61944 Federal Register / Vol. 79, No. 199 / Wednesday, October 15, 2014 / Rules and Regulations 

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. 

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: Carol.Gallagher@nrc.gov. For 

technical questions, contact the 

individuals listed in the FOR FURTHER

INFORMATION CONTACT section of this 

document. 

• NRC’s Agencywide Documents 

Access and Management System 

(ADAMS): You may obtain publicly 

available documents online in the 

ADAMS Public Documents collection at 

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 pdr.resource@nrc.gov. For the 

convenience of the reader, instructions 

about obtaining materials referenced in 

this document are provided in a table in 

Section VII, ‘‘Availability of 

Documents,’’ of this document. 

• NRC’s PDR: You may examine and 

purchase copies of public documents at 

the NRC’s PDR, Room O1–F21, One 

White Flint North, 11555 Rockville 

Pike, Rockville, Maryland 20852. 

FOR FURTHER INFORMATION CONTACT: 

George M. Tartal, Office of New 

Reactors, telephone: 301–415–0016, 

email: George.Tartal@nrc.gov; or David 

Misenhimer, Office of New Reactors, 

telephone: 301–415–6590, email: 

David.Misenhimer@nrc.gov; U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001. 

SUPPLEMENTARY INFORMATION: 

Executive Summary 

A. Need for the Regulatory Action 

The NRC is amending its regulations 

related to licenses, certifications, and 

approvals for nuclear power plants. This 

final rule certifies the ESBWR standard 

plant design. This action is necessary so 

that applicants or licensees intending to 

construct and operate an ESBWR design 

may do so by referencing this DCR. 

B. Major Provisions 

Major provisions of the final rule 

include changes to: 

• specify which documents contain 

the requirements for the ESBWR design, 

• specify how a nuclear power plant 

license applicant can reference the 

ESBWR design, 

• describe how the NRC considers 

matters within the scope of the design 

to be resolved for proceedings involving 

a license or application referencing the 

ESBWR design, and 

• describe the processes for changes 

to and departures from the ESBWR 

design. 

C. Costs and Benefits 

The NRC did not prepare a regulatory 

analysis to determine the expected 

quantitative or qualitative costs and 

benefits of the final rule. The NRC 

prepares regulatory analyses for 

rulemakings that establish generic 

regulatory requirements applicable to all 

licensees. Design certifications are not 

generic rulemakings in the sense that 

design certifications do not establish 

standards or requirements with which 

all licensees must comply. Rather, 

design certifications are NRC approvals 

of specific nuclear power plant designs 

by rulemaking, which then may be 

voluntarily referenced by an applicant 

for a combined license (COL). 

Furthermore, design certification 

rulemakings are initiated by an 

applicant for a design certification, 

rather than the NRC. Preparation of a 

regulatory analysis in this circumstance 

would not be useful because the design 

to be certified is proposed by the 

applicant rather than the NRC. For these 

reasons, the NRC concludes that 

preparation of a regulatory analysis is 

neither required nor appropriate. 

Table of Contents 

I. Background 

II. Summary and Analysis of Public 

Comments on the ESBWR Proposed Rule 

and Supplemental Proposed Rule 

A. Overview of Public Comments 

B. Comments Regarding Technical Content 

in the Design Control Document 

C. Comments Regarding NRC’s Response to 

Fukushima Dai-ichi Accident 

III. Regulatory and Policy Issues 

A. How the ESBWR Design Addresses 

Fukushima Near Term Task Force 

(NTTF) Recommendations 

B. Incorporation by Reference of Public 

Documents and Issue Resolution 

Associated With Non-Public Documents 

C. Changes to Tier 2* Information 

D. Change Control for Severe Accident 

Design Features 

E. Access to Safeguards Information (SGI) 

and Sensitive Unclassified NonSafeguards Information (SUNSI) 

F. Human Factors Engineering (HFE) 

Operational Program Elements Exclusion 

From Finality 

G. Other Changes to the ESBWR Rule 

Language and Difference Between the 

ESBWR Rule and Other DCRs 

IV. Technical Issues 

A. Regulatory Treatment of Nonsafety 

Systems (RTNSS) 

B. Containment Performance 

C. Control Room Cooling 

D. Feedwater Temperature Operating 

Domain 

E. Steam Dryer Analysis Methodology 

F. Aircraft Impact Assessment (AIA) 

G. American Society of Mechanical 

Engineers (ASME) Code Case N–782 

H. Exemption for the Safety Parameter 

Display System 

I. Hurricane-Generated Winds and Missiles 

J. Loss of One or More Phases of Offsite 

Power 

K. Spent Fuel Assembly Integrity in Spent 

Fuel Racks 

L. Turbine Building Offgas System Design 

Requirements 

M. ASME Boiler and Pressure Vessel Code 

(BPV Code) Statement in Chapter 1 of the 

ESBWR Design Control Document (DCD) 

N. Clarification of ASME Component 

Design Inspections, Tests, Analyses, and 

Acceptance Criteria (ITAACs) 

O. Corrections, Editorial, and Conforming 

Changes 

V. Rulemaking Procedure 

A. Exclusions From Issue Finality and 

Issue Resolution for Spent Fuel Pool 

Instrumentation 

B. Incorporation by Reference of Public 

Documents 

C. Changes to Tier 2* Information 

D. Other Changes to the ESBWR Rule 

Language and Difference From Other 

DCRs 

E. Exclusions From Issue Finality and Issue 

Resolution for Hurricane-Generated 

Winds and Missiles 


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F. Loss of One or More Phases of Offsite 

Power 

G. Spent Fuel Assembly Integrity in Spent 

Fuel Racks 

H. Turbine Building Offgas System Design 

Requirements 

I. ASME BPV Code Statement in Chapter 

1 of the ESBWR DCD 

J. Clarification of ASME Component Design 

Inspections, Tests, Analyses, and 

Acceptance Criteria (ITAACs) 

K. Changes to the Supplemental FSER 

After Publication of the Supplemental 

Proposed Rule 

L. Corrections, Editorial, and Conforming 

Changes 

VI. Planned Withdrawal of the ESBWR 

Standard Design Approval (SDA) 

VII. Section-by-Section Analysis 

A. Introduction (Section I) 

B. Definitions (Section II) 

C. Scope and Contents (Section III) 

D. Additional Requirements and 

Restrictions (Section IV) 

E. Applicable Regulations (Section V) 

F. Issue Resolution (Section VI) 

G. Duration of This Appendix (Section VII) 

H. Processes for Changes and Departures 

(Section VIII) 

I. Inspections, Tests, Analyses, and 

Acceptance Criteria (Section IX) 

J. Records and Reporting (Section X) 

VIII. Agreement State Compatibility 

IX. Availability of Documents 

X. Voluntary Consensus Standards 

XI. Finding of No Significant Environmental 

Impact: Availability 

XII. Paperwork Reduction Act 

XIII. Regulatory Analysis 

XIV. Regulatory Flexibility Certification 

XV. Backfitting and Issue Finality 

XVI. Congressional Review Act 

XVII. Plain Writing 

XVIII. Availability of Guidance 

I. Background 

Part 52 of Title 10 of the Code of 

Federal Regulations (10 CFR), 

‘‘Licenses, Certifications, and Approvals 

for Nuclear Power Plants,’’ subpart B, 

presents the process for obtaining 

standard design certifications. On 

August 24, 2005, GEH tendered its 

application for certification of the 

ESBWR standard plant design (ADAMS 

Accession No. ML052450245) with the 

NRC. The NRC published a notice of 

receipt of the application in the Federal 

Register (70 FR 56745; September 28, 

2005). GEH submitted this application 

in accordance with subpart B of 10 CFR 

part 52. On December 1, 2005, the NRC 

formally accepted the application as a 

docketed application for design 

certification (Docket No. 52–010) (70 FR 

73311; December 9, 2005). The preapplication 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 selfcharacterized as ‘‘petitions’’ or comment 

submissions related to such ‘‘petitions’’ 

(petitions). 

The NRC received four unique 

comment submissions, including three 

comment submissions from private 

citizens and one comment submission 

from a non-government organization. 

Table 1 provides summary information 

on the unique comment submissions 

and their ADAMS Accession numbers. 

In addition, in light of the Fukushima 

Dai-ichi accident and during the public 

comment period on the proposed rule, 

the NRC received a series of petitions to 

suspend adjudicatory, licensing, and 


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rulemaking activities, including the 

ESBWR design certification rulemaking. 

The NRC subsequently authorized 

responsive and supplemental filings on 

these petitions. In its Memorandum and 

Order, CLI–11–05, September 9, 2011, 

74 NRC 141 (2011) (this decision is 

available on the NRC Web site in 

Volume 74 at 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

Comment 

submission No. Commenter ADAMS Accession No. 

1 ........................ Paul Daugherty .................................................................................................................................................... ML110880057 

2 ........................ Farouk Baxter ...................................................................................................................................................... ML110880315 

3 ........................ Patricia T. Birnie, Chairman, General Electric Stockholders’ Alliance ................................................................ ML11158A088 

4 ........................ Anonymous .......................................................................................................................................................... ML11187A303 

TABLE 2—COMMENT SUBMISSIONS SELF-CHARACTERIZED AS PETITIONS AND RESPONSIVE FILINGS

Comment 

submission No. Commenter ADAMS Accession No. 

1 (Note 1) .......... Various organizations and individuals ................................................................................................................. ML111040472 

2 (Note 1) .......... Various organizations and individuals ................................................................................................................. ML111080855 

3 ........................ Various organizations and individuals ................................................................................................................. ML111100618 

4 ........................ Jerald G. Head, Senior VP, Regulatory Affairs, GE Hitachi Nuclear Energy ..................................................... ML11124A103 

5 ........................ Various organizations and individuals ................................................................................................................. ML111260637 

6 ........................ ESBWR Intervenors ............................................................................................................................................. ML112430118 

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, safetyrelated battery design, control rod drive 

design, and control room flood 

protection. Comments regarding 

Fukushima lessons learned include 

delaying certification of the ESBWR 

design until lessons learned have been 

incorporated and the NRC’s obligation 

under the National Environmental 

Policy Act (NEPA) to evaluate new 

information (such as the NTTF report, 

ADAMS Accession No. ML111861807) 

relevant to the environmental impact of 

its actions prior to certifying the ESBWR 

design. The NRC received comments 

related to the draft EA for this rule but 

those comments did not include 

anything to suggest that: (i) A rule 

certifying the ESBWR standard design 

would be a major Federal action, or (ii) 

the severe accident mitigation design 

alternatives (SAMDA) evaluation 

omitted a design alternative that should 

have been considered or incorrectly 

considered the costs and benefits of the 

alternatives it did consider. Therefore, 

no change to the EA was warranted. The 

NRC received no comments on the two 

specific topics in the supplemental 

proposed rule. The detailed comment 

summaries and the NRC’s responses are 

provided in Sections II.B and II.C of this 

document. 

Comment Identification Format 

All comments are identified uniquely 

by using the format [W][X]–[Y], where: 

[W] represents the comment 

submission type (S = unique comment 

submission, P = petition). 

[X] represents the comment 

submission identification number (refer 

to the comment submission tables). 

[Y] represents the comment number, 

which the NRC assigned to the 

comment. In some instances, lower-case 

alphabetic characters [Ya, Yb, Yc * * *] 

were added to a comment number after 

the initial designation of comments. 

The NRC has created a document 

(ADAMS Accession No. ML113130141) 

which compiles all comment 

submissions and annotates each 

comment submission with the comment 

number indicated in the right hand 

margin. 

B. Comments Regarding Technical 

Content in the DCD 

Design-Basis Accidents 

Comment: Beyond-Design-Basis 

Accidents (DBAs) should be included in 

the design, final safety analysis report 

(FSAR), and Technical Specifications 

(TS). (S1–1) 

NRC Response: The NRC agrees that 

beyond-DBAs should be considered in 

the ESBWR design and the FSAR. In its 

1985 policy statement on severe 

accidents (50 FR 32138), the 

Commission defined the term ‘‘severe 

accident’’ as an event that is ‘‘beyond 


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the substantial coverage of design basis 

events,’’ (DBE) including events in 

which there is substantial damage to the 

reactor core (whether or not there are 

serious offsite consequences). 

Consistent with the objectives of 

standardization and early resolution of 

design issues, 10 CFR 52.47(a)(23) 

requires applicants for design 

certification to include a description 

and analysis of severe accident 

prevention and mitigation features in 

the new reactor designs. These features 

are discussed in Chapter 19 of the DCD 

(equivalent to an FSAR), and the staff’s 

evaluation of them is found in Chapter 

19 of the FSER. 

The NRC disagrees that beyond-DBAs 

should be included in the TS. The TS 

prescribe safety limits, limiting safety 

system settings, limiting conditions for 

operation, surveillance requirements, 

and administrative controls associated 

with DBEs, but need not prescribe limits 

or settings for conditions that could be 

experienced during a beyond-DBE. 

No change was made to the rule, the 

DCD, or the EA as a result of this 

comment. 

Comment: The NRC’s current 

regulatory scheme requires significant 

re-evaluation and revision in order to 

expand or upgrade the design-basis for 

reactor safety as recommended by its 

NTTF report. (P6–1) 

NRC Response: The NRC considers 

this comment to be outside the scope of 

the ESBWR design certification 

rulemaking. The comment deals with 

the adequacy of the NRC’s overall 

regulatory scheme for nuclear power 

reactors and does not directly address 

the adequacy of the ESBWR design 

certification. 

Nonetheless, the NRC disagrees with 

the comment. The NRC’s rules and 

regulations provide reasonable 

assurance of adequate protection of 

public health and safety and the 

common defense and security. However, 

the Commission has ‘‘initiated a 

comprehensive examination of the 

implications of the Fukushima 

accident. . . . As a result [of that 

examination], the NRC may implement 

changes to its regulations and regulatory 

processes.’’ CLI–11–05, 74 NRC at 168. 

If such changes are warranted, the 

NRC’s ‘‘regulatory processes provide 

sufficient time and avenues to ensure 

that design certifications and COLs 

satisfy any Commission-directed 

changes before any new power plant 

commences operations. . . . Whether 

[the Commission] adopt[s] the Task 

Force recommendations or require[s] 

more, or different, actions associated 

with certified designs or COL 

applications, [the Commission has] the 

authority to ensure that certified designs 

and combined licenses include 

appropriate Commission-directed 

changes before operation.’’ Id. at 162– 

163. 

No change was made to the rule, the 

DCD, or the EA as a result of this 

comment. 

Comment: The ESBWR environmental 

documents do not address the 

radiological consequences of DBAs or 

demonstrate that those reactors can be 

operated without undue risk to the 

health and safety of the public and 

conclude that any health effects 

resulting from the DBAs are negligible. 

This conclusion is based on a review of 

the DBAs considered in the ESBWR DCD 

(WEC 2008) and NUREG–0800, 

Standard Review Plan (SRP). The 

findings of the Fukushima NTTF report 

call into question whether this 

represents a full, accurate description 

and examination of all DBAs having the 

potential for releases to the 

environment. See Makhijani Declaration 

at 7. If the design-basis for the reactors 

does not incorporate accidents that 

should be considered in order to satisfy 

the adequate protection standard, then 

it is not possible to reach a conclusion 

that the design of the reactor adequately 

protects against accident risks. See 

Makhijani Declaration at 9. (P6–3) 

NRC Response: The NRC disagrees 

with this comment. The NRC notes that 

the Makhijani Declaration citations do 

not address DBAs as discussed in the 

comment, but rather the declaration 

specifically refers to beyond-DBEs. The 

NRC interprets the comment to be 

referring to the environmental report 

required to be provided by the design 

certification applicant per 10 CFR 52.47, 

‘‘Contents of applications; technical 

information,’’ and 10 CFR 51.55, 

‘‘Environmental report—standard 

design certification.’’ The 

environmental report (NEDO–33306; 

ADAMS Accession No. ML102990433) 

referenced in Chapter 19 of the ESBWR 

DCD and evaluated in Chapter 19 of the 

FSER, as well as the NRC’s EA, 

addresses costs and benefits of severe 

accident mitigation design alternatives. 

Conversely, DBAs for the ESBWR, and 

their associated radiological 

consequences, are not addressed in the 

environmental report but rather are 

addressed in Chapter 15 of the ESBWR 

DCD and evaluated in Chapter 15 of the 

FSER. The environmental report 

addresses the costs and benefits of 

severe accident mitigation design 

alternatives but does not address the 

design basis accidents discussed in the 

comment. In any event, the Commission 

has stated that, if warranted and after ‘‘a 

comprehensive examination of the 

implications of the Fukushima accident 

. . ., the NRC may implement changes 

to its regulations and regulatory 

processes.’’ CLI–11–05, 74 NRC at 168. 

The NRC’s ‘‘regulatory processes 

provide sufficient time and avenues to 

ensure that design certifications and 

COLs satisfy any Commission-directed 

changes before any new power plant 

commences operations. . . .’’ Id. at 162– 

163. 

No change was made to the rule, the 

DCD, or the EA as a result of this 

comment. 

Electrical Systems 

Comment: The ESBWR design is 

flawed because it has failed to comply 

with the requirements of Institute of 

Electrical and Electronics Engineers 

(IEEE) Standard 603, which requires the 

electrical portion of the safety systems 

that perform safety functions— 

specifically, alternating current (ac) 

power from the Ancillary Diesel 

Generators (ADGs)—be classified as 

Class 1E. The DCD acknowledges that 

ac power from the ADGs is not needed 

for the first 72 hours of an accident, but 

are needed to perform Class 1E 

functions (recharging the Class 1E direct 

current (dc) batteries that provide power 

during the first 72 hours of an accident) 

when no other sources of power are 

available. The ESBWR design has 

classified these ac power sources as 

commercial grade, nonsafety-related, 

and non-Class 1E (S2–1, referencing 

ADAMS Accession No. ML102350160). 

NRC Response: The NRC disagrees 

with the comment. The NRC’s position 

remains as stated in the separate 

correspondence between the commenter 

and the NRC that is attached to the 

comment letter. Specifically, the NRC 

stated that the events described in the 

commenter’s previous letters (no ac 

power available to the plant for 72 hours 

after initiation of the accident and all 

batteries are depleted) are not DBEs but 

are beyond design-basis, for which the 

requirements of IEEE Standard 603 do 

not apply. As stated in the staff 

requirements memorandum (SRM), 

dated January 15, 1997, concerning 

SECY–96–128, ‘‘Policy and Key 

Technical Issues Pertaining to the 

Westinghouse AP600 Standardized 

Passive Reactor Design,’’ dated June 12, 

1996, the Commission approved Item 

IV—Post-72 Hour Actions. The approval 

specified that the post-72 hour systems, 

structures, and components (SSCs) are 

not required to be safety-related. In 

addition, as stated in NUREG–1242, 

Volume 3, Part 1, ‘‘NRC Review of 

Electric Power Research Institute’s 

Advanced Light Water Reactor Utility 

Requirements Document: Passive Plant 


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Designs, Chapter 1,’’ August 1994, a 

passive advanced light-water reactor, 

such as the ESBWR design, need not 

include or rely upon an active safetyrelated 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 longterm core cooling and containment 

integrity functions. These nonsafetyrelated 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 safetyrelated 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 windgenerated 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 vendorspecified 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 designbasis flood and water stops at 

construction joints to prevent inleakage. 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 designbasis 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 


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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 safetyrelated 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 safetyrelated 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. Safetyrelated 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, nonsafetyrelated 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 safetyrelated 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 safetyrelated 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 


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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 vendorspecified 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 


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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 ‘‘decisionmaking 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 


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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. 


tkelley on DSK3SPTVN1PROD with RULES2
Federal Register / Vol. 79, No. 199 / Wednesday, October 15, 2014 / Rules and Regulations 61953 

1Some 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. 

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 nonsafeguards 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


tkelley on DSK3SPTVN1PROD with RULES2
61954 Federal Register / Vol. 79, No. 199 / Wednesday, October 15, 2014 / Rules and Regulations 

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 To ̄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 


tkelley on DSK3SPTVN1PROD with RULES2
Federal Register / Vol. 79, No. 199 / Wednesday, October 15, 2014 / Rules and Regulations 61955 

2For 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. 

3For purposes of this discussion, ‘‘securityrelated information’’ means information subject to 

non-disclosure under 10 CFR 2.390(a)(7)(vi). 

4The non-publicly available documents contain 

proprietary, security-related, and/or safeguards 

information. 

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 safetyrelated 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). 

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 nonpublic documents to support the public 

commenting process and transparency 

has been a long-standing practice for 


tkelley on DSK3SPTVN1PROD with RULES2
61956 Federal Register / Vol. 79, No. 199 / Wednesday, October 15, 2014 / Rules and Regulations 

both design certification rulemakings 

and licensing actions. 

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 nonpublic 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 publiclyavailable 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

Document No. Document title 

Publiclyavailable 

ADAMS 

Accession No. 

Non-publicly available 

ADAMS Accession No. 

NEDE–33391, NEDO–33391 .. GE Hitachi Nuclear Energy, ‘‘ESBWR Safeguards Assessment Report,’’ NEDE–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.

ML14093A138 N/A (Safeguards information 

cannot be placed in 

ADAMS) 

NEDC–31959P, NEDO–31959 GE Nuclear Energy, ‘‘Fuel Rod Thermal-Mechanical Analysis 

Methodology (GSTRM),’’ NEDC–31959P (Proprietary), 

April 1991, and NEDO–31959 (Non-proprietary), April 1991.

ML14093A145 ML14093A146 

NEDC–32992P–A, NEDO– 

32992–A.

GE Nuclear Energy, J.S. 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.

ML14093A250 ML012610605 

NEDC–33139P–A, NEDO–

33139–A.

Global Nuclear Fuel, ‘‘Cladding Creep Collapse,’’ NEDC– 

33139P–A, Class III (Proprietary), July 2005, and NEDO– 

33139–A, Class I (Non-proprietary), July 2005.

ML14094A227 ML14094A228 

NEDE–31758P–A, NEDO– 

31758–A.

GE Nuclear Energy, ‘‘GE Marathon Control Rod Assembly,’’ 

NEDE–31758P–A (Proprietary), October 1991, and 

NEDO–31758–A (Non-proprietary), October 1991.

ML14093A142 ML14093A143 

NEDC–32084P–A, NEDO– 

32084–A.

GE Nuclear Energy, ‘‘TASC–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.

ML100220484 ML100220485 

NEDC–32601 P–A, NEDO– 

32601–A.

GE Nuclear Energy, ‘‘Methodology and Uncertainties for 

Safety Limit MCPR Evaluations,’’ NEDC–32601P–A, Class 

III (Proprietary), and NEDO–32601–A, Class I (Non-proprietary), August 1999.

ML14093A216 ML003740145 

NEDC–32983P–A, NEDO– 

32983–A.

GE Nuclear Energy, ‘‘GE 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.

ML072480121 ML072480125 

NEDC–33075P–A, NEDO– 

33075–A.

GE Hitachi Nuclear Energy, ‘‘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.

ML080310396 ML080310402 

NEDC–33079P, NEDO–33079 GE Nuclear Energy, ‘‘ESBWR 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.

ML053460471 ML051390233 


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Federal Register / Vol. 79, No. 199 / Wednesday, October 15, 2014 / Rules and Regulations 61957 

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—Continued 

Document No. Document title 

Publiclyavailable 

ADAMS 

Accession No. 

Non-publicly available 

ADAMS Accession No. 

NEDC–33083P–A, NEDO– 

33083–A.

GE Nuclear Energy, ‘‘TRACG 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.

ML102770606 ML102770608 

NEDC–33237P–A, NEDO– 

33237–A.

Global Nuclear Fuel, ‘‘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.

ML102770246 ML102770244 

NEDC–33238P, NEDO–33238 Global Nuclear Fuel, ‘‘GE14 Pressure Drop Characteristics,’’ 

NEDC–33238P, Class III (Proprietary), and NEDO–33238, 

Class I (Non-proprietary), December 2005.

ML060050328 ML060050330 

NEDC–33239P–A, NEDO– 

33239P–A.

Global Nuclear Fuel, ‘‘GE14 for ESBWR Nuclear Design Report,’’ NEDC–33239P–A, Class III (Proprietary), and 

NEDO–33239–A, Class I (Non-proprietary), Revision 5, 

October 2010.

ML102800405 ML102800408 (part 1) 

ML102800425 (part 2) 

NEDC–33240P–A, NEDO– 

33240–A.

Global Nuclear Fuel, ‘‘GE14E Fuel Assembly Mechanical Design Report,’’ NEDC–33240P–A, Revision 1, Class III (Proprietary), and NEDO–33240–A, Revision 1, Class I (Nonproprietary), September 2010.ML102770060 ML102770061 

NEDC–33242P–A, NEDO– 

33242–A.

Global Nuclear Fuel, ‘‘GE14 for ESBWR Fuel Rod ThermalMechanical Design Report,’’ NEDC–33242P–A, Revision 2, 

Class III (Proprietary), and NEDO–33242–A, Revision 2, 

Class I (Non-proprietary), September 2010.

ML102730885 ML102730886 

NEDC–33326P–A, NEDO– 

33326–A.

Global Nuclear Fuel, ‘‘GE14E for ESBWR Initial Core Nuclear Design Report,’’ NEDC–33326P–A, Revision 1, Class 

III (Proprietary), and NEDO–33326–A, Revision 1, Class I 

(Non-proprietary), September 2010.

ML102740191 ML102740193 (part 1) 

ML102740194 (part 2) 

NEDC–33374P–A, NEDO– 

33374–A.

GE-Hitachi Nuclear Energy, ‘‘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.

ML102860687 ML102860688 

NEDC–33456P, NEDO–33456 Global Nuclear Fuel, ‘‘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.

ML090920867 ML090920868 

NEDE–10958–PA, NEDO– 

10958–A.

General Electric Company, ‘‘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.

ML102290144 ML092820214 

NEDE–24011–P–A–16, 

NEDO–24011–A–16.

Global Nuclear Fuel, ‘‘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.

ML091340077 ML091340081 

NEDE–24011–P–A–US–16, 

NEDO–24011–A–US–16.

Global Nuclear Fuel, ‘‘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.

ML091340080 ML091340082 

NEDE–30130–P–A, NEDO–

30130–A.

General Electric Company, ‘‘Steady State Nuclear Methods,’’ 

NEDE–30130–P–A, Class III (Proprietary), April 1985, and 

NEDO–30130–A, Class I (Non-proprietary), May 1985.

ML14104A064 ML070400570 

NEDE–31152P, NEDO–31152 Global Nuclear Fuel, ‘‘Global Nuclear Fuels Fuel Bundle Designs,’’ NEDE–31152P, Revision 9, Class III (Proprietary), 

May 2007, and NEDO–33152, Revision 9, Class I (Nonproprietary), May 2007.

ML071510287 ML071510289 

NEDE–32176P, NEDO–32176 GE Hitachi Nuclear Energy, 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.

ML080370271 ML080370276 


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61958 Federal Register / Vol. 79, No. 199 / Wednesday, October 15, 2014 / Rules and Regulations 

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—Continued 

Document No. Document title 

Publiclyavailable 

ADAMS 

Accession No. 

Non-publicly available 

ADAMS Accession No. 

NEDE–33083 Supplement 1P– 

A, NEDO–33083 Supplement 1–A.

GE Hitachi Nuclear Energy, 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.

ML102770552 ML102770550 

NEDE–33083 Supplement 2P– 

A, NEDO–33083 Supplement 2–A.

GE Hitachi Nuclear Energy, ‘‘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.

ML103000353 ML103000355 

NEDE–33083 Supplement 3P– 

A, NEDO–33083 Supplement 3–A.

GE Hitachi Nuclear Energy, ‘‘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.

ML102770606 ML102770608 

NEDE–33197P–A, NEDO– 

33197–A.

GE Hitachi Nuclear Energy, ‘‘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.

ML102810320 ML102810341 

NEDE–33217P, NEDO–33217 GE Hitachi Nuclear Energy, ‘‘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.

ML100480284 ML100480285 

NEDE–33220P, NEDO–33220 GE Hitachi Nuclear Energy, ‘‘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.

ML100480209 ML100480202 

NEDE–33221P, NEDO–33221 GE Hitachi Nuclear Energy, ‘‘ESBWR Human Factors Engineering Task Analysis Implementation Plan,’’ NEDE– 

33221P, Class III (Proprietary), and NEDO–33221, Class I 

(Non-proprietary), Revision 4, February 2010.

ML100480212 ML100480213 

NEDE–33226P, NEDO–33226 GE Hitachi Nuclear Energy, ‘‘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.

ML100550837 ML100550844 

NEDE–33243P–A, NEDO– 

33243–A.

GE Hitachi Nuclear Energy, ‘‘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.

ML102740171 ML102740178 

NEDE–33244P–A, NEDO– 

33244–A.

GE Hitachi Nuclear Energy, ‘‘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.

ML102770208 ML102770209 

NEDE–33245P, NEDO–33245 GE Hitachi Nuclear Energy, ‘‘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.

ML100550839 ML100550847 

NEDE–33259P–A, NEDO– 

33259–A.

GE Hitachi Nuclear Energy, ‘‘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.

ML102920241 ML102920248 

NEDE–33261P, NEDO–33261 GE Hitachi Nuclear Energy, ‘‘ESBWR Containment Load 

Definition,’’ NEDE–33261P, Class III (Proprietary), and 

NEDO–33261, Class I (Non-proprietary), Revision 2, June 

2008.

ML082600720 ML082600721 

NEDE–33268P, NEDO–33268 GE Hitachi Nuclear Energy, ‘‘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.

ML100480179 ML100480180 


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Federal Register / Vol. 79, No. 199 / Wednesday, October 15, 2014 / Rules and Regulations 61959 

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—Continued 

Document No. Document title 

Publiclyavailable 

ADAMS 

Accession No. 

Non-publicly available 

ADAMS Accession No. 

NEDE–33276P, NEDO–33276 GE Hitachi Nuclear Energy, ‘‘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.

ML100480182 ML100480183 

NEDE–33295P, NEDO–33295 GE Hitachi Nuclear Energy, ‘‘ESBWR Cyber Security Program Plan,’’ NEDE–33295P, Class III (Proprietary), Revision 2, September 2010, and NEDO–33295, Class I (Nonproprietary), Revision 2, September 2010.ML102880103 ML102880104 

NEDE–33304P, NEDO–33304 GE Hitachi Nuclear Energy, ‘‘GEH ESBWR Setpoint Methodology,’’ NEDE–33304P, Class III (Proprietary), and NEDO– 

33304, Class I (Non-proprietary), Revision 4, May 2010.

ML101450251 ML101450253 

NEDE–33312P, NEDO–33312 GE Hitachi Nuclear Energy, ‘‘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.

ML13344B157 ML13344B163 

NEDE–33313P, NEDO–33313 GE Hitachi Nuclear Energy, ‘‘ESBWR Steam Dryer Structural 

Evaluation,’’ NEDE–33313P, Class III (Proprietary), Revision 5, December 2013, and NEDO–33313, Class I (Nonproprietary), Revision 5, December 2013.

ML13344B158 ML13344B164 

NEDE–33408P, NEDO–33408 GE Hitachi Nuclear Energy, ‘‘ESBWR Steam Dryer—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.

ML13344B159 ML13344B176 (part 1) 

ML13344B175 (part 2) 

NEDE–33440P, NEDO–33440 GE Hitachi Nuclear Energy ‘‘ESBWR Safety Analysis—Additional Information,’’ NEDE–33440P, Class III (Proprietary), 

and NEDO–33440, Class I (Non-proprietary), Revision 2, 

March 2010.

ML100920316 ML100920317 (part 1) 

ML100920318 (part 2) 

NEDE–33516P–A, NEDO– 

33516–A.

GE Hitachi Nuclear Energy, ‘‘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.

ML102880499 ML102880500 

NEDE–33536P, NEDO–33536 GE Hitachi Nuclear Energy, ‘‘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.

ML102780329 ML102780330 

NEDE–33572P, NEDO–33572 GE Hitachi Nuclear Energy, ‘‘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.

ML102740579 ML102740566 

Letter w/attachment ................. Letter from R.J. Reda (GE) 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.

ML14093A140 ML14094A240 

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 redesignation 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) pressureretaining 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 


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61960 Federal Register / Vol. 79, No. 199 / Wednesday, October 15, 2014 / Rules and Regulations 

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 NonSafeguards 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 


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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) Hurricanegenerated 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 defensein-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 


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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, ‘‘HumanSystem 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. 


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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 nonpublic 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. 


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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). 


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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 designbasis 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 


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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 hurricanegenerated 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 hurricanegenerated 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-WaterCooled 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 


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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, 

‘‘§ 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 § 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 asbuilt 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 


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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 publiclyavailable 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 publiclyavailable 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 


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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 HurricaneGenerated 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 hurricanegenerated 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 


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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, plantspecific departures from the generic 

DCD that the Commission imposed on 

the licensee by order, and any plantspecific 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 plantspecific 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 plantspecific 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 twotiered 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 designrelated 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 nonsystem 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 


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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 designrelated 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 firstthree-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. 


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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 nonsafetyrelated 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 sitespecific 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 plantspecific DCD to either include or 

incorporate by reference the generic 

DCD information. Further, this 

paragraph also requires the plantspecific 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 


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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 hurricanegenerated missile loads on those SSCs 

described in Section 3.5.2 of the generic 

DCD are either bounded by tornadogenerated missile loads analyzed in 

Section 3.5.1.4 of the generic DCD or 

will meet applicable NRC requirements 

with consideration of hurricanegenerated 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 securityrelated 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 caseby-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 


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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 securityrelated 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 


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5Certain 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. 

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 plantspecific 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. 

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 postfuel 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 securityrelated 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 plantspecific 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 plantspecific departures as ‘‘change 

processes.’’ 


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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 plantspecific 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 Commissionordered Tier 2 departures do not have 

to outweigh any decrease in safety that 

may result from the reduction in 

standardization caused by the plantspecific 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 


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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 


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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 plantspecific 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 plantspecific 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 plantspecific 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 plantspecific 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. 


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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 securityrelated 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 securityrelated 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 securityrelated 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 plantspecific 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 sitespecific 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 plantspecific 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 


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once a year, this should not be an 

excessive burden on the applicant. 

Paragraph X.B.3.b also requires semiannual 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. 

Document 

ADAMS 

Accession No./ 

web 

link/ Federal 

Register citation 

Proposed Rule Documents: 

SECY–11–0006, ‘‘Proposed Rule—ESBWR Design Certification’’ .......................................................................................... ML102220172 

Staff Requirements Memorandum for SECY–11–0006, ‘‘Proposed Rule—ESBWR Design Certification’’ ............................. ML110670047 

General Electric Company Application for Final Design Approval and Design Certification of ESBWR Standard Plant Design.ML052450245 

ESBWR Design Control Document, Revision 9 ........................................................................................................................ ML103440266 

ESBWR Final Safety Evaluation Report (NUREG–1966) ......................................................................................................... ML14100A304 

ESBWR FSER Final Chapters .................................................................................................................................................. ML103470210 

Final Design Approval for the Economic Simplified Boiling Water Reactor ............................................................................. ML110540310 

ESBWR Draft Environmental Assessment ................................................................................................................................ ML102220247 

ESBWR Proposed Rule Federal Register Notice, 76 FR 16549, March 24, 2011 ................................................................ ML110610353 

Public Comments on the March 2011 Proposed Rule: 

Comment (1) from Farouk D. Baxter on Environmental Impact Statement for Two AP1000 Units at Levy County Site ........ ML102350160

Comment submission S1 from Paul C. Daugherty ................................................................................................................... ML110880057 

Comment submission S2 from Farouk D. Baxter ..................................................................................................................... ML110880315 

Comment submission S3 from Patricia T. Birnie, Chair, General Electric Stockholders’ Alliance ........................................... ML11158A088 

Comment submission S4 from anonymous .............................................................................................................................. ML11187A303 

Comment submission P1, 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 

(initial).

ML111040472 

Comment submission P2, 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 

(amended).

ML111080855 

Comment submission P3, Declaration of Dr. 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.

ML111100618 

Comment submission P4, Comment of Jerald 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.

ML11124A103 

Comment submission P5, Petitioners’ Reply 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.

ML111260637 

Comment submission P6, Comments of Terry J. Lodge on PR 52, NEPA Requirement To Address Safety and Environmental Implications of the Fukushima Task Force Report From ESBWR, Fermi 3 Intervenors.ML112430118 

Public Comments Compilation—Final Rule—ESBWR Design Certification (RIN 3150–AI85) ................................................ ML113130141 

Supplemental Safety Evaluation for the ESBWR Design Certification: 

Advanced Supplemental Safety Evaluation Report for the Economic Simplified Boiling-Water Reactor Standard Plant Design.ML14043A134 

Supplemental Safety Evaluation Report for the Economic Simplified Boiling-Water Reactor Standard Plant Design ............ ML14155A333 

Supplemental Proposed Rule Documents: 

ESBWR Design Control Document, Rev. 10 ............................................................................................................................ ML14104A929 

ESBWR Supplemental Proposed Rule Federal Register Notice, 79 FR 25715, May 6, 2014 .............................................. ML14043A508 

Final Rule Documents: 

SECY–14–0081, ‘‘Final Rule—ESBWR Design Certification’’ .................................................................................................. ML111730346 

Staff Requirements Memorandum for SECY–14–0081, ‘‘Final Rule—ESBWR Design Certification’’ ..................................... ML14259A545 

ESBWR Final Environmental Assessment ................................................................................................................................ ML111730382 

Other Documents Relevant to the ESBWR Rulemaking: 

NEDO–33306, Revision 4, ‘‘ESBWR Severe Accident Mitigation Design Alternatives’’ .......................................................... ML102990433 

NEDO–33312, Rev. 5, ‘‘ESBWR Steam Dryer Acoustic Load Definition’’ ............................................................................... ML13344B157 


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Document 

ADAMS 

Accession No./ 

web 

link/ Federal 

Register citation 

NEDO–33313, Rev. 5, ‘‘ESBWR Steam Dryer Structural Evaluation’’ ..................................................................................... ML13344B158 

NEDO–33338, Revision 1, ‘‘ESBWR Feedwater Temperature Operating Domain Transient and Accident Analysis’’ ........... ML091380173 

NEDO–33408P, Revision 5, ‘‘ESBWR Steam Dryer—Plant-Based Load Evaluation Methodology, PBLE01 Model Description’’.ML13344B159 

Commission Memorandum and Order (CLI–11–05), September 9, 2011 (available on the NRC Web site in Volume 74 at 

http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr0750/).

ML112521106 

Commission Order, ‘‘Scheduling Order of the Secretary Regarding Petitions To Suspend Adjudicatory, Licensing and 

Rulemaking Activities (PR 52 re ESBWR Design Certification)’’.

ML111101277 

Order EA–12–049, ‘‘Order Modifying Licenses With Regard to Requirements for Mitigation Strategies for Beyond-DesignBasis External Events’’.ML12054A735 

Order EA 12–051, ‘‘Order Modifying Licenses With Regard to Reliable Spent Fuel Pool Instrumentation’’ ........................... ML12054A679 

Staff Requirements Memorandum for SECY–90–377, ‘‘Requirements for Design Certification Under 10 CFR Part 52’’ ...... ML003707892 

SECY–94–084, ‘‘Policy and Technical Issues Associated With the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs’’.ML003708068 

Staff Requirements Memorandum for SECY–96–077, ‘‘Certification of Two Evolutionary Designs’’ ...................................... ML003754873 

SECY–96–077, ‘‘Certification of Two Evolutionary Designs’’ ................................................................................................... ML003708129 

Staff Requirements Memorandum for SECY–11–0093, ‘‘Near-Team Report and Recommendations for Agency Actions 

Following the Events in Japan’’.

ML112310021 

SECY–11–0093, ‘‘Enclosure: The Near-Term Task Force Review of Insights From the Fukushima Dai-ichi Accident’’ ........ ML111861807 

Staff Requirements Memorandum for SECY–11–0117, ‘‘Proposed Charter for the Longer-Term Review of Lessons 

Learned From the March 11, 2011, Japanese Earthquake and Tsunami’’.

ML112920034 

SECY–11–0117, ‘‘Proposed Charter for the Longer-Term Review of Lessons Learned From the March 11, 2011, Japanese Earthquake and Tsunami’’.ML11231A723 

SECY–11–0124, ‘‘Recommended Actions To Be Taken Without Delay From The Near-Term Task Force Report’’ ............. ML11245A127 

SECY–11–0137, ‘‘Prioritization of Recommended Actions To Be Taken in Response to Fukushima Lessons Learned’’ ..... ML11269A204 

Staff Requirements Memorandum for SECY–12–0025, ‘‘Proposed Orders and Requests for Information in Response to 

Lessons Learned From Japan’s March 11, 2011, Great To ̄hoku Earthquake and Tsunami’’.

ML120690347 

SECY–12–0025, ‘‘Proposed Orders and Requests for Information in Response to Lessons Learned From Japan’s March 

11, 2011, Great To ̄hoku Earthquake and Tsunami’’.

ML12039A103 

SECY–14–0046, ‘‘Fifth 6-Month Status Update on Response to Lessons Learned From Japan’s March 11, 2011, Great 

To ̄hoku Earthquake and Subsequent Tsunami’’.

ML14064A523 

Regulatory Guide 1.13, ‘‘Spent Fuel Storage Facility Design Basis’’ ....................................................................................... ML070310035 

Regulatory Guide 1.20, ‘‘Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and 

Initial Startup Testing’’.

ML070260376 

Regulatory Guide 1.27, ‘‘Ultimate Heat Sink for Nuclear Power Plants (for Comment)’’ ......................................................... ML003739996 

Regulatory Guide 1.76, ‘‘Design-Basis Tornado and Tornado Missiles for Nuclear Power Plants’’ ........................................ ML070360253 

Regulatory Guide 1.117, ‘‘Tornado Design Classification’’ ....................................................................................................... ML003739346 

Regulatory Guide 1.143, ‘‘Design Guidance for Radioactive Waste Management Systems, Structures, and Components 

Installed in Light-Water-Cooled Nuclear Power Plants’’.

ML003740200 

Regulatory Guide 1.206, Section C.I.1, ‘‘Standard Format and Content of Combined License Applications—Introduction 

and General Description of the Plant’’.

ML070630005 

Regulatory Guide 1.221, ‘‘Design-Basis Hurricane and Hurricane Missiles for Nuclear Power Plants’’ ................................. ML110940303 

NUREG–0700, Revision 2, ‘‘Human-Systems Interface Design Review Guidelines’’ (three volumes) ................................... ML021700337 

ML021700342 

ML021700371 

NUREG–0711, Revision 2, ‘‘Human Factors Engineering Program Review Model’’ ............................................................... ML040770540 

NUREG–0711, Revision 3, ‘‘Human Factors Engineering Program Review Model’’ ............................................................... ML12324A013 

NUREG–0800, Section 3.8.4, Revision 2, ‘‘Other Seismic Category I Structures,’’ Appendix D, ‘‘Guidance on Spent Fuel 

Pool Racks’’.

ML070550054 

NUREG–0800, Section 3.9.2, Revision 3, ‘‘Dynamic Testing and Analysis of Systems, Structures, and Components’’ ....... ML070230008 

NUREG–0800, Section 3.9.5, Revision 3, ‘‘Reactor .................................................................................................................

Pressure Vessel Internals’’ ........................................................................................................................................................

ML070230009 

NUREG–0800, SRP Section 6.4, Revision 3, ‘‘Control Room Habitability System’’ ................................................................ ML070550069 

NUREG–0800, SRP Section 9.1.2, Revision 4, ‘‘New and Spent Fuel Storage’’ .................................................................... ML070550057 

NUREG–0800, SRP Section 13.4, Revision 3, ‘‘Operational Programs’’ ................................................................................. ML070470463 

NUREG–0800, SRP Section 13.5.2.1, Revision 2, ‘‘Operating and Emergency Operating Procedures’’ ............................... ML070100635 

NUREG–0800, SRP Section 18, Revision 2, ‘‘Human Factors Engineering’’ .......................................................................... ML070670253 

NUREG–1242, ‘‘NRC Review of Electric Power Research Institute’s Advanced Light Water Reactor Utility Requirements 

Document, Evolutionary Plant Designs’’ (five volumes).

ML100610048 

ML100430013 

ML063620331 

ML070600372 

ML070600373 

NRC Bulletin 2012–01: Design Vulnerability in Electric Power System ................................................................................... ML12074A115 

Interim Staff Guidance DC/COL–ISG–8, ‘‘Necessary Content of Plant-Specific Technical Specifications’’ ............................ ML083310259 

JLD–ISG–2012–03 Revision 0, ‘‘Compliance With Order EA–12–051, Reliable Spent Fuel Pool Instrumentation,’’ ............. ML12221A339 

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’’.ML122400399 

‘‘Clarifications Requested by NRC Staff on Economic Simplified Boiling Water Reactor Fuel Design’’ .................................. ML11269A093 

Audit Report, ‘‘ESBWR Fuel Seismic Audit Summary’’ ............................................................................................................ ML112860614 


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6The 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. 

Document 

ADAMS 

Accession No./ 

web 

link/ Federal 

Register citation 

Notice of Violation, ‘‘ESBWR AIA Inspection Report Inspection, NRC Inspection Report No. 0520000/10/2010–201 and 

Notice of Violation’’.

ML102740292 

Reply to Notice of Violation, NRC Inspection Report 052000010–10–201 .............................................................................. ML103010047 

GE–Hitachi Nuclear Energy Americas, LLC, Reply to Notice of Violation, NRC IR 052000010–10–201 ............................... ML103400150 

ACRS Memorandum—Final Rule—ESBWR Design Certification (RIN 3150–AI85) ................................................................ ML113120076 

ACRS Memorandum—ESBWR Design Certification Rulemaking and Supplemental Final Safety Evaluation Report ........... ML11340A043 

ACRS Memorandum—Supplemental Final Safety Evaluation Report on the General Electric-Hitachi Nuclear Energy 

(GEH) Application for Certification of the Economic Simplified Boiling Water Reactor (ESBWR) Design.

ML14107A263 

ACRS Memorandum—Final Rule—ESBWR Design Certification (RIN 3150–AI85) ................................................................ ML14196A207 

Regulatory History of Design Certification 6 .............................................................................................................................. ML003761550 

X. Voluntary Consensus Standards 

The National Technology Transfer 

and Advancement Act of 1995 (Act), 

Pub. L. 104–113, requires that Federal 

agencies use technical standards that are 

developed or adopted by voluntary 

consensus standards bodies unless the 

use of such a standard is inconsistent 

with applicable law or otherwise 

impractical. In this final rule, the NRC 

is approving the ESBWR standard plant 

design for use in nuclear power plant 

licensing under 10 CFR part 50 or part 

52. Design certifications are not generic 

rulemakings establishing a generally 

applicable standard with which all 10 

CFR parts 50 and 52 nuclear power 

plant licensees or applicants for SDAs, 

design certifications, or manufacturing 

licenses must comply. Design 

certifications are NRC approvals of 

specific nuclear power plant designs by 

rulemaking. Furthermore, design 

certifications are initiated by an 

applicant for rulemaking, rather than by 

the NRC. For these reasons, the NRC 

concludes that the Act does not apply 

to this final rule. 

XI. Finding of No Significant 

Environmental Impact: Availability 

The NRC has determined under 

NEPA, and the NRC’s regulations in 

subpart A, ‘‘National Environmental 

Policy Act; Regulations Implementing 

Section 102(2),’’ of 10 CFR part 51, 

‘‘Environmental Protection Regulations 

for Domestic Licensing and Related 

Regulatory Functions,’’ that this DCR is 

not a major Federal action significantly 

affecting the quality of the human 

environment and, therefore, an 

environmental impact statement (EIS) is 

not required. The NRC’s generic 

determination in this regard is reflected 

in 10 CFR 51.32(b)(1). The basis for the 

NRC’s categorical exclusion in this 

regard, as discussed in the 2007 final 

rule amending 10 CFR parts 51 and 52 

(August 28, 2007; 72 FR 49352–49566), 

is based upon the following 

considerations. A DCR does not 

authorize the siting, construction, or 

operation of a facility referencing any 

particular design; it only codifies the 

ESBWR design in a rule. The NRC will 

evaluate the environmental impacts and 

issue an EIS as appropriate under NEPA 

as part of the application for the 

construction and operation of a facility 

referencing any particular DCR. 

In addition, consistent with 10 CFR 

51.30(d) and 10 CFR 51.32(b), the NRC 

has prepared a final EA (ADAMS 

Accession No. ML111730382) for the 

ESBWR design addressing various 

design alternatives to prevent and 

mitigate severe accidents. The EA is 

based, in part, upon the NRC’s review 

of GEH’s evaluation of various design 

alternatives to prevent and mitigate 

severe accidents in NEDO–33306, 

Revision 4, ‘‘ESBWR Severe Accident 

Mitigation Design Alternatives.’’ Based 

upon review of GEH’s evaluation, the 

Commission concludes that: (1) GEH 

identified a reasonably complete set of 

potential design alternatives to prevent 

and mitigate severe accidents for the 

ESBWR design; (2) none of the potential 

design alternatives are justified on the 

basis of cost-benefit considerations; and 

(3) it is unlikely that other design 

changes would be identified and 

justified during the term of the design 

certification on the basis of cost-benefit 

considerations because the estimated 

core damage frequencies for the ESBWR 

are very low on an absolute scale. These 

issues are considered resolved for the 

ESBWR design. 

The NRC requested comments on the 

draft EA but the comments received did 

not include anything to suggest that: (i) 

A rule certifying the ESBWR standard 

design would be a major Federal action, 

or (ii) the SAMDA evaluation omitted a 

design alternative that should have been 

considered or incorrectly considered the 

costs and benefits of the alternatives it 

did consider. Therefore, no change to 

the EA was warranted. All 

environmental issues concerning 

SAMDAs associated with the 

information in the final EA and NEDO– 

33306 are considered resolved for 

facility applications referencing the 

ESBWR design if the site characteristics 

at the site proposed in the facility 

application fall within the site 

parameters specified in NEDO–33306. 

The final EA, upon which the 

Commission’s finding of no significant 

impact is based, and the ESBWR DCD 

are available for examination and 

copying at the NRC’s PDR, One White 

Flint North, Room O–1 F21, 11555 

Rockville Pike, Rockville, Maryland 

20852. 

XII. Paperwork Reduction Act 

This rule contains new or amended 

information collection requirements that 

are subject to the Paperwork Reduction 

Act of 1995 (44 U.S.C. 3501, et seq.). 

These requirements were approved by 

the Office of Management and Budget 

(OMB), control number 3150–0151. The 

burden to the public for these 

information collections is estimated to 

average 15 hours per response. 

Send comments on any aspect of 

these information collections, including 

suggestions for reducing the burden, to 

the Records and FOIA/Privacy Services 

Branch (T–5 F52), U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001, or by Internet 

electronic mail to 

INFOCOLLECTS.RESOURCE@

NRC.GOV; and to the Desk Officer, 

Office of Information and Regulatory 

Affairs, NEOB–10202, (3150–0151), 

Office of Management and Budget, 

Washington, DC 20503. 

Public Protection Notification 

The NRC may not conduct or sponsor, 

and a person is not required to respond 


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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 

■ 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). 

■ 2. In § 52.11, paragraph (b) is revised 

to read as follows: 

§ 52.11 Information collection 

requirements: OMB approval. 

* * * * * 

(b) The approved information 

collection requirements contained in 

this part appear in §§ 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. 

■ 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 plantspecific changes to generic DCD information. 

D. Tier 1 means the portion of the designrelated 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; 


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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 designrelated 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 §§ 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 PDR.Resource@nrc.gov. 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: Library.Resource@nrc.gov. All 

approved material is available for inspection 

at the National Archives and Records 

Administration (NARA). For information on 

the availability of this material at NARA, call 

1–202–741–6030 or go to 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, Revision 10, Tier 1, dated April 2014 .... ML14104A929 (package) 

26A6642AB Rev. 10 ..................... ESBWR Design Control Document, Revision 10, Tier 2, dated April 2014 .... ML14104A929 (package) 

Bechtel Power Corporation: 

BC–TOP–3–A ............................... ‘‘Tornado and Extreme Wind Design Criteria for Nuclear Power Plants,’’ 

Topical Report, Revision 3, August 1974.

ML14093A218 

BC–TOP–9A ................................. ‘‘Design of Structures for Missile Impact,’’ Topical Report, Revision 2, September 1974.ML14093A217 

General Electric: 

GEZ–4982A .................................. General Electric Large Steam Turbine Generator Quality Control Program, 

The STG Global Supply Chain Quality Management System (MFGGLO– 

GEZ–0010) Revision 1.2, February 7, 2006.

ML14093A215 

GE Nuclear Energy: 

NEDO–11209–04A ....................... ‘‘GE Nuclear Energy Quality Assurance Program Description,’’ Class 1, Revision 8, March 31, 1989.ML14093A209 

NEDO–31960–A ........................... ‘‘BWR Owners’ Group Long-Term Stability Solutions Licensing Methodology,’’ Class I, November 1995.ML14093A212 

NEDO–31960–A—Supplement 1 ‘‘BWR Owners’ Group Long-Term Stability Solutions Licensing Methodology,’’ Class I, November 1995.ML14093A211 

NEDO–32465–A ........................... GE Nuclear Energy and BWR Owners’ Group, ‘‘Reactor Stability Detect and 

Suppress Solutions Licensing Basis Methodology for Reload Applications,’’ Class I, August 1996.

ML14093A210 

GE-Hitachi Nuclear Energy: 

NEDO–33181 ................................ ‘‘NP–2010 COL Demonstration Project Quality Assurance Plan,’’ Revision 6, 

August 2009.

ML14248A297 

NEDO–33219 ................................ ‘‘ESBWR Human Factors Engineering Functional Requirements Analysis 

Implementation Plan,’’ Revision 4, Class I, February 2010.

ML100350104 

NEDO–33260 ................................ ‘‘Quality Assurance Requirements for Suppliers of Equipment and Services 

to the GEH ESBWR Project,’’ Revision 5, Class I, April 2008.

ML14248A648 

NEDO–33262 ................................ ‘‘ESBWR Human Factors Engineering Operating Experience Review Implementation Plan,’’ Revision 3, Class I, January 2010.ML100340030 

NEDO–33266 ................................ ‘‘ESBWR Human Factors Engineering Staffing and Qualifications Implementation Plan,’’ Revision 3, Class I, January 2010.ML100350167 


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TABLE 1—DOCUMENTS APPROVED FOR INCORPORATION BY REFERENCE—Continued 

Document No. Document title ADAMS Accession No. 

NEDO–33267 ................................ ‘‘ESBWR Human Factors Engineering Human Reliability Analysis Implementation Plan,’’ Revision 4, Class I, January 2010.ML100330609 

NEDO–33277 ................................ ‘‘ESBWR Human Factors Engineering Human Performance Monitoring Implementation Plan,’’ Revision 4, Class I, January 2010.ML100270770 

NEDO–33278 ................................ ‘‘ESBWR Human Factors Engineering Design Implementation Plan,’’ Revision 4, Class I, January 2010.ML100270468 

NEDO–33289 ................................ ‘‘ESBWR Reliability Assurance Program,’’ Revision 2, Class II, September 

2008.

ML14248A662 

NEDO–33337 ................................ ‘‘ESBWR Initial Core Transient Analyses,’’ Revision 1, Class I, April 2009 .... ML091130628 

NEDO–33338 ................................ ‘‘ESBWR Feedwater Temperature Operating Domain Transient and Accident Analysis,’’ Revision 1, Class I, May 2009.ML091380173 

NEDO–33373–A ........................... ‘‘Dynamic, Load-Drop, and Thermal-Hydraulic Analyses for ESBWR Fuel 

Racks,’’ Revision 5, Class I, October 2010.

ML102990226 (part 1) 

ML102990228 (part 2) 

NEDO–33411 ................................ ‘‘Risk Significance of Structures, Systems and Components for the Design 

Phase of the ESBWR,’’ Revision 2, Class I, February 2010.

ML100610417 

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 §§ 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 § 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 tornadogenerated 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 GEHitachi 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 § 52.63(a)(5) in subsequent 

proceedings for issuance of a COL, 

amendment of a COL, or renewal of a COL, 

proceedings held under § 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 


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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 

§ 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 § 52.85, the hearing provided 

under § 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 §§ 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 

§ 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 § 52.63(a)(4). 

4. Exemptions from Tier 1 information are 

governed by the requirements in 

§§ 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 plantspecific 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 

§ 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 


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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 exvessel severe accident previously reviewed. 

d. A proposed departure from Tier 2 

information required by § 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 

§ 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 § 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 § 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 thermalmechanical 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 

§ 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 plantspecific 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 § 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 § 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 plantspecific 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 nonsafeguards 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 plantspecific DCD to accurately reflect both 

generic changes to the generic DCD and 

plant-specific departures made under Section 


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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 § 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 

§ 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 § 52.103(g), the reports and 

updates to the plant-specific DCD must be 

submitted, along with updates to the sitespecific 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 


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