U.S.C. 552(a) and 1 CFR part 
51. You may obtain copies of the generic DCD from Jerald G. Head, 
Senior Vice President, Regulatory Affairs, GE-Hitachi Nuclear 
Energy, 3901 Castle Hayne Road, MC A-18, Wilmington, NC 28401, 
telephone: 1-910-819-5692. You can view the generic DCD online in 
the NRC Library at https://www.nrc.gov/reading-rm/adams.html. In 
ADAMS, search under the ADAMS Accession No. listed in Table 1. If 
you do not have access to ADAMS or if you have problems accessing 
documents located in ADAMS, contact the NRC's Public Document Room 
(PDR) reference staff at 1-800-397-4209, 1-301-415-3747, or by email 
at PDR.Resource@nrc.gov. These documents can also be viewed at the 
Federal rulemaking Web site, https://www.regulations.gov, by 
searching for documents filed under Docket ID NRC-2010-0135. Copies 
of these documents are available for examination and copying at the 
NRC's PDR located at Room O-1F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852. Copies are also available 
for examination at the NRC Library located at Two White Flint North, 
11545 Rockville Pike, Rockville, Maryland 20852, telephone: 301-415-
5610, email: Library.Resource@nrc.gov. All approved material is 
available for inspection at the National Archives and Records 
Administration (NARA). For information on the availability of this 
material at NARA, call 1-202-741-6030 or go to https://www.archives.gov/federal-register/cfr/ibrlocations.html.

                           Table 1--Documents Approved for Incorporation by Reference
----------------------------------------------------------------------------------------------------------------
             Document No.                       Document title                     ADAMS Accession No.
----------------------------------------------------------------------------------------------------------------
GE Hitachi:
    26A6642AB Rev. 10.................  ESBWR Design Control Document,  ML14104A929 (package)
                                         Revision 10, Tier 1, dated
                                         April 2014.
    26A6642AB Rev. 10.................  ESBWR Design Control Document,  ML14104A929 (package)
                                         Revision 10, Tier 2, dated
                                         April 2014.
Bechtel Power Corporation:
    BC-TOP-3-A........................  ``Tornado and Extreme Wind      ML14093A218
                                         Design Criteria for Nuclear
                                         Power Plants,'' Topical
                                         Report, Revision 3, August
                                         1974.
    BC-TOP-9A.........................  ``Design of Structures for      ML14093A217
                                         Missile Impact,'' Topical
                                         Report, Revision 2, September
                                         1974.
General Electric:
    GEZ-4982A.........................  General Electric Large Steam    ML14093A215
                                         Turbine Generator Quality
                                         Control Program, The STG
                                         Global Supply Chain Quality
                                         Management System (MFGGLO-GEZ-
                                         0010) Revision 1.2, February
                                         7, 2006.
GE Nuclear Energy:
    NEDO-11209-04A....................  ``GE Nuclear Energy Quality     ML14093A209
                                         Assurance Program
                                         Description,'' Class 1,
                                         Revision 8, March 31, 1989.
    NEDO-31960-A......................  ``BWR Owners' Group Long-Term   ML14093A212
                                         Stability Solutions Licensing
                                         Methodology,'' Class I,
                                         November 1995.
    NEDO-31960-A--Supplement 1........  ``BWR Owners' Group Long-Term   ML14093A211
                                         Stability Solutions Licensing
                                         Methodology,'' Class I,
                                         November 1995.
    NEDO-32465-A......................  GE Nuclear Energy and BWR       ML14093A210
                                         Owners' Group, ``Reactor
                                         Stability Detect and Suppress
                                         Solutions Licensing Basis
                                         Methodology for Reload
                                         Applications,'' Class I,
                                         August 1996.
GE-Hitachi Nuclear Energy:
    NEDO-33181........................  ``NP-2010 COL Demonstration     ML14248A297
                                         Project Quality Assurance
                                         Plan,'' Revision 6, August
                                         2009.
    NEDO-33219........................  ``ESBWR Human Factors           ML100350104
                                         Engineering Functional
                                         Requirements Analysis
                                         Implementation Plan,''
                                         Revision 4, Class I, February
                                         2010.
    NEDO-33260........................  ``Quality Assurance             ML14248A648
                                         Requirements for Suppliers of
                                         Equipment and Services to the
                                         GEH ESBWR Project,'' Revision
                                         5, Class I, April 2008.
    NEDO-33262........................  ``ESBWR Human Factors           ML100340030
                                         Engineering Operating
                                         Experience Review
                                         Implementation Plan,''
                                         Revision 3, Class I, January
                                         2010.
    NEDO-33266........................  ``ESBWR Human Factors           ML100350167
                                         Engineering Staffing and
                                         Qualifications Implementation
                                         Plan,'' Revision 3, Class I,
                                         January 2010.

[[Page 61985]]

 
    NEDO-33267........................  ``ESBWR Human Factors           ML100330609
                                         Engineering Human Reliability
                                         Analysis Implementation
                                         Plan,'' Revision 4, Class I,
                                         January 2010.
    NEDO-33277........................  ``ESBWR Human Factors           ML100270770
                                         Engineering Human Performance
                                         Monitoring Implementation
                                         Plan,'' Revision 4, Class I,
                                         January 2010.
    NEDO-33278........................  ``ESBWR Human Factors           ML100270468
                                         Engineering Design
                                         Implementation Plan,''
                                         Revision 4, Class I, January
                                         2010.
    NEDO-33289........................  ``ESBWR Reliability Assurance   ML14248A662
                                         Program,'' Revision 2, Class
                                         II, September 2008.
    NEDO-33337........................  ``ESBWR Initial Core Transient  ML091130628
                                         Analyses,'' Revision 1, Class
                                         I, April 2009.
    NEDO-33338........................  ``ESBWR Feedwater Temperature   ML091380173
                                         Operating Domain Transient
                                         and Accident Analysis,''
                                         Revision 1, Class I, May 2009.
    NEDO-33373-A......................  ``Dynamic, Load-Drop, and       ML102990226 (part 1)
                                         Thermal-Hydraulic Analyses     ML102990228 (part 2)
                                         for ESBWR Fuel Racks,''
                                         Revision 5, Class I, October
                                         2010.
    NEDO-33411........................  ``Risk Significance of          ML100610417
                                         Structures, Systems and
                                         Components for the Design
                                         Phase of the ESBWR,''
                                         Revision 2, Class I, February
                                         2010.
----------------------------------------------------------------------------------------------------------------

    B. An applicant or licensee referencing this appendix, in 
accordance with Section IV of this appendix, shall incorporate by 
reference and comply with the requirements of this appendix, 
including Tier 1, Tier 2 (including the availability controls in 
Appendix 19ACM of the DCD), and the generic TS except as otherwise 
provided in this appendix. Conceptual design information in the 
generic DCD and the evaluation of severe accident mitigation design 
alternatives in NEDO-33306, Revision 4, ``ESBWR Severe Accident 
Mitigation Design Alternatives,'' are not part of this appendix.
    C. If there is a conflict between Tier 1 and Tier 2 of the DCD, 
then Tier 1 controls.
    D. If there is a conflict between the generic DCD and either the 
application for design certification of the ESBWR design or NUREG-
1966, ``Final Safety Evaluation Report Related to Certification of 
the ESBWR Standard Design,'' (FSER) and Supplement No. 1 to NUREG-
1966, then the generic DCD controls.
    E. Design activities for structures, systems, and components 
that are wholly outside the scope of this appendix may be performed 
using site characteristics, provided the design activities do not 
affect the DCD or conflict with the interface requirements.

IV. Additional Requirements and Restrictions

    A. An applicant for a COL who references this appendix shall, in 
addition to complying with the requirements of Sec. Sec.  52.77, 
52.79, and 52.80, comply with the following requirements:
    1. Incorporate by reference, as part of its application, this 
appendix.
    2. Include, as part of its application:
    a. A plant-specific DCD containing the same type of information 
and using the same organization and numbering as the generic DCD for 
the ESBWR design, either by including or incorporating by reference 
the generic DCD information, and as modified and supplemented by the 
applicant's exemptions and departures;
    b. The reports on departures from and updates to the plant-
specific DCD required by paragraph X.B of this appendix;
    c. Plant-specific TS, consisting of the generic and site-
specific TS that are required by 10 CFR 50.36 and 50.36a;
    d. Information demonstrating that the site characteristics fall 
within the site parameters and that the interface requirements have 
been met;
    e. Information that addresses the COL action items;
    f. Information required by Sec.  52.47(a) that is not within the 
scope of this appendix;
    g. Information demonstrating that hurricane loads on those 
structures, systems, and components described in Section 3.3.2 of 
the generic DCD are either bounded by the total tornado loads 
analyzed in Section 3.3.2 of the generic DCD or will meet applicable 
NRC requirements with consideration of hurricane loads in excess of 
the total tornado loads; and hurricane-generated missile loads on 
those structures, systems, and components described in Section 3.5.2 
of the generic DCD are either bounded by tornado-generated missile 
loads analyzed in Section 3.5.1.4 of the generic DCD or will meet 
applicable NRC requirements with consideration of hurricane-
generated missile loads in excess of the tornado-generated missile 
loads; and
    h. Information demonstrating that the spent fuel pool level 
instrumentation is designed to allow the connection of an 
independent power source, and that the instrumentation will maintain 
its design accuracy following a power interruption or change in 
power source without requiring recalibration.
    3. Include, in the plant-specific DCD, the sensitive, 
unclassified, non-safeguards information (including proprietary 
information and security-related information) and safeguards 
information referenced in the ESBWR generic DCD.
    4. Include, as part of its application, a demonstration that an 
entity other than GE-Hitachi Nuclear Energy is qualified to supply 
the ESBWR design unless GE-Hitachi Nuclear Energy supplies the 
design for the applicant's use.
    B. The Commission reserves the right to determine in what manner 
this appendix may be referenced by an applicant for a construction 
permit or operating license under 10 CFR part 50.

V. Applicable Regulations

    A. Except as indicated in paragraph B of this section, the 
regulations that apply to the ESBWR design are in 10 CFR parts 20, 
50, 73, and 100, codified as of October 6, 2014, that are applicable 
and technically relevant, as described in the FSER (NUREG-1966) and 
Supplement No. 1.
    B. The ESBWR design is exempt from portions of the following 
regulations:
    1. Paragraph (f)(2)(iv) of 10 CFR 50.34--Contents of 
Applications: Technical Information--codified as of October 6, 2014.

VI. Issue Resolution

    A. The Commission has determined that the structures, systems, 
components, and design features of the ESBWR design comply with the 
provisions of the Atomic Energy Act of 1954, as amended, and the 
applicable regulations identified in Section V of this appendix; and 
therefore, provide adequate protection to the health and safety of 
the public. A conclusion that a matter is resolved includes the 
finding that additional or alternative structures, systems, 
components, design features, design criteria, testing, analyses, 
acceptance criteria, or justifications are not necessary for the 
ESBWR design.
    B. The Commission considers the following matters resolved 
within the meaning of Sec.  52.63(a)(5) in subsequent proceedings 
for issuance of a COL, amendment of a COL, or renewal of a COL, 
proceedings held under Sec.  52.103, and enforcement proceedings 
involving plants referencing this appendix:
    1. All nuclear safety issues associated with the information in 
the FSER and Supplement No. 1; Tier 1, Tier 2 (including referenced 
information, which the context indicates is intended as 
requirements, and the availability controls in Appendix 19ACM of the 
DCD), the 20 documents referenced in Table 1 of paragraph III.A, and 
the rulemaking record for certification of the ESBWR design, with 
the exception of: generic TS and other operational requirements such 
as human factors engineering procedure development and training 
program development in Sections 18.9 and 18.10 of the generic DCD; 
hurricane loads on those structures, systems, and components 
described in Section 3.3.2 of the generic DCD that are not bounded 
by the total tornado loads analyzed in Section 3.3.2 of the generic 
DCD; hurricane-generated missile loads on those structures, systems, 
and

[[Page 61986]]

components described in Section 3.5.2 of the generic DCD that are 
not bounded by tornado-generated missile loads analyzed in Section 
3.5.1.4 of the generic DCD; and spent fuel pool level 
instrumentation design in regard to the connection of an independent 
power source, and how the instrumentation will maintain its design 
accuracy following a power interruption or change in power source 
without recalibration;
    2. All nuclear safety and safeguards issues associated with the 
referenced information in the 50 non-public documents in Tables 1.6-
1 and 1.6-2 of Tier 2 of the DCD which contain sensitive 
unclassified non-safeguards information (including proprietary 
information and security-related information) and safeguards 
information and which, in context, are intended as requirements in 
the generic DCD for the ESBWR design, with the exception of human 
factors engineering procedure development and training program 
development in Chapters 18.9 and 18.10 of the generic DCD;
    3. All generic changes to the DCD under and in compliance with 
the change processes in paragraphs VIII.A.1 and VIII.B.1 of this 
appendix;
    4. All exemptions from the DCD under and in compliance with the 
change processes in paragraphs VIII.A.4 and VIII.B.4 of this 
appendix, but only for that plant;
    5. All departures from the DCD that are approved by license 
amendment, but only for that plant;
    6. Except as provided in paragraph VIII.B.5.f of this appendix, 
all departures from Tier 2 under and in compliance with the change 
processes in paragraph VIII.B.5 of this appendix that do not require 
prior NRC approval, but only for that plant;
    7. All environmental issues concerning severe accident 
mitigation design alternatives associated with the information in 
the NRC's Environmental Assessment for the ESBWR design (ADAMS 
Accession No. ML111730382) and NEDO-33306, Revision 4, ``ESBWR 
Severe Accident Mitigation Design Alternatives,'' (ADAMS Accession 
No. ML102990433) for plants referencing this appendix whose site 
characteristics fall within those site parameters specified in NEDO-
33306.
    C. The Commission does not consider operational requirements for 
an applicant or licensee who references this appendix to be matters 
resolved within the meaning of Sec.  52.63(a)(5). The Commission 
reserves the right to require operational requirements for an 
applicant or licensee who references this appendix by rule, 
regulation, order, or license condition.
    D. Except under the change processes in Section VIII of this 
appendix, the Commission may not require an applicant or licensee 
who references this appendix to:
    1. Modify structures, systems, components, or design features as 
described in the generic DCD;
    2. Provide additional or alternative structures, systems, 
components, or design features not discussed in the generic DCD; or
    3. Provide additional or alternative design criteria, testing, 
analyses, acceptance criteria, or justification for structures, 
systems, components, or design features discussed in the generic 
DCD.
    E. The NRC will specify at an appropriate time the procedures to 
be used by an interested person who seeks to review portions of the 
design certification or references containing safeguards information 
or sensitive unclassified non-safeguards information (including 
proprietary information, such as trade secrets and commercial or 
financial information obtained from a person that are privileged or 
confidential (10 CFR 2.390 and 10 CFR part 9), and security-related 
information), for the purpose of participating in the hearing 
required by Sec.  52.85, the hearing provided under Sec.  52.103, or 
in any other proceeding relating to this appendix in which 
interested persons have a right to request an adjudicatory hearing.

VII. Duration of This Appendix

    This appendix may be referenced for a period of 15 years from 
November 14, 2014, except as provided for in Sec. Sec.  52.55(b) and 
52.57(b). This appendix remains valid for an applicant or licensee 
who references this appendix until the application is withdrawn or 
the license expires, including any period of extended operation 
under a renewed license.

VIII. Processes for Changes and Departures

A. Tier 1 information

    1. Generic changes to Tier 1 information are governed by the 
requirements in Sec.  52.63(a)(1).
    2. Generic changes to Tier 1 information are applicable to all 
applicants or licensees who reference this appendix, except those 
for which the change has been rendered technically irrelevant by 
action taken under paragraphs A.3 or A.4 of this section.
    3. Departures from Tier 1 information that are required by the 
Commission through plant-specific orders are governed by the 
requirements in Sec.  52.63(a)(4).
    4. Exemptions from Tier 1 information are governed by the 
requirements in Sec. Sec.  52.63(b)(1) and 52.98(f). The Commission 
will deny a request for an exemption from Tier 1, if it finds that 
the design change will result in a significant decrease in the level 
of safety otherwise provided by the design.

B. Tier 2 information

    1. Generic changes to Tier 2 information are governed by the 
requirements in 10 CFR 52.63(a)(1).
    2. Generic changes to Tier 2 information are applicable to all 
applicants or licensees who reference this appendix, except those 
for which the change has been rendered technically irrelevant by 
action taken under paragraphs B.3, B.4, B.5, or B.6 of this section.
    3. The Commission may not require new requirements on Tier 2 
information by plant-specific order while this appendix is in effect 
under 10 CFR 52.55 or 52.61, unless:
    a. A modification is necessary to secure compliance with the 
Commission's regulations applicable and in effect at the time this 
appendix was approved, as set forth in Section V of this appendix, 
or to ensure adequate protection of the public health and safety or 
the common defense and security; and
    b. Special circumstances as defined in 10 CFR 50.12(a) are 
present.
    4. An applicant or licensee who references this appendix may 
request an exemption from Tier 2 information. The Commission may 
grant such a request only if it determines that the exemption will 
comply with the requirements of 10 CFR 50.12(a). The Commission will 
deny a request for an exemption from Tier 2, if it finds that the 
design change will result in a significant decrease in the level of 
safety otherwise provided by the design. The grant of an exemption 
to an applicant must be subject to litigation in the same manner as 
other issues material to the license hearing. The grant of an 
exemption to a licensee must be subject to an opportunity for a 
hearing in the same manner as license amendments.
    5.a. An applicant or licensee who references this appendix may 
depart from Tier 2 information, without prior NRC approval, unless 
the proposed departure involves a change to or departure from Tier 1 
information, Tier 2* information, or the TS, or requires a license 
amendment under paragraph B.5.b or B.5.c of this section. When 
evaluating the proposed departure, an applicant or licensee shall 
consider all matters described in the plant-specific DCD.
    b. A proposed departure from Tier 2, other than one affecting 
resolution of a severe accident issue identified in the plant-
specific DCD or one affecting information required by Sec.  
52.47(a)(28) to address aircraft impacts, requires a license 
amendment if it would:
    (1) Result in more than a minimal increase in the frequency of 
occurrence of an accident previously evaluated in the plant-specific 
DCD;
    (2) Result in more than a minimal increase in the likelihood of 
occurrence of a malfunction of a structure, system, or component 
(SSC) important to safety and previously evaluated in the plant-
specific DCD;
    (3) Result in more than a minimal increase in the consequences 
of an accident previously evaluated in the plant-specific DCD;
    (4) Result in more than a minimal increase in the consequences 
of a malfunction of an SSC important to safety previously evaluated 
in the plant-specific DCD;
    (5) Create a possibility for an accident of a different type 
than any evaluated previously in the plant-specific DCD;
    (6) Create a possibility for a malfunction of an SSC important 
to safety with a different result than any evaluated previously in 
the plant-specific DCD;
    (7) Result in a design-basis limit for a fission product barrier 
as described in the plant-specific DCD being exceeded or altered; or
    (8) Result in a departure from a method of evaluation described 
in the plant-specific DCD used in establishing the design bases or 
in the safety analyses.
    c. A proposed departure from Tier 2 affecting resolution of an 
ex-vessel severe accident design feature identified in the plant-
specific DCD, requires a license amendment if:
    (1) There is a substantial increase in the probability of an ex-
vessel severe accident

[[Page 61987]]

such that a particular ex-vessel severe accident previously reviewed 
and determined to be not credible could become credible; or
    (2) There is a substantial increase in the consequences to the 
public of a particular ex-vessel severe accident previously 
reviewed.
    d. A proposed departure from Tier 2 information required by 
Sec.  52.47(a)(28) to address aircraft impacts shall consider the 
effect of the changed design feature or functional capability on the 
original aircraft impact assessment required by 10 CFR 50.150(a). 
The applicant or licensee shall describe in the plant-specific DCD 
how the modified design features and functional capabilities 
continue to meet the aircraft impact assessment requirements in 10 
CFR 50.150(a)(1).
    e. If a departure requires a license amendment under paragraph 
B.5.b or B.5.c of this section, it is governed by 10 CFR 50.90.
    f. A departure from Tier 2 information that is made under 
paragraph B.5 of this section does not require an exemption from 
this appendix.
    g. A party to an adjudicatory proceeding for either the 
issuance, amendment, or renewal of a license or for operation under 
Sec.  52.103(a), who believes that an applicant or licensee who 
references this appendix has not complied with paragraph VIII.B.5 of 
this appendix when departing from Tier 2 information, may petition 
to admit into the proceeding such a contention. In addition to 
compliance with the general requirements of 10 CFR 2.309, the 
petition must demonstrate that the departure does not comply with 
paragraph VIII.B.5 of this appendix. Further, the petition must 
demonstrate that the change bears on an asserted noncompliance with 
an ITAAC acceptance criterion in the case of a Sec.  52.103 
preoperational hearing, or that the change bears directly on the 
amendment request in the case of a hearing on a license amendment. 
Any other party may file a response. If, on the basis of the 
petition and any response, the presiding officer determines that a 
sufficient showing has been made, the presiding officer shall 
certify the matter directly to the Commission for determination of 
the admissibility of the contention. The Commission may admit such a 
contention if it determines the petition raises a genuine issue of 
material fact regarding compliance with paragraph VIII.B.5 of this 
appendix.
    6.a. An applicant who references this appendix may not depart 
from Tier 2* information, which is designated with italicized text 
or brackets and an asterisk in the generic DCD, without NRC 
approval. The departure will not be considered a resolved issue, 
within the meaning of Section VI of this appendix and Sec.  
52.63(a)(5).
    b. A licensee who references this appendix may not depart from 
the following Tier 2* matters without prior NRC approval. A request 
for a departure will be treated as a request for a license amendment 
under 10 CFR 50.90.
    (1) Fuel mechanical and thermal-mechanical design evaluation 
reports, including fuel burnup limits.
    (2) Control rod mechanical and nuclear design reports.
    (3) Fuel nuclear design report.
    (4) Critical power correlation.
    (5) Fuel licensing acceptance criteria.
    (6) Control rod licensing acceptance criteria.
    (7) Mechanical and structural design of spent fuel storage 
racks.
    (8) Steam dryer pressure load analysis methodology.
    c. A licensee who references this appendix may not, before the 
plant first achieves full power following the finding required by 
Sec.  52.103(g), depart from the following Tier 2* matters except 
under paragraph B.6.b of this section. After the plant first 
achieves full power, the following Tier 2* matters revert to Tier 2 
status and are subject to the departure provisions in paragraph B.5 
of this section.
    (1) ASME Boiler and Pressure Vessel Code, Section III, 
Subsections NE (Division 1) and CC (Division 2) for containment 
vessel design.
    (2) American Concrete Institute 349 and American National 
Standards Institute/American Institute of Steel Construction--N690.
    (3) Power-operated valves.
    (4) Equipment seismic qualification methods.
    (5) Piping design acceptance criteria.
    (6) Instrument setpoint methodology.
    (7) Safety-Related Distribution Control and Information System 
performance specification and architecture.
    (8) Safety System Logic and Control hardware and software.
    (9) Human factors engineering design and implementation.
    (10) First of a kind testing for reactor stability (first plant 
only).
    (11) Reactor precritical heatup with reactor water cleanup/
shutdown cooling (first plant only).
    (12) Isolation condenser system heatup and steady state 
operation (first plant only).
    (13) Power maneuvering in the feedwater temperature operating 
domain (first plant only).
    (14) Load maneuvering capability (first plant only).
    (15) Defense-in-depth stability solution evaluation test (first 
plant only).
    d. Departures from Tier 2* information that are made under 
paragraph B.6 of this section do not require an exemption from this 
appendix.
    C. Operational requirements.
    1. Generic changes to generic TS and other operational 
requirements that were completely reviewed and approved in the 
design certification rulemaking and do not require a change to a 
design feature in the generic DCD are governed by the requirements 
in 10 CFR 50.109. Generic changes that require a change to a design 
feature in the generic DCD are governed by the requirements in 
paragraphs A or B of this section.
    2. Generic changes to generic TS and other operational 
requirements are applicable to all applicants who reference this 
appendix, except those for which the change has been rendered 
technically irrelevant by action taken under paragraphs C.3 or C.4 
of this section.
    3. The Commission may require plant-specific departures on 
generic TS and other operational requirements that were completely 
reviewed and approved, provided a change to a design feature in the 
generic DCD is not required and special circumstances as defined in 
10 CFR 2.335 are present. The Commission may modify or supplement 
generic TS and other operational requirements that were not 
completely reviewed and approved or require additional TS and other 
operational requirements on a plant-specific basis, provided a 
change to a design feature in the generic DCD is not required.
    4. An applicant who references this appendix may request an 
exemption from the generic TS or other operational requirements. The 
Commission may grant such a request only if it determines that the 
exemption will comply with the requirements of Sec.  52.7. The grant 
of an exemption must be subject to litigation in the same manner as 
other issues material to the license hearing.
    5. A party to an adjudicatory proceeding for the issuance, 
amendment, or renewal of a license, or for operation under Sec.  
52.103(a), who believes that an operational requirement approved in 
the DCD or a TS derived from the generic TS must be changed may 
petition to admit such a contention into the proceeding. The 
petition must comply with the general requirements of 10 CFR 2.309 
and must demonstrate why special circumstances as defined in 10 CFR 
2.335 are present, or demonstrate compliance with the Commission's 
regulations in effect at the time this appendix was approved, as set 
forth in Section V of this appendix. Any other party may file a 
response to the petition. If, on the basis of the petition and any 
response, the presiding officer determines that a sufficient showing 
has been made, the presiding officer shall certify the matter 
directly to the Commission for determination of the admissibility of 
the contention. All other issues with respect to the plant-specific 
TS or other operational requirements are subject to a hearing as 
part of the license proceeding.
    6. After issuance of a license, the generic TS have no further 
effect on the plant-specific TS. Changes to the plant-specific TS 
will be treated as license amendments under 10 CFR 50.90.

IX. [Reserved]

X. Records and Reporting

A. Records

    1. The applicant for this appendix shall maintain a copy of the 
generic DCD that includes all generic changes it makes to Tier 1 and 
Tier 2, and the generic TS and other operational requirements. The 
applicant shall maintain the sensitive unclassified non-safeguards 
information (including proprietary information and security-related 
information) and safeguards information referenced in the generic 
DCD for the period that this appendix may be referenced, as 
specified in Section VII of this appendix.
    2. An applicant or licensee who references this appendix shall 
maintain the plant-specific DCD to accurately reflect both generic 
changes to the generic DCD and plant-specific departures made under 
Section

[[Page 61988]]

VIII of this appendix throughout the period of application and for 
the term of the license (including any period of renewal).
    3. An applicant or licensee who references this appendix shall 
prepare and maintain written evaluations that provide the bases for 
the determinations required by Section VIII of this appendix. These 
evaluations must be retained throughout the period of application 
and for the term of the license (including any period of renewal).
    4.a. The applicant for the ESBWR design shall maintain a copy of 
the aircraft impact assessment performed to comply with the 
requirements of 10 CFR 50.150(a) for the term of the certification 
(including any period of renewal).
    b. An applicant or licensee who references this appendix shall 
maintain a copy of the aircraft impact assessment performed to 
comply with the requirements of 10 CFR 50.150(a) throughout the 
pendency of the application and for the term of the license 
(including any period of renewal).

B. Reporting

    1. An applicant or licensee who references this appendix shall 
submit a report to the NRC containing a brief description of any 
plant-specific departures from the DCD, including a summary of the 
evaluation of each. This report must be filed in accordance with the 
filing requirements applicable to reports in Sec.  52.3.
    2. An applicant or licensee who references this appendix shall 
submit updates to its plant-specific DCD that reflect the generic 
changes to and plant-specific departures from the generic DCD made 
under Section VIII of this appendix. These updates shall be filed 
under the filing requirements applicable to final safety analysis 
report updates in 10 CFR 52.3 and 50.71(e).
    3. The reports and updates required by paragraphs X.B.1 and 
X.B.2 of this appendix must be submitted as follows:
    a. On the date that an application for a license referencing 
this appendix is submitted, the application must include the report 
and any updates to the generic DCD.
    b. During the interval from the date of application for a 
license to the date the Commission makes its finding required by 
Sec.  52.103(g), the report must be submitted semi-annually. Updates 
to the plant-specific DCD must be submitted annually and may be 
submitted along with amendments to the application.
    c. After the Commission makes the finding required by Sec.  
52.103(g), the reports and updates to the plant-specific DCD must be 
submitted, along with updates to the site-specific portion of the 
final safety analysis report for the facility, at the intervals 
required by 10 CFR 50.59(d)(2) and 50.71(e)(4), respectively, or at 
shorter intervals as specified in the license.

    Dated at Rockville, Maryland, this 6th day of October, 2014.
    For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
[FR Doc. 2014-24362 Filed 10-14-14; 8:45 am]
BILLING CODE 7590-01-P