Economic Simplified Boiling-Water Reactor Design Certification

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 Non-Safeguards Information (SUNSI)

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

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

IV. Technical Issues

A. Regulatory Treatment of Nonsafety Systems (RTNSS)

B. Containment Performance

C. Control Room Cooling

D. Feedwater Temperature Operating Domain

E. Steam Dryer Analysis Methodology

F. Aircraft Impact Assessment (AIA)

G. American Society of Mechanical Engineers (ASME) Code Case N-782

H. Exemption for the Safety Parameter Display System

I. Hurricane-Generated Winds and Missiles

J. Loss of One or More Phases of Offsite Power

K. Spent Fuel Assembly Integrity in Spent Fuel Racks

L. Turbine Building Offgas System Design Requirements

M. ASME Boiler and Pressure Vessel Code (BPV Code) Statement in Chapter 1 of the ESBWR Design Control Document (DCD)

N. Clarification of ASME Component Design Inspections, Tests, Analyses, and Acceptance Criteria (ITAACs)

O. Corrections, Editorial, and Conforming Changes

V. Rulemaking Procedure

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

B. Incorporation by Reference of Public Documents

C. Changes to Tier 2* Information

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

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

F. Loss of One or More Phases of Offsite Power

G. Spent Fuel Assembly Integrity in Spent Fuel Racks

H. Turbine Building Offgas System Design Requirements

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

J. Clarification of ASME Component Design Inspections, Tests, Analyses, and Acceptance Criteria (ITAACs)

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

L. Corrections, Editorial, and Conforming Changes

VI. Planned Withdrawal of the ESBWR Standard Design Approval (SDA)

VII. Section-by-Section Analysis

A. Introduction (Section I)

B. Definitions (Section II)

C. Scope and Contents (Section III)

D. Additional Requirements and Restrictions (Section IV)

E. Applicable Regulations (Section V)

F. Issue Resolution (Section VI)

G. Duration of This Appendix (Section VII)

H. Processes for Changes and Departures (Section VIII)

I. Inspections, Tests, Analyses, and Acceptance Criteria (Section IX)

J. Records and Reporting (Section X)

VIII. Agreement State Compatibility

IX. Availability of Documents

X. Voluntary Consensus Standards

XI. Finding of No Significant Environmental Impact: Availability

XII. Paperwork Reduction Act

XIII. Regulatory Analysis

XIV. Regulatory Flexibility Certification

XV. Backfitting and Issue Finality

XVI. Congressional Review Act

XVII. Plain Writing

XVIII. Availability of Guidance

I. Background

Part 52 of Title 10 of the

Code of Federal Regulations

(10 CFR), “Licenses, Certifications, and Approvals for Nuclear Power Plants,” subpart B, presents the process for obtaining standard design certifications. On August 24, 2005, GEH tendered its application for certification of the ESBWR standard plant design (ADAMS Accession No. ML052450245) with the NRC. The NRC published a notice of receipt of the application in the

Federal Register

(70 FR 56745; September 28, 2005). GEH submitted this application in accordance with subpart B of 10 CFR part 52. On December 1, 2005, the NRC formally accepted the application as a docketed application for design certification (Docket No. 52-010) (70 FR 73311; December 9, 2005). The pre-application information submitted before the NRC formally accepted the application can be found in ADAMS under Docket No. PROJ0717 (Project No. 717).

The NRC staff issued a final safety evaluation report (FSER) for the ESBWR design in March 2011. The FSER is available in ADAMS under Accession No. ML103470210. The NRC subsequently published the FSER in April 2014 as NUREG-1966, “Final Safety Evaluation Report Related to the Certification of the Economic Simplified Boiling-Water Reactor Standard Design” (ADAMS Accession No. ML14100A304). The NRC also published a proposed rule to certify the ESBWR design in the

Federal Register

on March 24, 2011 (76 FR 16549), and a supplemental proposed rule on May 6, 2014 (79 FR 25715). The FSER and the proposed rule were based on the NRC's review of Revision 9 of the ESBWR DCD.

On April 17, 2014, the NRC issued an advanced supplemental safety evaluation report (SER) (ADAMS Accession No. ML14043A134) to address several matters identified by the NRC and revisions to the ESBWR DCD in Revision 10. The advanced supplemental SER was referenced in the supplemental proposed rule (79 FR 25715; May 6, 2014). The supplemental FSER will be published as Supplement No. 1 to NUREG-1966 before this final rule becomes effective. Because Revision 10 of the DCD was issued after the ESBWR proposed rule was published, all of the substantive changes in Revision 10 of the DCD are addressed in the

SUPPLEMENTARY INFORMATION

section of this document, including a discussion of why the change was or was not addressed in a supplemental proposed rule.

In its application for design certification, GEH also requested the NRC to provide an SDA for the ESBWR design. An SDA for the ESBWR design was issued in March 2011 (ADAMS Accession No. ML110540310) following the NRC staff's issuance of the ESBWR FSER. On June 3, 2014, GEH requested that the NRC retire the SDA at the time of issuance of the final ESBWR design certification rule (ADAMS Accession No. ML14154A094). After this final rule is published, the NRC intends, as a separate action from this rulemaking, to withdraw the SDA.

The application for design certification of the ESBWR design has been referenced in the following COL applications as of the date of this document: (1) Detroit Edison Company, Fermi Unit 3, Docket No. 52-033 (73 FR 73350; December 2, 2008); (2) Dominion Virginia Power, North Anna Unit 3, Docket No. 52-017 (73 FR 6528; February 4, 2008); (3) Entergy Operations, Inc., Grand Gulf Unit 3, Docket No. 52-024 (73 FR 22180; April 24, 2008) (APPLICATION SUSPENDED); (4) Entergy Operations, Inc., River Bend Unit 3, Docket No. 52-036 (73 FR 75141; December 10, 2008) (APPLICATION SUSPENDED); and (5) Exelon Nuclear Texas Holdings, LLC, Victoria County Station Units 1 and 2, Docket Nos. 52-031 and 52-032 (73 FR 66059; November 6, 2008) (APPLICATION WITHDRAWN).

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

A. Overview of Public Comments

The NRC published a proposed rule to certify the ESBWR design in the

Federal Register

on March 24, 2011 (76 FR 16549). The period for submitting comments on the proposed DCR, ESBWR DCD, or draft environmental assessment (EA) closed on June 7, 2011. The NRC received a total of 10 public comments on the proposed rule. The types of comments, the organization of comments, the comment identification format, and comment responses follow.

The NRC also published a supplemental proposed rule to request public comments on two specific topics regarding the ESBWR design certification. The supplemental proposed rule was published in the

Federal Register

on May 6, 2014 (79 FR 25715). The period for submitting comments on these specific topics closed on June 5, 2014. The NRC received no public comments on the supplemental proposed rule.

Types of Comments

The NRC received two types of comment submissions on the proposed rule for the ESBWR design certification. A comment submission means a communication or document, submitted to the NRC by an individual or entity, with one or more individual comments addressing a subject or an issue. The two types of comment submissions were:

1. Comment submissions that were not identical or similar in content (unique comment submissions); and

2. Comment submissions self-characterized as “petitions” or comment submissions related to such “petitions” (petitions).

The NRC received four unique comment submissions, including three comment submissions from private citizens and one comment submission from a non-government organization. Table 1 provides summary information on the unique comment submissions and their ADAMS Accession numbers.

In addition, in light of the Fukushima Dai-ichi accident and during the public comment period on the proposed rule, the NRC received a series of petitions to suspend adjudicatory, licensing, and

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, safety-related battery design, control rod drive design, and control room flood protection. Comments regarding Fukushima lessons learned include delaying certification of the ESBWR design until lessons learned have been incorporated and the NRC's obligation under the National Environmental Policy Act (NEPA) to evaluate new information (such as the NTTF report, ADAMS Accession No. ML111861807) relevant to the environmental impact of its actions prior to certifying the ESBWR design. The NRC received comments related to the draft EA for this rule but those comments did not include anything to suggest that: (i) A rule certifying the ESBWR standard design would be a major Federal action, or (ii) the severe accident mitigation design alternatives (SAMDA) evaluation omitted a design alternative that should have been considered or incorrectly considered the costs and benefits of the alternatives it did consider. Therefore, no change to the EA was warranted. The NRC received no comments on the two specific topics in the supplemental proposed rule. The detailed comment summaries and the NRC's responses are provided in Sections II.B and II.C of this document.

Comment Identification Format

All comments are identified uniquely by using the format [W][X]-[Y], where:

[W] represents the comment submission type (S = unique comment submission, P = petition).

[X] represents the comment submission identification number (refer to the comment submission tables).

[Y] represents the comment number, which the NRC assigned to the comment. In some instances, lower-case alphabetic characters [Ya, Yb, Yc * * *] were added to a comment number after the initial designation of comments.

The NRC has created a document (ADAMS Accession No. ML113130141) which compiles all comment submissions and annotates each comment submission with the comment number indicated in the right hand margin.

B. Comments Regarding Technical Content in the DCD

Design-Basis Accidents

Comment: Beyond-Design-Basis Accidents (DBAs) should be included in the design, final safety analysis report (FSAR), and Technical Specifications (TS).

(S1-1)

NRC Response:

The NRC agrees that beyond-DBAs should be considered in the ESBWR design and the FSAR. In its 1985 policy statement on severe accidents (50 FR 32138), the Commission defined the term “severe accident” as an event that is “beyond

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

Designs, Chapter 1,” August 1994, a passive advanced light-water reactor, such as the ESBWR design, need not include or rely upon an active safety-related ac power source to support safety system functions after 72 hours from the onset of an accident, but may rely on electrical power sources that are not safety-related after that time. Specifically, the ESBWR is designed so that safety-related passive systems are able to perform all safety functions for 72 hours after initiation of a DBE without the need for operator actions. The DBE is assumed to be resolved (except for long-term cooling) within 72 hours, and thus, the Class 1E batteries are designed for and need only function for 72 hours without being recharged.

In the ESBWR, the ADGs, which are the subject of the commenter's concern, are not used to recharge the Class 1E batteries. Rather, the ADGs provide power directly to post accident monitoring instrumentation, main control room lighting, the reactor pressure vessel (RPV) makeup pump, and containment cooling systems, among others. After 72 hours, consistent with NUREG-1242, nonsafety-related systems other than the ADGs are used to replenish safety-related passive systems so that they will perform long-term core cooling and containment integrity functions. These nonsafety-related systems are designed in accordance with quality standards commensurate with the importance of these functions and that provide reasonable assurance they will function when needed. In the event that the ADGs are not available, the Seismic Category I firewater storage tanks and Seismic Category I diesel pump and fire protection piping can be used to provide post-accident makeup water to the Isolation Condenser and Passive Containment Cooling System (PCCS) pools and Spent Fuel Pool (SFP) using the Fuel and Auxiliary Plant Cooling System (FAPCS) for long-term cooling beyond 72 hours.

The NRC also stated in its May 15, 2009, letter (in the referenced document) that the offsite power system, a nonsafety-related power source, is the preferred source of power for safety-related systems at all current plants. Further, the station blackout (SBO) rule, 10 CFR 50.63, “Loss of all alternating current power,” does not require the use of safety-related alternative ac power sources to cope with an SBO. Therefore, neither of these ac power sources—offsite power or alternate ac power source—is required to be safety-related or classified as Class 1E under IEEE 603. Thus, the ADGs need not be classified as Class 1E power sources as well.

In summary, the design bases of the passive safety systems are centered on the 72-hour capability and these safety-related systems must remain functional to assure the integrity of the reactor coolant pressure boundary and the capability to shut down the reactor and maintain it in a safe shutdown condition without operator action or support from nonsafety systems for the first 72 hours following the initiation of a DBE. Beyond 72 hours, these systems must continue to remain functional to provide such assurance for the following 4 days, with allowance for operator actions and support from nonsafety SSCs consistent with NUREG-1242.

No change was made to the rule, the DCD, or the EA as a result of this comment.

Comment: The NRC should require GEH to relocate the safety-related dc batteries and their related systems above grade level so that they are not subject to external flooding. This recommendation is supported by the following points:

1. There is a fair chance of a failure of the dc supply as safety-related battery banks (Class-1E grade batteries) are housed below grade in the reactor building, as well as their electrical penetration to primary containment. In a natural disaster they may not remain watertight, as water may enter through the doors and incapacitate the battery banks.

2. Water may also enter the battery rooms if those doors are open for maintenance, testing, or replacement of cells.

3. ESBWR emergency core cooling systems (ECCS) are dependent on this dc supply. If the dc supply is lost, emergency cooling and depressurization systems will fail. There is no diversity for the core cooling and depressurization systems if the dc supply fails.

(S4-1)

NRC Response:

The NRC disagrees with the comment. The safety-related dc batteries and their related systems do not need to be relocated above grade level. The NRC has reviewed the ESBWR DCD and has determined that the ESBWR safety-related SSCs (including the reactor building, which houses the dc batteries) are designed to withstand the effects of external flooding. With the exception of loads due to hurricane winds and wind-generated missiles beyond those considered in the ESBWR DCD, the NRC concluded that the ESBWR DCD meets the requirements of 10 CFR part 50, appendix A, “General Design Criteria for Nuclear Power Plants,” (GDC) 2, which requires the design bases of SSCs important to safety to include protection against natural phenomena (including earthquakes, tornadoes, floods, hurricanes, and tsunami) such that these SSCs will not lose the capability to perform their safety functions as a result of such phenomena. This conclusion is documented in the NRC's FSER for the ESBWR design.

In the following paragraphs, the NRC addresses each of the three supporting points for the comment.

Supporting Point 1:

The NRC agrees that safety-related batteries are located below grade per the ESBWR DCD, Tier 2, Figure 1.2-2. This is acceptable because all components of safety-related dc electric systems are housed in structures which provide protection against external flood damage. The structures that may be subjected to a design-basis flood are designed to withstand the flood level by locating the plant grade elevation 1 ft. (0.30 m) above the flood level and incorporating structural provisions into the plant design to protect the SSCs from the postulated flood conditions. GEH's application for design certification was submitted with proposed vendor-specified site parameters. These values are provided in Table 2.0-1 (Tier 2) and in Table 5.1-1 (Tier 1) of the DCD. For the ESBWR design, the maximum groundwater level is 2 ft. (0.61 m) below plant grade and the maximum flood level is 1 ft. (0.30 m) below plant grade. The ESBWR design was evaluated using the vendor-specified flood levels and found to be safe. All exterior access openings are above flood level. The flood design incorporates reinforced concrete walls designed to resist the static and dynamic forces of the design-basis flood and water stops at construction joints to prevent in-leakage. External surfaces below flood and ground water levels are waterproofed. Penetrations are sealed and also capable of withstanding the static and dynamic forces of the design-basis flood. Watertight doors provide physical separation of flood zones. In addition, the applicant has specified the site parameters, design characteristics, and any additional requirements and restrictions necessary for a COL applicant to ensure that safety-related SSCs will be adequately protected from the site-specific probable maximum flood conditions. Based on the evaluation in Section 3.4 of the FSER, the NRC concludes that the ESBWR design regarding flood protection provides reasonable assurance that safety-related SSCs (including the safety-related dc batteries and their

related systems) will maintain their structural integrity or are located within structures that will maintain their integrity, and will perform their intended safety functions when subjected to a design-basis flood, and therefore, satisfy the requirements of GDC 2.

Supporting Point 2:

The comment stated that water may enter the battery rooms if the watertight doors are open for maintenance, testing, or replacement of the battery cells. The NRC agrees that this scenario is possible for one division of safety-related battery banks. The ESBWR TS, under limiting condition of operation 3.8.1, restricts maintenance, testing, or replacement of the battery cells during plant operation to only one required division of safety-related battery banks. In addition, the COL applicant is required to develop plant operating and maintenance procedures that provide control for activities that are important to the safe operation of the facility, including limiting conditions of operation. However, there are four divisions of safety-related battery banks, which are physically separated by concrete walls and watertight doors. Only two divisions of dc systems are required for safe shutdown of the plant. If one of the safety-related battery room doors is open during a flood, as suggested in the comment, the other batteries will still be adequately protected by design features for physical separation to ensure the safety-related SSCs can perform their functions.

Supporting Point 3:

The comment stated that the ESBWR ECCS is dependent on dc power, and if dc power is lost, emergency cooling and depressurization systems will fail. The ESBWR ECCS consists of the Gravity Driven Cooling System, the Isolation Condenser System, the Standby Liquid Control System, and the Automatic Depressurization System. The Gravity Driven Cooling System, Standby Liquid Control System, and the Automatic Depressurization System do rely on dc power for actuation (as pointed out in the comment). The four trains of Isolation Condenser System, on the other hand, automatically begin removal of decay heat and control RPV level above the top of active fuel upon loss of all ac and dc power because the only valve in the system relied upon to change position upon initiation of the system fails in the safe (open) position upon loss of power. Beginning 4 hours after the start of an accident, the Isolation Condenser System upper and lower header vent valves are opened periodically to remove non-condensable gases to maintain optimum heat removal and allow continued reactor cooldown. These valves are solenoid-operated valves and rely upon electric power to open.

The comment also suggests that there is no diversity for several systems that rely on the dc power supply. The NRC agrees that the Automatic Depressurization System, Gravity Driven Cooling System, the Suppression Pool Equalization Line Valves, and the Standby Liquid Control System all require safety-related dc power in order to perform their safety functions and therefore lack diversity in that regard, but does not agree that the Basemat Internal Melt Arrest Coolability (BiMAC) cooling system requires safety-related dc power to perform its safety function. As discussed below, the BiMAC cooling system—a non-safety system—is designed to automatically fire squib valves and drain water to the area below the RPV upon sensing high temperatures in the BiMAC without dependence on any of the four safety-related power sources. Also, as discussed above, the four trains of the Isolation Condenser System automatically begin removal of decay heat and control RPV level above the top of active fuel upon loss of all ac and dc power because the only valve in the system relied upon to change position upon initiation of the system fails in the safe (open) position upon loss of power. Decay heat can be removed with the Isolation Condenser System for 72 hours without any additional action. The ESBWR is designed such that the Isolation Condenser System heat exchanger pool can be replenished after 72 hours with the diesel driven fire pump to allow continued cooling with the Isolation Condenser System. Safety-related dc power is not needed to operate this pump. In light of these facts, the NRC concludes that the capability of the ESBWR to remove decay heat from the reactor core following an accident is sufficiently diverse. It should also be noted that the ESBWR safety-related 120 volts ac uninterruptible power supply (UPS) input is normally supplied by offsite power or a nonsafety-related onsite power system. During a loss of offsite and nonsafety-related onsite power, the UPS gets its power from 250 volts dc batteries. The ESBWR design includes an offsite power system, nonsafety-related standby diesel generators, and ADGs, any of which can mitigate the consequences of an accident if available. Safety-related UPS systems are housed in seismic Category I structures and meet GDCs 2, 4, and 17.

Common cause failure of the safety-related batteries in the ESBWR design would clearly be an event of substantial safety significance because dc power is used to power the distributed control and instrumentation system, which is used to actuate passive safety systems. However, the ESBWR design includes a number of defense-in-depth features for reducing the likelihood of losing all ability to accomplish key safety functions. As previously stated, the Isolation Condenser System automatically begins removal of decay heat and controls RPV level above the top of active fuel upon loss of all ac and dc power. All safety divisions (including concrete walls and watertight doors that separate the four safety-related battery banks) are physically separated.

The ESBWR design also includes design features specifically for the purpose of injecting water into the containment to flood the containment floor and cover core debris. The BiMAC cooling system is designed to automatically fire squib valves and drain water to the area below the RPV upon sensing high temperatures in the BiMAC, indicating core debris below the RPV. This occurs without operator action and without dependence on any of the four safety-related power sources.

No change was made to the rule, the DCD, or the EA as a result of this comment.

Control Rod Drive System

Comment: Two Control Rod Drives (CRD) are scrammed by one hydraulic control unit (HCU). A single failure of one HCU will affect the scram function of two CRDs. It is done for cost saving. This is not acceptable in a safety system.

(S4-2)

NRC Response:

The NRC disagrees with the comment. In Section 4.6.3 of the FSER, the NRC stated that a single failure in an HCU may result in the failure of two control rods. The DCD describes that the control rods are assigned to HCUs in a manner such that no 4X4 array of rods contain both rods connected to the same HCU. This arrangement assures that shutdown is achieved (among other things) assuming a single failure of an HCU. The NRC reviewed the effects of an HCU failure and concluded in Section 4.3 of the FSER that sufficient shutdown margin exists in the case of an HCU failure. In addition, TS 3.1.5 requires that all control rod scram accumulators are operable during Modes 1 (Power Operation) and 2 (Start-Up). If an accumulator is inoperable, the associated control rod pair is declared inoperable and Limiting Condition of Operation (LCO) 3.1.3, Control Rod Operability, is entered. This would

result in requiring the affected control rod to be fully inserted and disarmed, thereby satisfying the intended function in accordance with actions of LCO 3.1.3. If an accumulator is inoperable, TS require the affected control rod to be inserted and hence the scram function of two CRDs is satisfied. Finally, the ESBWR has a diverse method to scram the reactor. An electric motor is provided for each CRD for scram in addition to the hydraulic scram using the accumulator. Accordingly, the NRC has determined that the CRD system design is adequate.

No change was made to the rule, the DCD, or the EA as a result of this comment.

Control Room

Comment: For safety reasons, the Control Room should be located at a sufficient height from the ground to prevent its flooding during a tsunami, tornado, hurricane, heavy rain, etc.

(S4-3)

NRC Response:

The NRC agrees that the control room should be protected from flooding. GEH's application for SDA and design certification was submitted with proposed vendor-specified site parameters. The values for maximum groundwater is 2 feet (0.61 m) below plant grade as provided in Table 2.0-1 (Tier 2) of the DCD and the maximum flood level is 1 foot (0.30 m) below plant grade as provided in Table 5.1-1 (Tier 1) of the DCD.

The ESBWR design was evaluated using the vendor-specified flood levels and found to be safe. As described in Chapter 3 of the DCD, the ESBWR construction incorporates several water proofing features: The external walls below groundwater and flood levels are designed to withstand hydrostatic loads, construction and expansion joints have water stops, external surfaces below groundwater and flood levels are waterproofed, penetrations below groundwater and flood levels are sealed, and there are no exterior openings below grade.

If a COL application referencing the ESBWR design is submitted to the NRC, the COL applicant must demonstrate that the site-specific characteristics are bounded by the DCD site parameters. During the review of a COL application using this design, the staff will perform an independent analysis to verify that the flood levels and other relevant site characteristics are within the DCD parameters.

No change was made to the rule, the DCD, or the EA as a result of this comment.

Spent Fuel Pool

Comment: The ESBWR design has an elevated SFP. This is a particularly troublesome feature in common with the Mark I BWR design, which is the design of the Fukushima reactors.

(P2-2)

NRC Response:

The NRC disagrees with this comment. The ESBWR SFP design is different from the Mark I BWR design in that the ESBWR SFP is located entirely below grade. The ESBWR design does include an additional buffer pool located above grade in the reactor building. The buffer pool contains a small array of spent fuel racks that is used for temporary storage of spent fuel during refueling operations and also includes a location to store new fuel assemblies during power operations.

GDC 2 requires that the ESBWR spent fuel storage facilities (SFP and buffer pool) and the structure within which they are housed, as SSCs important to safety, be protected against the effects of natural phenomena without loss of their safety function. In addition, GDC 61 requires that the design prevents drainage of coolant inventory below an adequate shielding depth, provides adequate coolant flow to the spent fuel racks, and provides a system for detecting and containing pool liner leakage.

The reactor building and the concrete containment, which houses the SFP and additional buffer pool, are seismic Category I structures that are designed to meet the requirements of GDC 2 for protection against natural phenomena such as an earthquake, tornado, or hurricane in combination with normal and accident condition loads considering the effects due to the elevated location of the buffer pool. Information relating to the analysis and design of the reactor building is provided in DCD Sections 3.7 and 3.8 and Appendices 3A, 3B, 3F, and 3G. Through analysis and review of the design, the NRC determined that the reactor building and the concrete containment are structurally adequate to withstand all design-basis loads. The NRC concluded in the FSER that both pools are adequately protected from the effects of natural phenomena without loss of capability to perform their safety functions.

The NRC also concluded in its FSER that, because the SFP and buffer pools have anti-siphoning devices on all submerged Fuel and Auxiliary Pools Cooling System (FAPCS) piping, and there are no other drainage paths by which the level in the SFP or buffer pool could be reduced, coolant will not drain below an adequate shielding depth in either pool.

Cooling of spent fuel located in either the SFP or buffer pool is provided by the FAPCS. In the unlikely event that a loss of active cooling to the spent fuel assemblies occurs, there is enough water to keep the fuel assemblies cooled for a minimum of 72 hours before operator actions are needed. After 72 hours, additional water can be provided through safety-related connections to the fire protection system or another onsite or offsite water source. The NRC concluded in the FSER that cooling for both ESBWR SFP and buffer pools will be maintained.

Finally, the NRC concluded in the FSER that, because the spent fuel pool and buffer pool are equipped with stainless steel liners, concrete walls, and leak detection drains, both detection and containment of pool liner leakage capability are provided.

No change was made to the rule, the DCD, or the EA as a result of this comment.

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

Some commenters favored delaying (in some fashion) the ESBWR rulemaking until lessons are learned from the Fukushima Dai-ichi Nuclear Power Plant (Fukushima) accident that occurred on March 11, 2011, and the NRC applies the lessons learned to United States (U.S.) nuclear power plants, including the ESBWR design. Background on how the Commission responded to the Fukushima accident and how the ESBWR design addresses Fukushima NTTF recommendations is discussed in Section III of the

SUPPLEMENTARY INFORMATION

section of this document.

As discussed in Section III of the

SUPPLEMENTARY INFORMATION

section of this document, the NRC concludes that no changes to the ESBWR design are warranted at this time to provide reasonable assurance of adequate protection of public health and safety. Moreover, even if the Commission concludes at a later time that some additional action is needed for the ESBWR design, the NRC has ample opportunity and legal authority to modify the ESBWR DCR to implement design changes, as well as to take any necessary action to ensure that COLs that reference the ESBWR make any necessary design changes.

Comment: The NRC should suspend the certification of the ESBWR reactor design and rescind the final design approval it granted on March 9, 2011. Based on the recent events at the Fukushima Dai-ichi site, the NRC should first undertake a far more

rigorous, long-term review of the design and the regulatory implication of the events, implement new regulations to protect public health and safety, and revise the environmental analyses to evaluate the potential health, environmental and economic costs of reactor and SFP accidents.

(S3-1, P3-1, P3-2)

NRC Response:

The NRC declines to suspend the ESBWR rulemaking. See

Memorandum and Order,

CLI-11-05, 74 NRC 141 (2011) (ADAMS Accession No. ML112521106).

Background on how the Commission responded to the Fukushima accident and how the ESBWR design addresses Fukushima NTTF recommendations is discussed in Section III of the

SUPPLEMENTARY INFORMATION

section of this document. In that section, the NRC concludes that no changes to the ESBWR design are required at this time to provide reasonable assurance of adequate protection of public health and safety. If the Commission concludes at a later time that some additional action is needed for the ESBWR design, the NRC has ample opportunity and legal authority to modify the ESBWR DCR to implement design changes, as well as to take any necessary action to ensure that COLs that reference the ESBWR also make any necessary design changes.

For these reasons the NRC does not regard delays in the ESBWR design certification process to be appropriate. No change was made to the rule, the DCD, or the EA as a result of this comment.

Comment: The Atomic Energy Act (AEA) and NEPA preclude the NRC from approving standardized plant designs until it has completed the investigation of the Fukushima accident and considered the safety and environmental implications of the accident with respect to its regulatory program. NEPA imposes on agencies a continuing obligation to gather and evaluate new information relevant to the environmental impact of its actions. The need to supplement under NEPA when there is new and significant information is also found throughout the NRC regulations, e.g., 10 CFR 51.92(a)(2), 51.50(c)(iii), 51.53(b), and 51.53(c)(3)(iv). The conclusions and recommendations presented in the NTTF report constitute “new and significant information” whose environmental implications must be considered before the NRC may certify the ESBWR design and operating procedures.

(P2-2, P6-2)

NRC Response:

The NRC disagrees with this comment. The comment did not explain what particular provision of the AEA precludes the NRC from issuing a standard DCR. Furthermore, NEPA has no “continuing obligation” to gather and evaluate new information relevant to the environmental impact of its actions, because the Commission has determined that issuance of a standard DCR is not a major Federal action significantly affecting the quality of the human environment. See the EA at page 1 (ADAMS Accession No. ML111730382).

No change was made to the rule, the DCD, or the EA as a result of this comment.

Comment: The whole nuclear culture must be reviewed before any reactor designs are certified for potential construction, and that all licensing of new reactor designs be put on hold until the NRC's systems of regulations, oversight, and enforcement are thoroughly reviewed and, where required, are made more restrictive.

(S3-2)

NRC Response:

The NRC considers this comment to be outside the scope of the ESBWR design certification rulemaking. The comment addresses overall nuclear industry safety culture and does not directly address the adequacy of the ESBWR design certification.

Nonetheless, the NRC disagrees with the comment. The NRC considers that its regulatory framework and requirements provide a rigorous and comprehensive design certification and license review process that examines the full extent of siting, system design, and operations of nuclear power plants.

The NRC will continue to process existing applications for new design certifications and licenses in accordance with the schedules that have been established.

Background on how the Commission responded to the Fukushima accident and how the ESBWR design addresses Fukushima near-term task force recommendations is discussed in Section III of the

SUPPLEMENTARY INFORMATION

section of this document. In that section, the NRC concludes that no changes to the ESBWR design are warranted at this time to provide reasonable assurance of adequate protection of public health and safety. Moreover, even if the Commission concludes at a later time that some additional action is needed for the ESBWR design, the NRC has ample opportunity and legal authority to modify the ESBWR DCR to implement design changes, as well as to take any necessary action to ensure that COLs that reference the ESBWR also make any necessary design changes.

For these reasons the NRC does not regard delays in the ESBWR design certification process to be appropriate. No change was made to the rule, the DCD, or the EA as a result of this comment.

Comment: The NRC should include a review of public health challenges worldwide from radiation in its decision-making process.

(S3-3)

NRC Response:

The NRC considers this comment to be outside the scope of the ESBWR DCR. The comment addresses the NRC's generic process and criteria for regulatory decision making, and does not directly address the adequacy of the ESBWR design.

Nonetheless, the NRC disagrees with the comment. The NRC interprets the comment's reference to the “decision-making process” to mean the Commission's decision whether to certify the ESBWR design. The NRC reviewed the design and has found that it complies with the NRC's regulations, which provide reasonable assurance of adequate protection of public health and safety, including protection of the public from radiation. The comment did not provide any data, analyses, or other technical information to suggest why the EBSWR design would be unable to provide adequate protection of the public from radiation. No change was made to the rule, the DCD, or the EA as a result of this comment.

Comment: The NTTF recommended that licensees reevaluate the seismic and flooding hazards at their sites and if necessary update the design-basis and SSCs important to safety to protect against the updated hazards. NTTF Report, page 30. The ESBWR environmental documents must be supplemented in light of this new and significant information. The NTTF's findings and recommendations are directly relevant to environmental concerns and have a bearing on the proposed action and its impacts. They demonstrate a need to reevaluate the seismic and flooding hazards on the ESBWR reactors, the environmental consequences such hazards could pose, and what, if any, design measures could be implemented (i.e., through NEPA's requisite “alternatives” analysis) to ensure that the public is adequately protected from these risks.

(P6-4)

NRC Response:

The NRC disagrees with the comment. Recommendation 2 of the NTTF, which is the subject of the comment, was focused on licensees of nuclear power reactors and was addressed through site-specific evaluations of the adequacy of the design of the reactors as applied to the site-specific seismic and flooding characteristics. By contrast, the ESBWR design certification—as any other design certification—is not approved for use on

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.

Nonetheless, the NRC disagrees with the comment. The NTTF suggestions that COL applicants or holders address Recommendations 4 and 7, rather than the design certification applicant during the certification process, would not necessitate those COLs to be considered “prototypes.” The Commission has stated that “the agency continues to evaluate the accident and its implications for U.S, facilities and the full picture of what happened at Fukushima is still far from clear. In short, we do not know today the full implications of the Japan event for U.S. facilities.” CLI-11-05, 74 NRC at 167. Should changes need to be made to the ESBWR design as a result of the evaluation of the Fukushima event, the Commission has stated that “we have the authority to ensure that certified designs and combined licenses include appropriate Commission-directed changes before operation.”

Id.

at 163. Further, it is not contrary to the certification process to require changes resulting from Fukushima lessons learned on COLs. The NRC may, under 10 CFR 52.97(c), place conditions upon the COL that the “Commission deems necessary and appropriate.” Further, the requirements under 10 CFR 52.63(a)(1) provide a mechanism for the NRC to modify certified designs. Such design changes would be applied to all COL holders referencing this design under 10 CFR 52.63(a)(3). As a result, all COL holders referencing the certified design would be required to make such changes. Moreover, in appropriate (but relatively limited) circumstances the NRC could also impose changes as an “administrative exemption” to the issue finality provisions of 10 CFR 52.63 and the ESBWR analogous to what the NRC did in the aircraft impact assessment (AIA) final rule, 10 CFR 50.150 (72 FR 56287; October 3, 2007).

No change was made to the rule, the DCD, or the EA as a result of this comment.

Emergency Petition

NRC Note: The Emergency Petition is comment submissions P1 and P2 in this ESBWR design certification rulemaking proceeding.

Comment: The emergency petition is out of process and should be dismissed on that basis alone. However, if this petition is not so dismissed, the NRC should treat this petition, for aspects related to the single issue specifically regarding the ESBWR design certification rulemaking, as a public comment on the proposed rule.

(P4-1)

NRC Response:

The NRC need not address, in this rulemaking, the comment's suggestion that the emergency petition is out of process because the Commission considered the merits of it and related filings in its

Memorandum and Order,

CLI-11-05, 74 NRC at 141 (2011) (ADAMS Accession No. ML112521106). The Commission determined that the Emergency Petition should be denied in the relevant adjudicatory proceedings and, on its own motion referred the emergency petition to the NRC staff for consideration as comments in the ESBWR rulemaking.

To the extent that it is relevant to the ESBWR design certification rulemaking, the NRC agrees that the Emergency Petition should be treated as a public comment on the proposed rule. Comments in the Emergency Petition are addressed in this comment response portion of this statement of considerations for the final ESBWR DCR.

No change was made to the rule, the DCD, or the EA as a result of this comment.

Comment: The responses, filed by various industry representatives and COL applicants in accordance with an April 19, 2011, Commission Order (ADAMS Accession No. ML111101277) and setting forth those representatives' and applicants' views on an “Emergency Petition” (ADAMS Accession No. ML111080855), were based on mischaracterizations of the Emergency Petition, incorrect representations regarding the NRC's response to the Three Mile Island accident, and incorrect interpretations of the law. Therefore, the responses should be rejected and the Emergency Petition should be granted.

(P5-1)

NRC Response:

On September 9, 2011, the Commission issued a Memorandum and Order on the Emergency Petition, CLI-11-05, 74 NRC 141 (ADAMS Accession No. ML112521106), which referred both the Emergency Petition and certain documents filed with the NRC to the NRC staff for “consideration as comments” in the applicable design certification rulemaking. CLI-11-05, 74 NRC at 176. Comment submission P5 was one of the documents referred by the Commission to the staff for consideration as comments. In accordance with the Commission's direction in CLI-11-05, comment submission P5 has been considered in the ESBWR rulemaking in a manner consistent with other comment submissions filed in the ESBWR rulemaking. Thus, the NRC reviewed the submission to determine the nature of the comments within this comment submission, if it is within the scope of the ESBWR rulemaking, and if so, what substantive response is appropriate. Based upon that review, the NRC determined that comment submission P5 is essentially a procedural reply to responses filed by other entities on the Emergency Petition. The NRC has determined that the reply does not contain any new substantive comments on the adequacy of the ESBWR design that were not already presented in the Emergency Petition and, therefore, has concluded that no further response is needed. No change was made to the rule, the DCD, or the EA as a result of this comment.

III. Regulatory and Policy Issues

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

Each of these issues identified above is discussed below.

1

1

Some of the regulatory and policy issues discussed below arose after the close of the public comment period on the March 24, 2011, proposed rule. The public was afforded an opportunity to comment on some of these issues in the May 16, 2014, supplemental proposed rule. Section V of the

SUPPLEMENTARY INFORMATION

section of this document describes the NRC's bases for not offering a comment opportunity for some of the regulatory and policy issues that arose after the close of the public comment period on the proposed rule.

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 Tōhoku 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

functions relied upon in the post-72-hour period are successful. Emergency procedures are to be developed by the COL applicant to support emergencies, which includes the period after 72 hours from the onset of the loss of all ac power. Further, the nonsafety-related equipment relied upon in the post-72-hour period has been designed in accordance with Commission policy (as described in Section 22.5.6.2 of the FSER) for use of augmented design standards for protection from external hazards and the NRC is engaging with COL applicants to ensure they have established appropriate availability controls for this equipment. Availability controls will be addressed in connection with a COL application referencing the ESBWR standard design.

The ESBWR design supports a COL applicant refilling the pools with offsite equipment, such as local fire pumpers. In the period beyond seven days from the onset of the event, the COL applicant will be responsible for describing how it will make available offsite sources, such as diesel fuel oil for the ancillary and standby diesel generators and water makeup to support long term cooling. The COL applicant must address the ability of offsite support to sustain these functions indefinitely, including procedures, guidance, training and acquisition, staging or installing needed equipment. Therefore, the NRC concludes that the ESBWR design, as described in the DCD, satisfies the underlying purpose of Order EA-12-049 insofar as it includes additional equipment to maintain or restore core and spent fuel pool cooling and containment function in the event of the loss of all ac power. While the ESBWR design includes all of the necessary design features in this respect, the COL applicant must address the programmatic aspects of Order EA-12-049. The NRC staff has already engaged with COL applicants on these arrangements. To the extent a COL applicant proposes to rely on additional equipment to perform required functions in the event of a loss of all ac power, that equipment is outside the scope of the standard ESBWR design and the NRC staff will evaluate it in connection with the COL application.

The NRC considered Recommendation 7.1, as modified by SRM-SECY-12-0025, using the requirements in Order EA-12-051, which describes the key parameters to be used to determine that a level instrument is considered reliable. JLD-ISG-2012-03, Revision 0, “Compliance with Order EA-12-051, Reliable Spent Fuel Pool Instrumentation,” (ADAMS Accession No. ML12221A339) endorses with exceptions and clarifications the methodologies described in the industry guidance document NEI 12-02, Revision 1, “Industry Guidance for Compliance with NRC Order EA-12-051, To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation,” (ADAMS Accession No. ML122400399) and provides an acceptable approach for satisfying the applicable requirements.

The NRC finds that the ESBWR design has design features that satisfy the underlying purpose of Order EA-12-051 for reliable SFP level instrumentation, except for two matters. The exceptions are whether the safety-related level instrumentation: (1) Are designed to allow the connection of an independent power source, and (2) will maintain its design accuracy following a power interruption or change in power source without recalibration. While the ESBWR design includes all of the necessary design features in this respect, the DCD did not include any information addressing these two matters. In addition, the NRC is currently developing a rulemaking which would address spent fuel pool instrumentation for beyond design basis events/accidents. This rulemaking may adopt different requirements than what is currently considered acceptable to meet the underlying purpose of order EA-12-051 and its related guidance. For these reasons, the NRC is excluding from issue finality and issue resolution these two aspects of the ESBWR spent fuel pool instrumentation design features. The exclusions have two consequences. First, any combined license applicant referencing the ESBWR design certification rule will have to provide information demonstrating that the NRC's requirements on these two matters are met. Second, the NRC need not address the factors of 10 CFR 52.63 either when it reviews the combined license application for adequacy with respect to these two matters, or in connection with any amendment of the ESBWR design certification rule imposing requirements to govern those matters.

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

In Section III, “Scope and Contents,” of the proposed ESBWR DCR (76 FR 16549; March 24, 2011), the only document for which the NRC proposed to obtain approval from the Office of the Federal Register (OFR) for incorporation by reference into the ESBWR design certification rule was the ESBWR DCD, Revision 9 (DCD Revision 9). Such approval would make DCD Revision 9 a legally-binding requirement on any referencing combined license applicant and holder by virtue of publication in the

Federal Register

as a final rule. This was based upon the assumption that the DCD specified all necessary requirements in Tier 1 and Tier 2 (with the exception of non-public documents containing proprietary information,

2

security-related information,

3

and SGI).

2

For purposes of this discussion, “proprietary information” constitutes trade secrets or commercial or financial information that are privileged or confidential, as those terms are used under the Freedom of Information Act and the NRC's implementing regulation at 10 CFR part 9.

3

For purposes of this discussion, “security-related information” means information subject to non-disclosure under 10 CFR 2.390(a)(7)(vi).

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

4

but for some of the documents which were not publicly available, GEH had not created a publicly-available version of that document to support the public comment process. The creation of publicly-available versions of non-public documents to support the public commenting process and transparency has been a long-standing practice for

both design certification rulemakings and licensing actions.

4

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

To address the NRC's concerns, for those non-public documents which include information intended to be treated as requirements and for which publicly-available versions were not previously created, GEH created publicly-available versions of those non-public documents. GEH also submitted Revision 10 to the DCD (DCD Revision 10), which included three tables in Section 1.6 that superseded Tables 1.6-1 and 1.6-2 in DCD Revision 9. These three tables—Tables 1.6-1, “GE/GEH Reports Incorporated by Reference,” 1.6-2, “Non-GE/GEH Reports Incorporated by Reference,” and 1.6-3, “Referenced Reports (not Incorporated by Reference,”—collectively clarify which documents are intended to be requirements and which documents are references only.

The supplemental proposed rule (79 FR 25715; May 6, 2014): (1) Announced the availability of DCD Revision 10; (2) described the distinction between those documents intended as requirements versus those which were for information only; (3) requested public comments on the NRC's intent to treat 50 non-public, referenced documents in DCD Revision 10 (listed in Table 2 of the supplemental proposed rule) as requirements and matters resolved in subsequent licensing and enforcement actions for plants referencing the ESBWR design certification; and (4) clarified, but did not request public comments on, the NRC's intent to obtain approval for incorporation by reference from the Director of the OFR for both DCD Revision 10 and the 20 publicly-available documents referenced in DCD Revision 10 (listed in Table 3 of the supplemental proposed rule), which are intended by the NRC to be requirements.

The 50 non-publicly available documents listed in Table 3 below are considered by the NRC to be requirements applicable to any combined license applicant or holder of a combined license referencing the ESBWR design certification rule, where the language of DCD Revision 10 makes clear that any one of those documents is intended to be a requirement. In addition, the 50 non-public documents are within the scope of issue resolution under Section VI of Appendix E, and are accorded issue finality protection under that Section VI and 10 CFR 52.63.

Table 3—50 Non-Public Documents Which the NRC Regards as Requirements, Are Matters Resolved Under Paragraph VI, ISSUE RESOLUTION, of the ESBWR Design Certification Rule, and Are Accorded Issue Finality Protection

Document No.

Document title

Publicly-

available

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

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 (Non-proprietary), September 2010

ML102770060

ML102770061

NEDC-33242P-A, NEDO-33242-A

Global Nuclear Fuel, “GE14 for ESBWR Fuel Rod Thermal-Mechanical Design Report,” NEDC-33242P-A, Revision 2, Class III (Proprietary), and NEDO-33242-A, Revision 2, Class I (Non-proprietary), September 2010

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 (Non-proprietary), 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

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

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 (Non-proprietary), 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 (Non-proprietary), 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 re-designation of Tier 2* information in paragraph VIII.B.6.c.(1) applies only to the ASME BPV Code, Section III, Subsections NE (Division 1) and CC (Division 2) for the design of ASME BPV Code Class MC (metal containment) and CC (concrete containment) pressure-retaining components (e.g., the containment vessel). This change does not apply to the design and construction of mechanical pressure-boundary components because they are required to meet the design and construction requirements in Section III for ASME BPV Code Class 1, 2, and 3 mechanical

pressure-boundary components, which are incorporated by reference into 10 CFR 50.55a. The regulations in 10 CFR 50.55a include provisions in paragraphs 50.55a(c)(3), (d)(2) and (e)(2) for reactor coolant pressure boundary, Quality Group B, and Quality Group C (i.e., ASME BPV Code Classes 1, 2, and 3 components, respectively. These paragraphs provide the necessary regulatory controls on the use of later edition and addenda to the ASME BPV Code, Section III through the conditions the NRC established on the use of paragraph NCA-1140 of the ASME BPV Code, Section III. As a result, these rule requirements adequately control the ability of a licensee to use later editions or addenda of the ASME BPV Code, Section III such that a Tier 2* designation is not necessary.

Second, paragraph VIII.B.6.c(3) is changed from “Motor-operated valves” to “Power-operated valves.” This change is necessary to correct an error in the proposed rule text. Consistent with Revisions 9 and 10 of the ESBWR DCD, which were the versions of the DCD available for public comment, the only valves that are described in Tier 2* information in an ESBWR nuclear power plant are air-operated rather than motor-operated.

Third, the NRC discussed in the supplemental proposed rule its proposal to designate the revised ESBWR steam dryer analysis methodology as Tier 2* information throughout the life of any license referencing the ESBWR DCR. This is a change from Revision 9 of the ESBWR DCD, which identified much of this information (in its earlier form before the revisions reflected in Revision 10) as Tier 2. Therefore, the ESBWR steam dryer analysis methodology was not identified as Tier 2* information in the proposed rule.

In the supplemental proposed rule, the NRC proposed to designate the revised ESBWR steam dryer pressure load analysis methodology as Tier 2* for two reasons. First, the NRC's experience with other applications using this methodology highlights the importance of the proper application of the steam dryer pressure load analysis methodology. Therefore, it is necessary for the NRC to review any changes a referencing applicant or licensee proposes to the methodology from that which the NRC previously reviewed and approved. Second, in Revision 10 to the ESBWR DCD, GEH revised the designation of this methodology to Tier 2* and, therefore, the rule's designation is consistent with GEH's designation in the DCD.

The supplemental proposed rule provided an opportunity for public comment on the proposed designation as Tier 2* of certain information related to the pressure load analysis methodology supporting the ESBWR steam dryer design. The NRC staff did not receive any public comments on the proposal to designate information related to the ESBWR steam dryer pressure load analysis methodology as Tier 2* information. Therefore, the final rule designates the revised ESBWR steam dryer pressure load analysis methodology as Tier 2* information throughout the life of any license referencing the ESBWR DCR.

D. Change Control for Severe Accident Design Features

The

SUPPLEMENTARY INFORMATION

section of the amendment to 10 CFR part 52 (72 FR 49392, at 49394; August 28, 2007), states that the Commission codified separate criteria in paragraph B.5.c of Section VIII of each DCR for determining if a departure from design information that resolves these severe accident issues would require a license amendment. Originally, the final rule was applied specifically to changes to ex-vessel severe accident design features. In the SRM to SECY-12-0081, “Risk-Informed Regulatory Framework for New Reactors,” dated October 22, 2012, the Commission directed the staff to make the change process in paragraph B.5.c of Section VIII applicable to severe accident design features, both ex-vessel and non-ex-vessel, that are described in the plant-specific DCD. This policy was changed after issuance of the proposed ESBWR rule. The policy was changed to ensure that, for changes to Tier 2 information, the effects on all severe accident design features—and not just ex-vessel severe accident design features—are considered.

However, the NRC has not changed the rule language in paragraph B.5.c of Section VIII for the ESBWR rulemaking because all of the relevant severe accident design features (i.e., those that are non-ex-vessel) are described in Tier 1 information. Tier 1 information, by definition, includes change controls in Section VIII of the rule text that meet the underlying purpose of the Commission's direction. Therefore, this change was not necessary for the ESBWR design certification.

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

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

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

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

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

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

The language of the ESBWR design certification rule differs from the rule language of other DCRs in two substantive areas. First, paragraph IX was reserved for future use because the substantive requirements in this paragraph (for other DCRs) has since been incorporated into 10 CFR part 52 in a 2007 rulemaking (72 FR 49352; August 28, 2007) and thus are no longer needed in the four existing DCR appendices. The NRC intends to remove these requirements from Section IX of the four existing DCR appendices in future amendment(s) separate from this rulemaking.

The second difference involves documents incorporated by reference into the ESBWR design certification rule. In the first four DCRs, the DCD is the only document identified in Section III of the rule language as being approved by the Office of the Federal Register for incorporation by reference. However, the ESBWR final rule identifies the ESBWR DCD and 20 publicly-available documents referenced in the DCD, Tier 2, Section 1.6 as approved for incorporation by reference. These 20 documents, which are intended by the NRC and GEH to be requirements, are listed in a table in Section III of the ESBWR final rule language. By being approved by the Office of the Federal Register for incorporation by reference, Revision 10 of the DCD and the 20 publicly-available documents are considered to be requirements as if they had been published in the

Federal Register

.

IV. Technical Issues

The NRC issued an FSER for the ESBWR design in March 2011, and subsequently published the FSER as NUREG-1966 in April 2014. The NRC issued an advanced supplemental SER in April 2014 (ADAMS Accession No. ML14043A134) and plans to publish Supplement No. 1 to NUREG-1966, as described in Section III of the

SUPPLEMENTARY INFORMATION

section of this document, before this final rule becomes effective. The FSER and its supplement provide the basis for issuance of a design certification under subpart B to 10 CFR part 52.

The significant technical issues that were resolved during the initial review of the ESBWR design (i.e., the NRC staff's review of Revision 9 of the ESBWR DCD and development of an FSER) are: (1) Regulatory treatment of nonsafety systems (RTNSS), (2) containment performance, (3) control room cooling, (4) feedwater temperature operating domain, (5) steam dryer analysis methodology, (6) aircraft impact assessment, (7) the use of ASME Code Case N-782, and (8) an exemption for the safety parameter display system. These issues were discussed in the March 2011 proposed rule. No public comments were received on these issues.

After publishing the proposed rule, the NRC addressed several issues that were changed in Revision 10 of the DCD or required a change to the FSER. The NRC staff reviewed these changes and developed an advanced supplemental SER as described above. The issues that were resolved in the advanced supplemental SER are: (1) Steam dryer analysis methodology, (2) loss of one or more phases of offsite power, (3) spent fuel assembly integrity in spent fuel racks, (4) Turbine Building Offgas System design requirements, (5) ASME Code statement in Chapter 1 of the ESBWR DCD, and (6) clarification of ASME component design ITAACs. The NRC also made changes to the advanced supplemental SER after the publication of the supplemental proposed rule.

After publication of the proposed rule, the NRC addressed two issues that were not addressed in Revision 10 of the DCD or in the advanced supplemental FSER. These issues are: (1) Hurricane-generated winds and missiles, and (2) changes to Tier 2* information.

Each of these issues identified above is discussed below. The public was afforded an opportunity to comment on some of these issues in the May 6, 2014 supplemental proposed rule. Section V of the

SUPPLEMENTARY INFORMATION

section of this document describes the NRC's bases for not offering a supplemental comment opportunity for any of the other technical issues that arose after the close of the public comment period on the proposed rule.

A. Regulatory Treatment of Nonsafety Systems (RTNSS)

The ESBWR safety analysis credits passive systems to perform safety functions for 72 hours following an initiating event. After 72 hours, nonsafety systems, either passive or active, replenish the passive systems in order to keep them operating or perform post-accident recovery functions directly. The ESBWR design also uses nonsafety-related active systems to provide defense-in-depth capabilities for key safety functions provided by passive systems. The challenge during the review was to identify the nonsafety SSCs that should receive enhanced regulatory treatment and to identify the appropriate regulatory treatment to be applied to these SSCs. Such SSCs are denoted as “RTNSS SSCs” in the context of the ESBWR design. As a result of the NRC's review, the applicant added Appendix 19A to the DCD to identify the nonsafety systems that perform these post-72 hour or defense-in-depth functions and the basis for their selection. The applicant's selection process was based on the guidance in SECY-94-084, “Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs.”

To provide reasonable assurance that RTNSS SSCs will be available if called upon to function, the applicant established availability controls in DCD Tier 2, Appendix 19ACM, and TS in DCD Tier 2, Chapter 16, when required by 10 CFR 50.36, “Technical specifications.” The applicant also included all RTNSS SSCs in the reliability assurance program described in Chapter 17 of DCD Tier 2 and applied augmented design standards as described in DCD Tier 2, Section 19A.8.3. For the reasons set forth in Section 22.5 of the FSER, the NRC finds the applicant's treatment of the RTNSS SSCs, as described in the DCD, acceptable.

B. Containment Performance

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

design basis are described in detail in Section 6.2.2 of the DCD Tier 2. The NRC identified a concern regarding the PCCS long-term cooling capability for the period from 72 hours to 30 days following a LOCA. To address this concern, the applicant proposed additional design features credited after 72 hours to reduce the long-term containment pressure. The features are the PCCS vent fans and passive autocatalytic hydrogen recombiners as described in DCD Tier 2, Section 6.2.1. These SSCs have been identified in DCD Appendix 19A as RTNSS SSCs.

The NRC staff's review of the PCCS design is documented in Section 6.2.2 of the FSER. The following is a summary of key points of that review. The applicant provided calculation results to demonstrate that the long-term containment pressure would be acceptable and that the design complies with GDC 38. The NRC's independent calculations confirmed the applicant's conclusion and the NRC accepts the proposed design and licensing basis. The NRC also raised a concern regarding the potential accumulation of high concentrations of hydrogen and oxygen in the PCCS and Isolation Condenser System, which could lead to combustion following a LOCA. The applicant modified the design of the PCCS and Isolation Condenser System heat exchangers to withstand potential hydrogen detonations. Accordingly, the NRC concludes that the design changes to the PCCS and Isolation Condenser System are acceptable and meet the applicable requirements.

C. Control Room Cooling

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

D. Feedwater Temperature Operating Domain

In operating BWRs, the recirculation pumps are used in combination with the control rods to control and maneuver reactor power level during normal power operation. The ESBWR design is unique in that the core is cooled by natural circulation during normal operation, and there are no recirculation pumps. In Chapter 15 of the DCD, GEH references licensing topical report (LTR) NEDO-33338, Revision 1, “ESBWR Feedwater Temperature Operating Domain Transient and Accident Analysis.” This LTR describes a broadening of the ESBWR operating domain, which allows for increased flexibility of operation by adjusting the feedwater temperature. This increased flexibility reduces the duty (mechanical stress) to the fuel and minimizes the probability of pellet-clad interactions and associated fuel failures.

By adjusting the feedwater temperature, the operator can control the reactor power level without control blade motion and with minimum impact on the fuel duty. Control blade maneuvering can also be performed at lower power levels.

To control the feedwater temperature, the ESBWR design includes a seventh feedwater heater with high-pressure steam. Feedwater temperature is controlled by either manipulating the main steam flow to the No. 7 feedwater heater to increase feedwater temperature above the temperature normally provided by the feedwater heaters with turbine extraction steam (normal feedwater temperature) or by directing a portion of the feedwater flow around the high-pressure feedwater heaters to decrease feedwater temperature below the normal feedwater temperature. An increase in feedwater temperature decreases reactor power, and a decrease in feedwater temperature increases reactor power. As described in Section 15.1.6 of the FSER, the applicant provided analyses that demonstrated ample margin to acceptance criteria. For the reasons set forth in Section 15.1.6 of the FSER, the NRC concludes that the applicant has adequately accounted for the effects of the proposed feedwater temperature operating domain extension on the nuclear design. Further, the applicant has demonstrated that the fuel design limits will not be exceeded during normal or anticipated operational transients and that the effects of postulated transients and accidents will not impair the capability to cool the core. Based on the evaluation documented in Section 15.1.6 of the FSER, the NRC concludes that the nuclear design of the fuel assemblies, control systems, and reactor core will continue to meet the applicable regulatory requirements.

E. Steam Dryer Analysis Methodology

As a result of RPV steam dryer issues at operating BWRs, the NRC issued revised guidance in Regulatory Guide (RG) 1.20, “Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Initial Startup Testing,” and SRP Sections 3.9.2, “Dynamic Testing and Analysis of Systems, Structures, and Components,” and 3.9.5, “Reactor Pressure Vessel Internals,” for the evaluation of the structural integrity of steam dryers in BWR nuclear power plants. The guidance requested that applicants for BWR nuclear power plant design certifications, licenses, or license amendments perform analyses to demonstrate that the steam dryer will maintain its structural integrity during plant operation when experiencing acoustic and hydrodynamic fluctuating pressure loads. This demonstration of RPV steam dryer structural integrity consists of three general steps:

(1) Predict the fluctuating pressure loads on the steam dryer,

(2) Use these fluctuating pressure loads in a structural analysis to demonstrate the adequacy of the steam dryer design, and

(3) Implement a steam dryer monitoring program for confirming the steam dryer design analysis results during the initial plant power ascension testing and periodic steam dryer inspections.

In its March 2011 FSER, the NRC staff described its review of the GEH methodology used to demonstrate the steam dryer structural integrity as described in Revision 9 of the ESBWR DCD and four referenced topical reports on which the NRC staff had issued separate SERs. The NRC staff concluded that the methodology was technically sound and provided a conservative analytical approach for definition of flow-induced acoustic pressure loading on the steam dryer, and that the design provided assurance of the structural integrity of the steam dryer and demonstrated conformance with GDCs 1, “Quality Standards and Records,” 2 “Design Bases for Protection Against Natural Phenomena,” and 4, “Environmental and Dynamic Effects Design Bases.” The NRC received no public comments on the proposed rule with respect to the steam dryer analysis methodology.

Following the publication of the proposed rule, the NRC staff identified safety issues applicable to the ESBWR steam dryer structural analysis based on information obtained during the NRC's review of a license amendment request for a power uprate at an operating BWR nuclear power plant. Consequently, the NRC staff communicated to GEH in a letter dated January 19, 2012 (ADAMS Accession No. ML120170304), that it was concerned that the bases for its FSER on the ESBWR DCD and its SERs on several applicable GEH topical reports were no longer valid. Specifically, errors were identified in the benchmarking GEH used as a basis for determining fluctuating pressure loading on the steam dryer and errors were identified in a number of GEH's modeling parameters. The NRC staff subsequently issued requests for additional information (RAIs) and held multiple public meetings and non-public meetings (in which the NRC staff and GEH discussed GEH proprietary information) to clarify and discuss the safety issues with the ESBWR steam dryer analysis methodology. The NRC staff also conducted an audit of the GEH steam dryer analysis methodology at the GEH facility in Wilmington, North Carolina, in March 2012, and a vendor inspection, at that facility, of the quality assurance program for GEH engineering methods in April 2012.

To document the resolution of those issues, GEH revised the ESBWR DCD by removing references to its LTRs that addressed the ESBWR steam dryer structural evaluation and to reference new engineering reports that describe the updated ESBWR steam dryer analysis methodology. The following four LTRs were removed by GEH (public and proprietary versions cited):

• NEDE-33313 and NEDE-33313P, “ESBWR Steam Dryer Structural Evaluation,” all revisions

• NEDE-33312 and NEDE-33312P, “ESBWR Steam Dryer Acoustic Load Definition,” all revisions

• NEDC-33408 and NEDC-33408P, “ESBWR Steam Dryer—Plant Based Load Evaluation Methodology,” all revisions

• NEDC-33408, Supplement 1, and NEDC-33408P, Supplement 1, “ESBWR Steam Dryer—Plant Based Load Evaluation Methodology Supplement 1,” all revisions

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

• NEDO-33312 and NEDE-33312P, Rev. 5, December 2013, “ESBWR Steam Dryer Acoustic Load Definition”

• NEDO-33408 and NEDE-33408P, Rev. 5, December 2013, “ESBWR Steam Dryer—Plant Based Load Evaluation Methodology—PBLE01 Model Description”

• NEDO-33313 and NEDE-33313P, Rev. 5, December 2013, “ESBWR Steam Dryer Structural Evaluation”

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

• Tier 1, Chapter 2, Section 2.1, “Nuclear Steam Supply”

• Tier 1, Chapter 2, Section 2.1.1, “Reactor Pressure Vessel and Internals”

• Tier 2, Chapter 1, Tables 1.6-1, 1.9-21, and 1D-1

• Tier 2, Chapter 3, Section 3.9.2, “Dynamic Testing and Analysis of Systems, Components and Equipment”

• Tier 2, Chapter 3, Section 3.9.5, “Reactor Pressure Vessel Internals”

• Tier 2, Chapter 3, Section 3.9.9, “COL Information”

• Tier 2, Chapter 3, Section 3.9.10, “References”

• Tier 2, Chapter 3, Appendix 3L, “Reactor Internals Flow Induced Vibration Program”

The revisions to these documents enhance the detailed design and evaluation process related to the structural integrity of the ESBWR steam dryer in several ways. For example, the source of data used to benchmark the analysis methodology was modified in Revision 10 to the ESBWR DCD to a different operating nuclear power plant for which the NRC recently authorized an extended power uprate. In addition, the details of the design methodology were made more restrictive in several respects, including limiting the analysis methods for fillet welds and using more conservative data and assumptions. The changes also designate additional information as Tier 2* and clarify regulatory process steps for completing the detailed design and startup testing of the ESBWR steam dryer, including COL information items to be satisfied by a COL applicant, ITAACs to be met by a COL licensee, and model license conditions that may be proposed by a COL applicant.

The NRC staff reviewed the revised ESBWR DCD sections, new GEH engineering reports, and RAI responses and prepared an advanced supplemental SER to replace Section 3.9.5, “Reactor Pressure Vessel Internals,” of the original FSER. To maintain the description of the regulatory evaluation of all ESBWR reactor vessel internals in the same location, the advanced supplemental SER replaced the entire Section 3.9.5 in the original FSER, although only the ESBWR steam dryer discussion has been modified in the advanced supplemental SER in any significant respect. The advanced supplemental SER documents the NRC staff conclusion that Revision 10 to the ESBWR DCD and the referenced engineering reports provide sufficient information to support the adequacy of the design basis for the ESBWR reactor vessel internals. The advanced supplemental SER also documents the NRC staff conclusion that the design process for the ESBWR reactor vessel internals is acceptable and meets the requirements of 10 CFR part 50, appendix A, GDC 1, 2, 4, and 10; 10 CFR 50.55a; and 10 CFR part 52. Finally, the advanced supplemental SER documents the NRC staff conclusion that the ESBWR design documentation for the reactor vessel internals in Revision 10 to the ESBWR DCD is acceptable and provides the bases for the NRC staff conclusion that GEH's application for the ESBWR design certification meets the requirements of 10 CFR part 52, subpart B, that are applicable and technically relevant to the ESBWR standard plant design. The NRC adopts the above conclusions and finds, based on the application materials discussed in the FSER as modified by the advanced supplemental SER, that the ESBWR steam dryer design meets all applicable NRC requirements and may be incorporated by reference in a COL application.

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).

G. ASME Code Case N-782

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

H. Exemption for the Safety Parameter Display System

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

I. Hurricane-Generated Winds and Missiles

Nuclear power plants must be designed to withstand the effects of natural phenomena, including those that could result in the most severe wind events (tornadoes and hurricanes). The design bases for plant structures, systems, and components must reflect consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin to account for the limited accuracy, quantity, and period of time in which the historical data have been accumulated. Initially, the U.S. Atomic Energy Commission, the predecessor to the NRC, considered tornadoes to be the bounding extreme wind events and issued RG 1.76, “Design-Basis Tornado for Nuclear Power Plants,” in April 1974, which reflected this technical position. RG 1.76 describes a design-basis tornado that a nuclear power plant should be designed to withstand without undue risk to the health and safety of the public. The design-basis tornado wind speeds were chosen so that the probability that a tornado exceeding the design-basis would occur was on the order of 10

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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

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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

ESBWR design is in conformance with GDCs 2 and 4 for hurricane wind and missile loads that are not bounded by the total tornado loads analyzed in the DCD. For these reasons, the NRC is only making a final safety determination on the acceptability of the ESBWR design with respect to loads on the applicable SSCs from hurricane winds and hurricane-generated missiles that are bounded by other loads analyzed in the DCD.

Accordingly, the NRC is excluding two issues from issue finality and issue resolution in the ESBWR DCD. First, with respect to the scope of the design in Section 3.3.2 of the ESBWR DCD, the NRC is excluding from finality the narrow issue of loads on applicable SSCs from hurricanes, but only to the extent that such loads are not bounded by other loads analyzed in the ESBWR DCD. Second, with respect to the scope of the design in Section 3.5.1.4 of the ESBWR DCD, the NRC is excluding from finality the narrow issue of loads on applicable SSCs from hurricane-generated missiles, but only to the extent that such loads are not bounded by other loads analyzed in the ESBWR DCD. This is accomplished in paragraph A.2.g of Section IV, “Additional Requirements and Restrictions,” and paragraph B.1 of Section VI, “Issue Resolution,” of the new appendix E to 10 CFR part 52, by excluding loads from hurricane winds and hurricane-generated missiles on the applicable SSCs from the finality accorded to the ESBWR design if they are not bounded as described. Under the exclusion, a COL applicant referencing the ESBWR DCR must demonstrate that loads from site-specific hurricane winds and hurricane-generated missiles are bounded by the total tornado load as analyzed in the ESBWR DCD. If the total tornado load analyses are not bounding, the COL applicant has several ways of addressing the exclusion, for example, demonstrating that the design can withstand the hurricane wind loads and hurricane-generated missile loads.

The NRC's narrow exclusion with respect to issue finality, as reflected in the ESBWR DCR language, does not require any change to the ESBWR design, the ESBWR DCD, or the NRC's EA supporting the ESBWR rulemaking. Nor are any changes required to the associated analyses for total tornado loads as described in the ESBWR DCD.

J. Loss of One or More Phases of Offsite Power

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

K. Spent Fuel Assembly Integrity in Spent Fuel Racks

Prior to publishing the proposed rule, the NRC performed its review of the integrity of spent fuel racks based on SRP Section 9.1.2, “New and Spent Fuel Storage.” This section states that “Designing the storage pool and fuel storage racks to meet seismic Category I requirements provides reasonable assurance that earthquakes will not cause a substantial coolant loss, a reduction in margin to criticality, or damage to the fuel assemblies.” This section supports the NRC's requirements in GDC 2, which requires that nuclear power plant SSCs important to safety be designed to withstand the effects of natural phenomena, such as an earthquake without loss of capability to perform their safety functions. The ESBWR FSER concluded that the design of the SFP, the buffer pool, and the fuel storage racks complied with the requirements of GDC 2 and met the guidance of SRP Section 9.1.2.

After publication of the proposed rule, the NRC recognized that Appendix D, “Guidance on Spent Fuel Racks,” to SRP Section 3.8.4, “Other Seismic Category I Structures,” states that, “It should be demonstrated that the consequent loads on the fuel assembly do not lead to damage of the fuel.” In other words, though the spent fuel rack may have remained intact during a seismic event, because there are gaps between the rack and the fuel assemblies, the applicant should demonstrate that the spent fuel assemblies in the rack have not sustained damage during that seismic event. During the NRC staff's review of the ESBWR design and prior to its publication of its FSER, the NRC staff did not specifically review the design of the spent fuel in the spent fuel racks against this guidance, but only against that of SRP Section 9.1.2 as described above.

To confirm the structural integrity of the fuel in the spent fuel racks, the NRC staff conducted an audit on August 5 and September 8, 2011. The audit summary is available under ADAMS Accession No. ML112860614. GEH subsequently submitted additional information (ADAMS Accession No. ML11269A093) to address whether the consequent loads on the fuel assembly that result from the design-basis seismic event would lead to fuel damage. For the reasons set forth in Section 3.8.4 of the supplemental FSER, the NRC finds that the fuel assemblies maintain structural integrity when subject to the design-basis seismic loads, the fuel assemblies in the fuel storage racks are structurally adequate to withstand the design-basis seismic loads, and the fuel assemblies are in compliance with GDC 2.

L. Turbine Building Offgas System Design Requirements

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

Because the NRC staff's evaluation of the turbine building structure came after completion of the FSER, issuance of the final SDA, and publication of the proposed rule, the NRC decided to

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

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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

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clarified that using alternatives to ASME Code requirements authorized in accordance with 10 CFR 50.55a is sufficient and does not require a COL holder to submit an exemption when changes involve a departure from only ASME Code requirements.

To clarify the use of alternatives when verifying compliance with ASME BPV Code ITAACs, GEH proposed to clarify in its Tier 1 definitions in Revision 10 to the ESBWR DCD, Section 1.1.1, “Definitions,” that “ASME Code” means ASME BPV Code requirements or any alternative authorized by the NRC pursuant to 10 CFR 50.55a(a)(3). This change does not affect previous NRC safety findings in the FSER or change the status of how the ESBWR standard design complies with ASME BPV Code requirements. For the reasons set forth in Section 14.3 of the supplemental FSER, the NRC finds that these changes to the definition of ASME Code are acceptable.

N. Clarification of ASME Component Design ITAACs

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

O. Corrections, Editorial, and Conforming Changes

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

V. Rulemaking Procedure

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

As described in Section III of the

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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 follo

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