[Federal Register Volume 79, Number 199 (Wednesday, October 15, 2014)]
[Rules and Regulations]
[Pages 61944-61988]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2014-24362]
[[Page 61943]]
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
[[Page 61944]]
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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 http://www.regulations.gov and search for Docket ID NRC-2010-0135. Address
questions about NRC dockets to Carol Gallagher, telephone: 301-287-
3422; email: [email protected]. 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 http://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 [email protected]. 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: [email protected]; or
David Misenhimer, Office of New Reactors, telephone: 301-415-6590,
email: [email protected]; 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
[[Page 61945]]
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
[[Page 61946]]
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 http://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
[[Page 61947]]
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.
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\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 http://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
[email protected]; 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 http://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 [email protected]. These documents can also be viewed at the
Federal rulemaking Web site, http://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: [email protected]. 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 http://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