ML18180A062

Issuance of Amendment Nos. 196 and 179 Regarding Application of the Seismic Probabilistic Risk Assessment Into the Previously Approved 10 CFR.50.69. Categorization Process

ML18180A062
Person / Time
Site:	Vogtle
Issue date:	08/10/2018
From:	Michael Orenak Plant Licensing Branch II
To:	Gayheart C Southern Nuclear Operating Co
Orenak, M D, NRR/DORL/LPLII-1, 415-3229
References
EPID L-2017-LLA-0248
Download: ML18180A062 (35)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 August 10, 2018 Ms. Cheryl A. Gayheart Regulatory Affairs Director Southern Nuclear Operating Company, Inc.

3535 Colonnade Parkway Birmingham, AL 35243

SUBJECT:

VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2- ISSUANCE OF AMENDMENTS REGARDING APPLICATION OF SEISMIC PROBABILISTIC RISK ASSESSMENT INTO THE PREVIOUSLY APPROVED 10 CFR 50.69 CATEGORIZATION PROCESS (EPID L-2017-LLA-0248)

Dear Ms. Gayheart:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 196 to Renewed Facility Operating License No. NPF-68 and Amendment No. 179 to Renewed Facility Operating License No. NPF-81 for the Vogtle Electric Generating Plant, Units 1 and 2 (Vogtle), respectively. The amendments consist of changes to the licenses in response to your application dated June 22, 2017, as supplemented by letters dated February 6, February 21, April 26, and August 6, 2018.

The amendments incorporate the use of the Vogtle seismic probabilistic risk assessment into the previously approved categorization process under Section 50.69, "Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors," of Title 10 of the Code of Federal Regulations.

A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/rhJ Gwiu Michael Orenak, Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-424 and 50-425

Enclosures:

1. Amendment No. 196 to NPF-68

2. Amendment No. 179 to NPF-81

3. Safety Evaluation cc: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 SOUTHERN NUCLEAR OPERATING COMPANY. INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON. GEORGIA DOCKET NO. 50-424 VOGTLE ELECTRIC GENERATING PLANT. UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 196 Renewed License No. NPF-68

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment to the Vogtle Electric Generating Plant, Unit 1 (the facility) Renewed Facility Operating License No. NPF-68 filed by the Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated June 22, 2017, as supplemented by letters dated February 6, February 21, April 26, and August 6, 2018, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes as indicated in the attachment to this license amendment, and paragraph 2.C.(11) of Renewed Facility Operating License No. NPF-68 is hereby amended to read as follows:

( 11 ) Additional Conditions The Additional Conditions contained in Appendix D, as revised through Amendment No. 196, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Additional Conditions.

In addition, the license is amended by changes as indicated in the attachment to this license amendment, and Appendix D, "Additional Conditions," to Renewed Facility Operating License No. NPF-68 is hereby amended to include a new license condition to read as follows:

Southern Nuclear Operating Company (SNC) is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the license amendments No. 173 (Unit 1) and No. 155 (Unit 2). SNC is approved to utilize the seismic probabilistic risk assessment (SPRA) model for use in the categorization process rather than the previously approved seismic margin approach.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.

3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~~~A~~

Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. NPF-68 and Appendix D Date of Issuance: August 1 o, 2018

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-425 VOGTLE ELECTRIC GENERATING PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 179 Renewed License No. NPF-81

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment to the Vogtle Electric Generating Plant, Unit 2 (the facility) Renewed Facility Operating License No. NPF-81 filed by the Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated June 22, 2017, as supplemented by letters dated February 6, February 21, April 26, and August 6, 2018, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 2

2. Accordingly, the license is amended by changes as indicated in the attachment to this license amendment, and paragraph 2.C.(5) of Renewed Facility Operating License No. NPF-81 is hereby amended to read as follows:

(5) Additional Conditions The Additional Conditions contained in Appendix D, as revised through Amendment No. 179, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Additional Conditions.

In addition, the license is amended by changes as indicated in the attachment to this license amendment, and Appendix D, "Additional Conditions," to Renewed Facility Operating License No. NPF-81 is hereby amended to include a new license condition to read as follows:

Southern Nuclear Operating Company (SNC) is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the license amendments No. 173 (Unit 1) and No. 155 (Unit 2). SNC is approved to utilize the seismic probabilistic risk assessment (SPRA) model for use in the categorization process rather than the previously approved seismic margin approach.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.

3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~~_) ~ A,~~-;;

Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. NPF-81 and Appendix D Date of Issuance: August 1 O, 2018

ATTACHMENT TO VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 LICENSE AMENDMENT NO. 196 RENEWED FACILITY OPERATING LICENSE NO. NPF-68 DOCKET NO. 50-424 AND LICENSE AMENDMENT NO. 179 RENEWED FACILITY OPERATING LICENSE NO. NPF-81 DOCKET NO. 50-425 Replace the following pages of the licenses and Appendix D, Additional Conditions, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Renewed Facility Operating License NPF-68 Remove Appendix D, Additional Conditions, to NPF-68 Remove D-3 D-3 Renewed Facility Operating License NPF-81 Remove Appendix D, Additional Conditions, to NPF-81 Remove D-3

7. Spent fuel pool mitigation measures (c) Actions to minimize release to include consideration of:

1. Water spray scrubbing

2. Dose to onsite responders (11) Additional Conditions The Additional Conditions contained in Appendix D, as revised through Amendment No. 196, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Additional Conditions.

D. The facility requires exemptions from certain requirements of 10 CFR Part 50 and 10 CFR Part 70. These include (a) an exemption from the requirements of 10 CFR 70.24 for two criticality monitors around the fuel storage area, and (b) an exemption from the requirements of Paragraph III.D.2(b )(ii) of Appendix J of 10 CFR 50, the testing of containment air locks at times when containment integrity is not required.

The special circumstances regarding exemption bare identified in Section 6.2.6 of SSER 5.

An exemption was previously granted pursuant to 10 CFR 70.24. The exemption was granted with NRC materials license No. SNM-1967, issued August 21, 1986, and relieved GPC from the requirement of having a criticality alarm system. GPC and Southern Nuclear are hereby exempted from the criticality alarm system provision of 10 CFR 70.24 so far as this section applies to the storage of fuel assemblies held under this license.

These exemptions are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. The exemptions in items b and c above are granted pursuant to 10 CFR 50.12. With these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.

E. Southern Nuclear shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

The plan, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Southern Nuclear Operating Company Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan," with revisions submitted through May 15, 2006.

Southern Nuclear shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Southern Nuclear CSP was approved by License Amendment No. 162, as supplemented by a change approved by License Amendment No. 175.

F. GPC shall comply with the antitrust conditions delineated in Appendix C to this license.

Renewed Operating License No. NPF-68 Amendment No. 196

Amendment Additional Condition Implementation Number Date 188 Southern Nuclear Operating Company (SNC) is approved to As stated in implement the Risk Informed Completion Time Program as the Additional specified in the license amendment request submittals dated Condition.

September 13, 2012, August 2, 2013, July 17, 2014, November 11, 2014, December 12, 2014, March 16, 2015, May 5, 2015, February 17, 2016, April 18, 2016, July 13, 2016, March 13, 2017, April 14, 2017, May 4, 2017, and June 2, 2017.

The licensee shall implement the items listed in Enclosure 1, Implementation items of SNC letter NL-15-0381 dated March 16, 2015 prior to the implementation of the Risk Informed Completion Time Program.

The risk assessment approach and methods, shall be acceptable to the NRC, be based on the as-built, as-operated, and maintained plant, and reflect the operating experience of the plant as specified in RG 1.200. Methods to assess the risk from extending the completion times must be PRA methods accepted as part of this license amendment, or other methods approved by the NRC for generic use. If the licensee wishes to change its methods, and the change is outside the bounds of this license condition, the licensee will seek prior NRC approval, via a license amendment.

196 Southern Nuclear Operating Company (SNC) is approved to Within 90 days of implement 10 CFR 50.69 using the processes for categorization of the issuance of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 the amendment.

structures, systems, and components (SSCs) specified in the license amendments No. 173 (Unit 1) and No. 155 (Unit 2). SNC is approved to utilize the seismic probabilistic risk assessment (SPRA) model for use in the categorization process rather than the previously approved seismic margin approach.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.

Vogtle Unit 1 D-3 Amendment No. 196

successfully demonstrated prior to the time and condition specified below for each:

a) DELETED b) DELETED c) SR 3.8.1.20 shall be successfully demonstrated at the first regularly scheduled performance after implementation of this license amendment.

(3) Southern Nuclear Operating Company shall be capable of establishing containment hydrogen monitoring within 90 minutes of initiating safety injection following a loss of coolant accident.

(4) Mitigation Strategy License Condition The licensee shall develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a) Fire fighting response strategy with the following elements:

1. Pre-defined coordinated fire response strategy and guidance

2. Assessment of mutual aid fire fighting assets

3. Designated staging areas for equipment and materials

4. Command and control

5. Training of response personnel (b) Operations to mitigate fuel damage considering the following:

1. Protection and use of personnel assets

2. Communications

3. Minimizing fire spread

4. Procedures for implementing integrated fire response strategy

5. Identification of readily-available pre-staged equipment

6. Training on integrated fire response strategy

7. Spent fuel pool mitigation measures (c) Actions to minimize release to include consideration of:

1. Water spray scrubbing

2. Dose to on site responders (5) Additional Conditions The Additional Conditions contained in Appendix D, as revised through Amendment No. 179, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Additional Conditions.

D. The facility requires exemptions from certain requirements of 10 CFR Part 50 and 10 CFR Part 70. These include (a) an exemption from the requirements of 10 CFR 70.24 for two criticality monitors around the fuel storage area, and (b) an exemption from the requirements of Paragraph III.D.2(b)(ii) of Appendix J of 10 CFR 50, the testing of containment air locks at times when containment integrity is not required. The special circumstances regarding exemption b are identified in Section 6.2.6 of SSER 8.

Renewed Operating License NPF-81 Amendment No. 179

Amendment Additional Condition Implementation Number Date 179 Southern Nuclear Operating Company (SNC} is approved to Within 90 days of implement 10 CFR 50.69 using the processes for categorization of the issuance of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 the amendment.

structures, systems, and components (SSCs} specified in the license amendments No. 173 (Unit 1} and No. 155 (Unit 2). SNC is approved to utilize the seismic probabilistic risk assessment (SPRA}

model for use in the categorization process rather than the previously approved seismic margin approach.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.

Vogtle Unit 2 D-3 Amendment No. 179

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 196 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-68 AND AMENDMENT NO. 179 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-81 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-424 AND 50-425

1.0 INTRODUCTION

By application dated June 22, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17173A875), as supplemented by letters dated February 6, February 21, April 26, and August 6, 2018 (ADAMS Accession Nos. ML180378121, ML180528342, ML18116A487, and ML18218A177, respectively), Southern Nuclear Operating Company, Inc. (SNC, the licensee), requested changes to the licenses for the Vogtle Electric Generating Plant, Units 1 and 2 (Vogtle). The supplements dated February 6, February 21, April 26, and August 6, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC, the Commission) staffs original proposed no significant hazards consideration determination as published the Federal Register on August 29, 2017 (82 FR 41072).

The proposed changes would incorporate the use of the Vogtle seismic probabilistic risk assessment (SPRA) into the Title 10 of the Code of Federal Regulations ( 10 CFR)

Section 50.69, "Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors," categorization process approved by the NRC in Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2 on December 17, 2014 (ADAMS Accession No. ML14237A034). The licensee's categorization process is based on the guidance in Nuclear Energy Institute (NEI) 00-04, Revision 0, "10 CFR 50.69 SSC Categorization Guideline," July 2005 (ADAMS Accession No. ML052910035), which was endorsed by the NRC in Regulatory Guide (RG) 1.201, Revision 1, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance" (ADAMS Accession No. ML061090627), which was issued in May 2006 for trial use.

Enclosure 3

By letters dated January 5, March 19, and March 28, 2018 (ADAMS Accession Nos. ML17354A782, ML18071A209, and ML18079A957, respectively), the NRC staff submitted requests for additional information (RAls) to the licensee. The licensee's RAI responses were provided in letters dated February 6, February 21, and April 26, 2018.

2.0 REGULATORY EVALUATION

2.1 Proposed Change In its submittal dated June 22, 2017, as supplemented by letter dated August 6, 2018, the licensee proposed the addition of the following condition to the Vogtle operating licenses to document the NRC's approval of the use of the seismic probabilistic risk assessment (SPRA) in the Vogtle 10 CFR 50.69 categorization process:

SNC is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the license amendments No. 173 (Unit 1) and No. 155 (Unit 2). SNC is approved to utilize the SPRA model for use in the categorization process rather than the previously approved seismic margin approach.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.

2.2 Regulations and Guidance On November 22, 2004 (69 FR 68008), the NRC amended 10 CFR Part 50 to add Section 50.69 that encompassed the risk-informed categorization and treatment of structures, systems, and components (SSCs) for nuclear power reactors. The regulations in 10 CFR 50.69 permit power reactor licensees and license applicants to implement an alternative regulatory framework with respect to "special treatment," which refers to those requirements that provide increased assurance beyond normal industry practices that SSCs will perform their design basis functions. Implementation of 10 CFR 50.69 requires that licensees first categorize safety-related and non-safety-related SSCs according to their safety significance. The SSCs are classified into high safety significant (HSS) and low safety significant (LSS) SSCs.

Alternative treatments per 10 CFR 50.69(b )( 1) and 10 CFR 50.69( d) can then be applied consistent with the categorization of the SSCs.

In May 2006, the NRC issued RG 1.201, Revision 1, for trial use. Revision 1 of RG 1.201 endorses, with conditions, NEI 00-04, Revision 0. Revision O of NEI 00-04 describes a process for determining the safety significance of SSCs and for categorizing them into the four risk-informed safety class (RISC) categories defined in 10 CFR 50.69. This categorization process uses an integrated decision-making process, incorporating both risk and traditional engineering insights. The NEI 00-04 guidance allows licensees to implement different approaches for categorization of SSCs, depending on the scope of their probabilistic risk assessment (PRA). The NEI 00-04 guidance allows the use of non-PRA type approaches for categorization when PRAs have not been performed by the licensee. These non-PRA type approaches include Fire-Induced Vulnerability Evaluation (FIVE), seismic margin analysis (SMA), and the use of the guidance in Nuclear Management and Resources Council (NUMARC) 91-06, "Guidelines for Industry Actions to Assess Shutdown Management,"

December 1991 (ADAMS Accession No. ML14365A203), to address shutdown operations.

Revision 1 of RG 1.201 states that the applicant is expected to document, at a minimum, the technical adequacy of the internal events PRA. Either PRAs or alternative approaches for hazards other than internal events may be used. One acceptable approach to determining the technical adequacy of a PRA is contained in RG 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," March 2009 (ADAMS Accession No. ML090410014). Revision 2 of RG 1.200 endorses, with clarifications, the American Society of Mechanical Engineers (ASME) and the American Nuclear Society (ANS) PRA standard ASME/ANS RA-Sa-2009 (henceforth referred to as the ASME/ANS 2009 Standard). The ASME/ANS 2009 Standard addresses PRAs for internal events and other hazards. Revision 2 of RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," May 2011 (ADAMS Accession No. ML100910006), provides guidance on the use of PRA findings and risk insights in support of licensee requests for changes to a plant's licensing basis. 1 Revision 2 of RG 1.174 provides risk acceptance guidelines for evaluating the results of such evaluations. The NRC staff notes that the consideration of RG 1.174, Revision 3, does not change the staff's conclusions documented in this safety evaluation (SE).

In Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2 issued on December 17, 2014, the NRC staff found that the licensee's process, as supplemented by the license conditions in the NRC staff's corresponding SE, was consistent with the NRG-endorsed NEI 00-04 guidance and was acceptable to satisfy the requirements of 10 CFR 50.69(c). In that categorization process, the licensee used the SMA, a non-PRA approach, for the assessment of the seismic risk. A license condition was added that stated, in part, that:

NRC prior approval, under 10 CFR 50.90, is required for a change to a categorization process that is outside the bounds specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

In the current submittal dated June 22, 2017, the licensee proposed the use of its SPRA in place of the SMA for categorization of SSCs under the licensee's previously approved 10 CFR 50.69 program. The applicable regulations for the review of this submittal are:

(1) Section 50.69(c), with a focus on Section 50.69(c)(1 )(i) that requires that the PRA must be of sufficient quality and level of detail to support the categorization process and must be subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC; (2) Section 50.69(b)(2)(iii) that requires the results of the PRA review process conducted to meet 10 CFR 50.69(c)(1 )(i) be submitted as part of the application; and (3) 10 CFR 50.69(e) that requires periodic updates to the licensee's PRA and SSC categorization.

3.0 TECHNICAL EVALUATION

In its submittal dated June 22, 2017, the licensee described the basis for the proposed change to the Vogtle operating licenses to incorporate the use of the Vogtle SPRA into the previously approved 10 CFR 50.69 categorization process. The NRC staff reviewed the submittal to determine if the proposed change satisfies the requirements of 10 CFR 50.69(c) by (1) verifying 1 RG 1.174, Revision 3 (ADAMS Accession No. ML17317A256), was issued in January 2018 during the NRC staffs review of this application. The NRC staff notes that the consideration of RG 1.174, Revision 3, does not change the staff's conclusions documented here.

conformance of the licensee's categorization process with the relevant NRC-endorsed guidance as it pertains to adoption of the SPRA, and (2) validating that the licensee's SPRA is acceptable for use in the categorization process for this specific application.

3.1 Overview of the Categorization Process As described in the August 31, 2012, application for Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2 (ADAMS Accession No. ML12248A035), the licensee's categorization process contains the following elements/steps:

• Qualitative assessment of system functions

• Component safety significance assessment

• Assessment of defense-in-depth (DID) and safety margin

• Risk sensitivity study

• Review by the Integrated Decision-Making Panel (IDP)

• Documentation

• Periodic reviews to ensure continued categorization validity and acceptable performance for those SSCs that have been categorized In Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2, the NRC approved the licensee's use of following approaches: (1) internal PRA to assess internal risk; (2) fire PRA to assess fire risk; (3) SMA to assess seismic risk; (4) Individual Plant Examination of External Events (IPEEE) screening to assess the risk from other external hazards (high winds, external floods); and (5) shutdown safety plan to assess shutdown risk. The additional license condition associated with Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2 states, in part, that a change to a categorization process must be submitted to the NRC under 10 CFR 50.90 for prior review and approval. The licensee stated in Section 3.1.3 of the current submittal that "the current version of the procedures describes both use of SPRA and SMA, and specifies use of the [Vogtle] SMA; this will be revised after approval is received to use the SPRA." The following sections summarize the NRC staff's review of each element in the licensee's proposed categorization process with regard to the proposed change to use SPRA.

3.2 Qualitative Assessment of System Functions The NRC staff's December 17, 2014, SE for Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2 describes the preliminary categorization of functions as "[o]nce a system is chosen for categorization, all of the system functions and all of the components belonging to that system are identified. The system functions are qualitatively categorized using a set of deterministic questions." The corresponding SE further states that "[t]he deterministic questions used by the licensee in the qualitative categorization of functions correspond to the seven questions provided in Section 9.2.2 of NEI 00-04." The preliminary categorization of functions was found to be acceptable by the NRC staff in Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2, and the proposed change does not alter any aspect of the preliminary categorization. Therefore, the NRC staff finds that the preliminary categorization of functions continues to be acceptable.

3.3 Component Safety Significance Assessment The submittal requests changing one of the approaches in the licensee's process to assess the safety significance of components. In the NEI 00-04 guidance, component risk significance is assessed separately for five hazard groups:

• Internal events

• Fire

• Seismic

• Other external hazards (tornadoes, external floods, etc.)

• Shutdown The regulation in 10 CFR 50.69(c)(1) requires the use of PRA to assess risk from internal events. For seismic hazards, 10 CFR 50.69(b )(2) allows the use of PRA or non-PRA approaches to assess risk.

The NRC approved the licensee's categorization process that uses PRA to assess risks from internal events (including internal flooding) and from fire in Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2. The current submittal requests to use a PRA for seismic risk in the categorization process instead of an SMA. The current submittal does not propose any changes to the licensee's process for using non-PRA approaches for the risk characterization of the following two risk hazard groups:

• IPEEE screening to assess the risk from other external hazards (high winds, external floods); and

• Shutdown safety plan to assess shutdown risk The use of SPRA by the licensee is consistent with the approaches in the NEI 00-04 guidance and is, therefore, acceptable. The capability and technical acceptability, as well as the application of the licensee's SPRA to the categorization of SSCs, is reviewed in Section 3.3.1 of this SE. Section 3.3.2 of this SE discusses the NRC staff's review of importance measures and sensitivity studies. Section 3.3.3 of this SE discusses the NRC staff's review of non-PRA approaches. Section 3.3.4 of this SE discusses the NRC staff's review of the safety significance assessment for passive components.

3.3.1 Acceptability of the SPRA to Support the Categorization Process In its submittal, the licensee proposes the addition of an SPRA to other parts of the licensee's PRA (internal events and fire PRA) that were reviewed and approved in Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2. The regulation in 10 CFR 50.69(c)(1 )(i) requires that the PRA must be of sufficient quality and level of detail to support the categorization process, and must be subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC. Furthermore, RG 1.201, Revision 1, states that if a licensee wishes to change its categorization approach, the NRC staff's review of the resulting submittal will focus on the acceptability of the methodology and analyses relied upon in the application.

The regulation in 10 CFR 50.69(b)(2)(iii) requires the licensee to submit the results of the PRA review process that was conducted to meet 10 CFR 50.69(c)(1 )(i) as part of the application.

The licensee submitted this information and, therefore, meets this requirement. The NRC staffs review of the submitted information is described throughout the rest of this section.

The regulation in 10 CFR 50.69(e) requires periodic updates to the licensee's PRA and SSC categorization. The licensee's periodic update program is discussed in Section 3.8 of this SE.

The NRC staff reviewed the periodic update program to determine whether the licensee's SPRA is acceptable to support the categorization of SSCs under the 10 CFR 50.69 program. Further, the NRC staff's review of the periodic update program also focused on ensuring that the categorization process and the results from the SPRA remain valid, as required by 10 CFR 50.69(e)(2) and (3).

Section 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance," June 2007 (ADAMS Accession No. ML071700658),

of NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition" (SRP), provides general guidance to the NRC staff for reviewing risk information used to support permanent plant-specific changes to the licensing basis. Section 19.2 of the SRP states, in part, that:

When licensees use RG 1.200 in support of an application, it should obviate the need for an in-depth review of the base PRA by NRC reviewers for those PRA aspects addressed, allowing the staff to focus their review on the application-specific impacts, key assumptions, and areas identified by peer reviewers and self-assessments as being of concern that are relevant to the application.

Revision 2 of RG 1.200 provides guidance for determining the acceptability of the base PRA used in risk-informed regulatory activities. Following the cited guidance documents, the NRC staff's review of the SPRA peer review process, key assumptions, and sources of uncertainty in the SPRA, and the results of the SPRA peer review are documented in the following subsections.

3.3.1.1 Evaluation of SPRA Peer Review Process The NRC staff reviewed the results of the peer review process for the SPRA presented in Section 3.2 and Attachment 2 to the submittal. The licensee's SPRA was subject to a self-assessment and a full-scope peer review conducted in November 2014. Section 3.2 of the submittal states that that the peer review was performed using the process defined in NEI 12-13, "External Hazards PRA Peer Review Process Guidelines," August 2012 (ADAMS Accession No. ML12240A027). The licensee further stated that no exceptions to the use of NEI 12-13 were noted in the peer review report. In RAI 1, the NRC staff requested that the licensee describe how the NRC comments on NEI 12-13 from a letter dated November 16, 2012 (ADAMS Accession No. ML12321A280), were addressed in the licensee's implementation of that guidance. 2 The NRC staff reviewed the licensee response to RAI 1 and finds it addresses the comments in the March 2018 acceptance letter of NEI 12-13. Therefore, the March 2018 letter does not change the NRC staff's conclusions in this SE.

2 The NRC staff issued a letter accepting the use of NEI 12-13, Revision 0, as modified by the NRC staff's comments, in March 2018 (ADAMS Accession No. ML18025C025). The acceptance letter states that the NRC staff's comments in the letter supersede those in the November 2012 letter.

In its February 6, 2018, response to RAI 1a, the licensee provided a description of the approach used by the licensee to ensure that the qualifications of the Vogtle SPRA peer review team complied with the corresponding requirements in the ASME/ANS 2009 Standard as endorsed in RG 1.200, Revision 2. The licensee stated that the qualifications of the Vogtle SPRA peer review team were reviewed against the requirements in the ASME/ANS 2009 Standard to establish that the peer review team met those requirements. The requirements address, among other things, the independence of the peer reviewers and specific training necessary for reviewing seismic fragilities. The licensee determined that the Vogtle SPRA peer review team had the individual and collective experience to meet the requirements in the ASME/ANS 2009 Standard through a combination of review of the team members' resumes, prior familiarity with team member experience, and consultation with the peer review team lead.

By letter dated March 27, 2017 (ADAMS Accession No. ML17088A130 ), the licensee responded to the March 12, 2012, 10 CFR 50.54(f) letter (ADAMS Accession No. ML12053A340) and included information regarding the qualifications of each Vogtle SPRA peer reviewer in Section A.3 of Appendix A of its response. Further, the licensee stated that the SPRA peer reviewers had no previous involvement in the Vogtle SPRA and cited the reviewers' signatures on the cover of the peer review report as certification of the reviewers' independence. Because the peer review team met the requirements in the ASME/ANS 2009 Standard, the NRC staff finds that the SPRA peer review team had the appropriate qualifications and independence to review the Vogtle SPRA used to support the current submittal.

Unreviewed Analysis Methods (UAMs) are a specific type of Facts and Observations (F&Os) assigned by peer reviewers and are defined in Section 3.2 of NEI 12-13. Further, the NRC staff's comments on NEI 12-13 in the November 16, 2012, letter stated that, "licensees that use UAMs for external hazards need to identify the UAMs in risk-informed applications to the NRC so that the NRC staff can evaluate the acceptability of these new methods in the context of their applications." In its February 6, 2018, response to RAI 1b, the licensee stated that the SPRA peer review team did not identify any UAMs in the licensee's SPRA. Therefore, further details regarding any UAM and a corresponding NRC staff review for this application are unnecessary.

The NRC staff finds that the licensee's response appropriately addresses the issue of UAMs in the Vogtle SPRA for this application.

In its February 6, 2018, response to RAI 1c, the licensee stated that the only application of expert judgment, as defined in Section 1-4.3 of the ASME/ANS 2009 Standard, is in the probabilistic seismic hazard analysis (PSHA). The licensee further stated that the Senior Seismic Hazard Analysis Committee (SSHAC) process was applied for determination of the seismic hazard for use in the SPRA. The SSHAC process, as described in NUREG/CR-6372, "Recommendations for Probabilistic Seismic Hazard Analysis: Guidance on Uncertainty and Use of Experts," April 1997 (Volumes 1 and 2 at ADAMS Accession Nos. ML080090003 and ML080090004, respectively), provides guidance for utilizing a structured expert elicitation process, including the technical and procedural aspects, for completing a PSHA. In Section A.4 of Appendix A of the licensee's March 27, 2017, response to the March 12, 2012, 10 CFR 50.54(f) letter, the licensee included excerpts from the peer review report. The excerpts from the review of the PSHA development for use in the SPRA state that "[t]he requirements of the SSHAC process satisfy the requirements of the standard." The licensee further stated that no other formal expert judgment was used in the SPRA fragility or logic model or was identified by the peer review. Because the peer reviewers did not identify any use of expert judgement in the licensee's SPRA and that the implementation of the SSHAC process for PSHA determination was peer reviewed, the NRC staff finds that the licensee satisfactorily addressed the use of expert judgement in its SPRA for this application.

In its February 6, 2018, response to RAI 1d, the licensee stated that the Vogtle SPRA was reviewed against Capability Category (CC) II of the ASME/ANS 2009 Standard for all applicable supporting requirements (SRs) and that any SRs that the reviewers found to meet only CC I had associated finding level F&Os. Because the licensee had a peer review for all SRs against CC II of the ASME/ANS 2009 Standard, the NRC staff finds that the licensee's SPRA was reviewed to the appropriate CC for this application.

The NRC staff's comments on NEI 12-13 in the November 16, 2012, letter included specific expectations related to an "in-process" peer review. In its February 6, 2018, response to RAI 1e, the licensee stated that an "in-process" peer review of the Vogtle SPRA was not performed. Instead, a final full-scope peer review was performed after the SPRA was fully developed to judge the technical adequacy of the SPRA model. Because an "in-process" review approach was not followed, the NRC staff does not need to review the details and process followed for the "in-process" reviews for the licensee's SPRA used to support this application.

Revision 2 of RG 1.200 endorses the ASME/ANS 2009 Standard Addendum A. The Vogtle SPRA peer review was performed using the SPRA requirements in Addendum B of the ASME/ANS 2009 Standard. Revision 2 of RG 1.200 does not endorse Addendum B, as noted in a July 6, 2011, letter to the ASME (ADAMS Accession No. ML111720067). Section 3.2 of the submittal provided discussions to address the NRC staff's comments in the July 6, 2011, letter in the context of establishing the technical capability of the Vogtle SPRA. In addition, the licensee's "basis for assessment" of the differences between each SR of Part 5 of Addendum B of the ASME/ANS 2009 Standard and to those in Addendum A in the context of the licensee's SPRA is provided in the Enclosure to the July 11, 2017, letter, "Response to Supplemental Information Needed for Acceptance of Systematic Risk-Informed Assessment of Debris Technical Report" (ADAMS Accession No. ML17192A245). The "basis of assessment" is applicable to the SPRA used for the current application because the same SPRA model and corresponding peer review were used to support both this application and the application regarding systematic risk-informed assessment of debris. The NRC staff reviewed the licensee's discussion in Section 3.2 of the current submittal and the information provided in conjunction with the July 11, 2017, letter. Based on the review of the "basis for assessment" for the difference between Addenda A and B, the NRC staff found that for SR SFR-C6, the licensee stated that its SPRA conformed to accepted current practices, but did not provide further details.

In RAI 10, the NRC staff requested that the licensee provide details of the approach followed to determine the median response and uncertainty for the soil structure interaction (SSI) analysis for the Vogtle SPRA and confirm the compatibility between the PSHA and SSI analysis. In its February 21, 2018, response to RAI 10, the licensee stated that the SSI input response spectra were from the site PSHA. The licensee further stated that site-specific dynamic soil profile properties that were developed were strain-compatible with the SSI input response spectra.

The licensee also provided additional details about the approach used for the SSI analysis, including the guidance followed and the justification for deviation from the guidance. The licensee stated that confirmatory analyses were performed to validate the accuracy of the SSI analysis of embedded structures. Based on its review of (1) the licensee's discussion in Section 3.2 of the submittal where the licensee addressed the NRC staff's comments on Addendum B in the context of the SPRA supporting this application; (2) the licensee's comparison of supporting requirement of Part 5 of Addendum B of the ASME/ANS 2009 Standard to those in Addendum A that was provided as part of a separate submittal, and (3) the supplemental information which provided details of the licensee's approach for performing the SSI analysis for use in its SPRA, the NRC staff finds that the licensee's use of Addendum B adequately addresses the technical elements for the

development of an SPRA. Therefore, the NRC staff concludes that the use of Addendum B is an acceptable alternative to the NRG-endorsed approach for the licensee's SPRA used to support this application.

Based on the findings that the licensee used the peer review guidance in NEI 12-13, appropriately addressed expert judgement and the NRC staff's comments on NEI 12-13, reviewed the SPRA to the appropriate CC, and that Addendum B is an acceptable alternative, the NRC staff concludes that the licensee appropriately implemented the peer review process in the context of the SPRA used to support this application.

3.3.1.2 Evaluation of Key Assumptions and Sources of Uncertainty Section 3.3 of RG 1.200, Revision 2, identifies two aspects necessary to demonstrate the technical acceptability of the PRA: ( 1) assurance that the pieces of the PRA used in the application have been performed in a technically correct manner, and (2) assurance that the assumptions and approximations used in developing the PRA are appropriate. Section 3.3.1 of RG 1.200, Revision 2, further discusses that various consensus PRA standards and industry PRA programs, as endorsed, may be interpreted to be adequate for the purpose of demonstrating that the first aspect (1, above) is met. Section 3.3.2 of RG 1.200, Revision 2, further discusses the second aspect (2, above) and clarifies that "[f]or each application that calls upon this regulatory guide, the applicant identifies the key assumptions and approximations relevant to that application. This will be used to identify sensitivity studies as input to the decision-making associated with the application." Revision 2 of RG 1.200 defines the terms "key assumption" and "key source of uncertainty" in Section 3.3.2, "Assessment of Assumptions and Approximations." Based on the above described guidance and the lack of identification of key assumptions and sources of uncertainty in the submittal, in RAI 5, the NRC staff requested the licensee to provide information on the approach for identifying key assumptions and sources of uncertainty in the SPRA, the results of the identification, and the disposition of the identified key assumptions and sources of uncertainty in the context of this application.

In its February 21, 2018, response to RAI 5a, the licensee described the approach followed to identify key assumptions and sources of uncertainty. The licensee stated that for the SPRA, two sets of assumptions, and the associated uncertainties, can impact the results. The first set of assumptions were those in the internal events PRA that were identified as potential sources of uncertainty in the licensee's August 31, 2012, submittal seeking approval of the 10 CFR 50.69 program. In the supplement dated February 21, 2018, the licensee provided a description of each of tho~e assumptions and the corresponding disposition or impact on the SPRA model in the context of this application. The second set of assumptions considered by the licensee were those made specifically for the SPRA. The licensee further stated that the definitions of "key assumption" and "key source of uncertainty" in RG 1.200, Revision 2, were used to guide and support the identification of such instances in the SPRA model. The approach used by the licensee identified three assumptions from the SPRA for more detailed review. For each assumption, the licensee performed a quantitative sensitivity study or a qualitative evaluation to demonstrate that the categorization in the 10 CFR 50.69 program would not be impacted by the corresponding assumption. The licensee's response included the description of each sensitivity study or qualitative evaluation and the corresponding disposition in the context of this application.

In its February 21, 2018, response to RAI 5b, the licensee listed the sensitivities performed in Section 5.7 of the licensee's March 27, 2017, response to the March 12, 2012, 10 CFR 50.54(f) letter. The licensee provided the rationale behind performing each specific sensitivity and

stated that none of those sensitivities addressed key assumptions and key sources of uncertainty, as defined in RG 1.200, Revision 2. The licensee referred to the response to RAI 7 in the February 21, 2018, supplement for the reasons for not considering the very small loss-of-coolant accident (VSLOCA) modeling in the Vogtle SPRA to be a key assumption or uncertainty. The NRC staff's review of the licensee's approach for VSLOCA modeling is provided in the discussion on F&O 16-18 in Section 3.3.1.3.2 of this SE.

In its February 21, 2018, response to RAI Sc, the licensee stated that when evaluated with respect to the definitions of key assumption and key uncertainty given in RG 1.200, Revision 2, none of the potential assumptions or uncertainties were determined to be key for this application and cited the sensitivity studies or qualitative evaluations discussed in the responses to RAI Sa and Sb to support this conclusion. Based on this conclusion, the licensee further stated that no additional sensitivity studies other than those in Table 5-4 of NEI 00-04, Revision 0, were identified in the licensee's characterization of PRA acceptability.

The NRC staff finds that the licensee searched for, identified, and evaluated sources of uncertainty in its SPRA consistent with the guidance in Section 3.3.2 of RG 1.200, Revision 2, to identify sensitivity studies as input to the decision-making associated with the application as well as the guidance in Table 5-4 of NEI 00-04 to identify additional "applicable sensitivity studies" for this application. The NRC staff concludes that the licensee provided sufficient information to appropriately disposition the identified assumptions and sources of uncertainty in the context of this application.

3.3.1.2.1 Westinghouse Generation Ill Shutdown Seals The licensee's internal events model was updated in 2015 and the major change during the update was the addition of the Westinghouse Owners Group (WOG) shutdown seal modeling.

The internal events model forms the basis for the SPRA model. Further, Section 3.1 of the Enclosure to the current submittal dated June 22, 2017, stated that the only additional sensitivity analysis required beyond those specified in NEI 00-04 was related to certain scenarios that affected the shutdown seal operation and its corresponding modeling. The NRC staff's SE for the topical report Pressurized Water Reactor Owners Group (PWROG)-14001-P, Revision 1, "PRA Model for the Generation Ill Westinghouse Shutdown Seal," dated August 23, 2017 (ADAMS Accession No. ML17200C876), imposed limitations and conditions on the use of the models and parameters presented in the report. One of the limitations and conditions regarding a shutdown seal temperature limit was directly related to the licensee's discussion in this application about the scenarios that affect the shutdown seal modeling. In RAI 3a, the NRC staff requested the licensee to clarify whether the shutdown seal model was peer reviewed as part of the licensee's internal events PRA or SPRA peer reviews and, if not, to justify why the addition of this shutdown seal model was not considered a PRA upgrade requiring a focused-scope peer review. In RAI 3b, the NRC staff requested the licensee to demonstrate how the limitations and conditions in the NRC staff's SE for PWROG-14001-P, Revision 1, are being met for the scenarios identified by the licensee in Section 3. 1 of the Enclosure to the current submittal for this application. The NRC staff reviewed the responses to RAI 3a and 3b, as well as follow-on RAI 3b-1, collectively.

In its February 6, 2018, response to RAI 3a, the licensee stated that the peer-reviewed Vogtle internal events PRA model, including internal flooding, that formed the basis for the SPRA did not include the Westinghouse Generation Ill low leakage (shutdown) seals. However, the peer-reviewed PRA model included the WOG 2000 reactor coolant pump (RCP) seal leakage model (ADAMS Accession No. ML031400376) to assess the plant's response to events that

result from a total loss of cooling to the RCP seals. The licensee cited the definition of a PRA upgrade in the ASME/ANS 2009 Standard (i.e., a new methodology, or a change in scope or change in capability that impacts the significant accident sequences or the significant accident progression sequences) to support its claim that the inclusion of the shutdown seals model did not constitute a PRA upgrade. The licensee's response, dated February 6, 2018, states, in part, that:

The change in the seal leakage model is not a new methodology since the new seal leakage model is simply an expansion of the current peer-reviewed model, with different failure probabilities and associated human action. [It is] not a change of scope of the model, i.e., the equipment, dependencies, and types of accident sequences remain the same. [It is] not a change in PRA model capability, i.e. the peer-reviewed PRA model can still evaluate the risk associated with station blackout and total loss of cooling events related to RCP seal failures.

The licensee further stated that although the lower seal failure rates, due to the inclusion of the shutdown seal model, will affect the ordering of the associated accident sequences, and reduce the core damage frequency (CDF) and large early release frequency (LERF) overall, the associated sequences are not significantly changed, and new sequences that had not already been modeled in the PRA and peer reviewed will not be generated.

In its February 6, 2018, response to RAI 3b, the licensee stated that the Vogtle SPRA model was revised after the June 22, 2017, submittal was transmitted to the NRC and the revised SPRA model incorporates the effects on RCP seal loss-of-coolant accidents (LOCAs) if the rated temperature of shutdown seal is exceeded in a timeframe insufficient to credit operator action following a seismic event. The licensee further stated that Vogtle SPRA model of record considered failure probabilities for each failure mode, common cause and associated human error probabilities (HEPs), as well as the seismic effects on HEPs. The Vogtle SPRA models specific actions required in response to the temperature limit being exceeded and that, "The model of record would be used to categorize SSC[s] per 10 CFR 50.69." The licensee provided information about how the Vogtle SPRA model of record addresses limitations and conditions 2, 4, and 5 in the SE for PWROG-14001-P, Revision 1 {limitations and conditions 1 and 3 are not relevant to the Vogtle SPRA model of record). The licensee stated that limitation and condition 2 is addressed probabilistically in the SPRA model of record, but did not provide further details. Therefore, in follow-on RAI 3b-1, the NRC staff requested the licensee to provide additional details on how the Vogtle SPRA met limitation and condition 2.

In its April 26, 2018, response to RAI 3b-1, the licensee stated that the event tree and fault tree modeling was used to incorporate the RCP shutdown seals into the SPRA and the potential for asymmetric cooling and high temperature failure of the seals was included in the logic model.

The licensee cited an analysis performed by the shutdown seal vendor for a case where feedwater flow is lost to a steam generator, termed asymmetric cooling, which concluded that the temperature in the cold leg of the effected loop would eventually exceed the shutdown seal temperature limitation in the SE for PWROG-14001-P, Revision 1. The licensee stated that the analysis further concluded that the temperature limit would not be reached if cooldown of the reactor coolant system via the secondary side was initiated before the affected steam generator dried out. The licensee further stated that the vendor's analysis assumed a dry-out time of 45 minutes and that the licensee assumed additional time for a "more realistic assumption" of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (i.e., an additional 15 minutes was assumed by the licensee) if the operator failed to initiate cooldown. The licensee explained the logic used for model asymmetric cooling in the SPRA by stating that following actuation of the shutdown seals, success of the seal operation

would occur if feedwater flow was available to all four steam generators or if there was auxiliary feedwater flow to two steam generators and the operator depressurized the steam generators and reactor coolant system within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

After reviewing the information regarding the modeling of the shutdown seals in the Vogtle SPRA, the NRC staff concludes that the licensee has demonstrated that the modeling of the shutdown seal model was performed in accordance with the SE for PWROG-14001-P, Revision 1, and that the applicable limitations and conditions therein were addressed. Further, the NRC staff finds the rationale for the additional 15 minutes used by the licensee for modeling shutdown seal failure under asymmetric cooling conditions is acceptable for this application.

The NRC staff does not, however, make any conclusions related to whether the inclusion of the shutdown seal model constitutes a PRA upgrade or the appropriateness of the value of 15 minutes used in the shutdown seal model implementation beyond this request to incorporate SPRA into the 10 CFR 50.69 categorization process.

3.3.1.2.3 SPRA Technical Acceptability-Based Sensitivity Studies Section 5.3 of NEI 00-04, Revision 0, recommends the completion of several sensitivity studies, including any applicable sensitivity studies identified in the characterization of PRA acceptability.

Table 5-4 of NEI 00-04 shows that one of the sensitivities to be performed for SPRA use in the categorization process is to use correlated fragilities for all SSCs in an area.

Section 4.4.2 of the licensee's March 27, 2017, response to the March 12, 2012, 10 CFR 50.54(f) letter indicates that equipment that is similar in design, with similar anchorage, and located in the same building on the same elevation is completely correlated in the SPRA model but also states that there were "few exceptions to this general correlation rule." In RAI 6a, the NRC staff requested the licensee to provide the bases for the "exceptions to the general correlation rule" and to clarify whether the sensitivity study for correlated fragilities will be performed for those exceptions. RAI 6b requested the licensee to clarify whether the sensitivity analyses identified in all other RAls (e.g., sensitivity analyses discussed in RAls 3, 5.c, 7.c.iv, 7.d, etc.) will be performed every time SSCs are categorized under 10 CFR 50.69.

RAI 6c requested the licensee to describe how the Vogtle 10 CFR 50.69 program continues to evaluate assumptions and sources of uncertainty when the Vogtle SPRA model is updated in the future to identify corresponding sensitivity analysis consistent with the guidance in NEI 00-04.

In its February 21, 2018, response to RAI 6a, the licensee stated that similar components at different locations, although at the same elevation, could have different demands, and thus different fragilities. The licensee further stated that detailed finite element models of the structures were developed as part of the Vogtle SPRA development and those models provided the seismic demands at different locations in a building. Seismic demands were used to develop the fragilities for components in different locations. The licensee stated that if there was a significant difference in fragilities, then the corresponding modeling resulted in the exception to the general correlation rule in that the lower capacity component could fail by itself, but was guaranteed to fail if the higher capacity component was failed. The licensee stated that the correlation sensitivity study in Table 5-4 of NEI 00-04 does not need to be performed because the correlation is for cases where full correlation is not used for all appropriate components and the Vogtle SPRA has full correlation implemented. The NRC staff finds that the licensee demonstrated that the exception to the general correlation rule implemented in the Vogtle SPRA represented a realistic correlation model in the context of this application. The NRC staff also finds that the Vogtle SPRA model includes full correlation for all relevant

components at the appropriate fragility value and, therefore, concludes that the sensitivity for correlated fragilities needs not be performed for this application as long as that modeling practice is implemented in the Vogtle SPRA.

In its February 21, 2018, response to RAI 6b, the licensee stated that no additional sensitivities were identified in the characterization of PRA adequacy that would be required to be performed for each categorization. Further, the licensee stated that none of the sensitivities discussed in the response to other RAls in the February 21, 2018, supplement were identified as a key source of uncertainty requiring specific sensitivities for categorization per Table 5-4 of NEI 00-04. The NRC staff's review of the licensee's responses to specific RAls, such as RAls 3, 5.c, 7.c.iv, and 7.d that discuss the need for additional sensitivity studies, is provided in the context of those particular RAls. In response to RAI 6c, the licensee stated that the evaluation of PRA model assumptions and sources of uncertainty is part of the licensee's PRA model maintenance process. The licensee further stated that if a change to the PRA model results in new assumptions or sources of uncertainty, those would be evaluated for impact on the 10 CFR 50.69 categorization process and consistent with NEI 00-04 and the Vogtle 10 CFR 50.69 program. The licensee explained that if a new assumption or source of uncertainty was identified that should be considered by the Integrated Decisionmaking Panel (IDP) in its consideration of categorization results, it would be added to the list of required sensitivities for the SPRA. Based on the NRC staff's review of specific RAls, such as RAls 3, 5.c, 7.c.iv, and 7.d identified above where the need for additional sensitivity studies is discussed, the NRC staff finds that additional sensitivities related to PRA adequacy were not identified and are, therefore, not required for this application. The NRC staff also concludes that the Vogtle PRA model maintenance process and the Vogtle 10 CFR 50.69 program addresses new assumptions or sources of uncertainty and corresponding sensitivity studies consistent with the NEI 00-04 guidance.

In summary, the NRC staff finds that the licensee has identified sources of uncertainty in its SPRA consistent with the guidance in RG 1.200, Revision 2. The NRC staff further finds that the licensee used the identified sources of uncertainly to determine sensitivity studies for input to the decision-making associated with the application, consistent with the guidance in Table 5-4 of NEI 00-04 to identify additional "applicable sensitivity studies" for this application. The NRC staff concludes that the licensee's process properly addresses new assumptions or sources of uncertainty and the corresponding sensitivity studies in the context of the use of its SPRA for categorization of SSCs.

3.3.1.3 Evaluation of Peer Review Findings 3.3.1.3.1 Internal Events Finding Level F&Os The licensee stated in its February 6, 2018, response to RAI 2 that there were 10 findings for the internal events PRA model, including internal flooding. The licensee further stated that

"[p]rior to conducting the [Vogtle] SPRA peer review, the finding-level F&Os from the [Vogtle]

internal events (including internal flooding) PRA peer review were dispositioned and incorporated into the PRA model as appropriate prior to use of the internal events PRA as the basis for the SPRA." The licensee also stated that the SPRA peer review team was provided with the internal events peer review report and the dispositions of the findings to facilitate their assessment of the acceptability of the internal events PRA model as the basis for the SPRA.

The licensee also provided a brief summary of the disposition of each of the 10 findings relative to the SPRA for this application. Further, Attachment 1 of Enclosure 1 of the licensee's supplemental response to Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on

Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors," dated September 13, 2004 (ADAMS Accession No. ML17116A096), includes a description of each of the 10 findings and details of the licensee's corresponding disposition. The NRC staff reviewed the licensee's response regarding the impact of the 10 findings from the internal events PRA model on the SPRA model in conjunction with the details of each of the findings and the corresponding resolution available in the above cited separate submittal. Because the licensee demonstrated that the internal events findings and their resolutions either do not impact the Vogtle SPRA or have been adequately considered in the Vogtle SPRA in the context of this application, the NRC staff finds that the licensee has established the technical acceptability of its internal events PRA model for use as the foundation for its SPRA.

3.3.1.3.2 SPRA Finding Level F&Os The NRC staff reviewed the licensee's resolution of all finding level F&Os from the Vogtle SPRA peer review provided in Attachment 2 to the current submittal and considered the potential impact of the findings on the acceptability of the application. The NRC staff requested additional information to clarify the licensee's disposition for some of the findings, as described in the following paragraphs. The licensee has stated that the resolution of each finding level F&O in Attachment 2 to the submittal is "for 50.69 and other applications," however, the NRC staff's review and conclusion on the resolution of the F&Os in Attachment 2 to the current submittal is restricted to this application only and does not extend beyond this application.

F&Os 12-23 and 12-24 (both related to SR SPR-E5), 12-27 (related to SR SPR-F2), and 12-29 (related to SPR-E2) discussed inconsistencies in the quantification of the Vogtle SPRA and the lack of information regarding the uncertainty analysis performed as part of the quantification.

The licensee stated that all the cited F&Os were related to parametric uncertainties in the Vogtle SPRA. The licensee provided additional details on F&Os 12-23, 12-24, and 12-29 and stated that the parametric uncertainty analysis has been re-performed for the SPRA submitted to support this application. The licensee further stated that the parametric uncertainty analysis included uncertainties in the seismic hazard, fragilities, non-seismic random unavailability, and HEPs and that the quantitative uncertainty results are now consistent with the point estimates.

With regard to F&O 12-27, the licensee stated that the F&O was related to the documentation of quantification of the plant response model including the uncertainty and importance analyses and that documentation has been updated to describe the quantification process in more detail.

The licensee further stated that the resolution of the F&Os 12-23, 12-24, 12-27, and 12-29 does not affect this application. Because the licensee has re-performed the parametric uncertainty analysis and the resolutions of the F&Os do not impact this application, the NRC staff finds that the licensee has resolved F&Os 12-13, 12-24, 12-27, and 12-29 for this application.

F&O 14-10, related to SR SFR-A2, assigns a CC I to that SR and states that significant conservatisms were noted is several sampled fragility calculations. The F&O cited specific components as examples of such instances and the resolution mentioned updates only to those components. Using conservative fragilities can lead to incorrect categorization and, therefore, incorrect treatment of SSCs based on the SPRA results. The licensee stated that the conservatisms identified in the cited F&O were related to the modeling of nozzle loads for mechanical equipment and the frequency range of interest used in the fragility evaluation. The licensee stated that a systematic approach was followed that involved initially applying realistic frequency range of interest for risk-significant components and subsequently applying realistic nozzle loads, where applicable, if the component remained risk significant. The licensee also provided details of the approach used to address the cited conservatism including the guidance that was followed. The NRC staff finds that the licensee has resolved F&O 14-10 for this

application because it has followed a systematic approach to address conservatisms in component fragility modeling beyond the examples identified in the F&O.

F&O 14-20, related to SRs SPR-B9 and SFR-E4, stated that the details of the walkdown procedure for seismic-fire interactions were missing, thereby implying that the peer reviewers did not have the opportunity to review the methodology employed during the walkdowns and the results therefrom. Section 4.2 of the licensee's March 27, 2017, response to the March 12, 2012, 10 CFR 50.54(f) letter provides a few examples of the type of seismic-fire interactions evaluated during the walkdown but does not provide details of the approach used during the walkdown. In RAI 7b, the NRC staff requested the licensee to describe the approach used during the walkdown for seismic-fire interactions. In its February 21 2018, response to RAI 7b, the licensee summarized the systematic process used for identifying the seismic-fire interaction sources. The licensee further stated that after the systematic screening of fire sources, there were no unique seismic induced fire scenarios in the Vogtle SPRA. The NRC staff finds that the licensee has resolved F&O 14-20 for this application because it used a systematic approach to identify seismic-fire interactions for inclusion in the SPRA model and the seismic-induced fire scenarios have been appropriately considered.

F&O 16-18, related to SR SPR-B8, noted the licensee's unique approach to screening out VSLOCAs based on walkdowns. The F&O stated that "little documentation exists of such walkdowns." The resolution treats the F&O as a documentation issue only. However, the F&O statements appeared to the NRC staff to indicate that the peer review team, due to the limitations cited in the F&O, did not review (or only partially reviewed) the associated documentation to determine the adequacy of the VSLOCA treatment. In RAI 7c, the NRC staff requested that the licensee justify the disposition of the F&O as a documentation issue and describe the approach followed for the systematic evaluation of the possible sources of VSLOCAs. The NRC staff also requested details of the sensitivity related to the VSLOCA modeling performed in the licensee's March 27, 2017, response to the March 12, 2012, 10 CFR 50.54(f) letter. In its February 21, 2018, response to RAI 7c, the licensee provided a description of the information that was available to the peer review team in the context of the statements in the cited F&O. The licensee stated that a dedicated discussion session on the treatment of VSLOCA was held with the peer reviewers and that the discussion included:

• A verbal description of the systematic walkdown process specific to the issue of VSLOCA,

• The amount and location of piping and tubing reviewed,

• The search for issues that would potentially cause piping/tubing failures,

• Photos of the in-vessel instrument lines and supports, and

• The engineered design for Vogtle piping and supports.

In addition, the licensee provided additional details of the approach taken to screen out VSLOCAs based on walkdowns. The licensee stated that the walkdowns sampled all areas of containment that were accessible with small-bore piping and tubing including all quadrants of containment, multiple elevations, as well as specific areas. The licensee stated that specific attention was paid to the design and supports for tubing to ensure adequate support, lack of seismic interactions issues that could fail the tubing, and sufficient flexibility to accommodate

differential movement between tubing anchor points. The licensee further stated that photos were taken of tubing in many areas of containment to document the walkdown process. The licensee stated that detailed fragility evaluations of LOCA-sensitive piping inside containment were also performed for several systems and that those fragility evaluations showed that the piping had very high seismic capacity. In regard to the sensitivity related to the VSLOCA modeling performed in the licensee's March 27, 2017, response to the March 12, 2012, 10 CFR 50.54(f) letter, the licensee stated that the sensitivity was performed using very conservative assumptions, including the use of small break LOCA logic, success criteria, and timing and that a more realistic model would have treated VSLOCA as a small leak with substantially relaxed success criteria. According to the licensee, the screening out of VSLOCA was a more realistic treatment rather than the generic assumption of a VSLOCA and, therefore, was not considered as a key uncertainty since it was well supported and demonstrated by the extensive walkdown, and the fragility evaluations of LOCA-sensitive piping. The NRC staff finds that the licensee has resolved F&O 16-18 for this application because the licensee provided sufficient information to support its approach of screening out VSLOCAs from the Vogtle SPRA based on walkdowns and the peer reviewers had sufficient information for the review of this topic.

F&O 16-5, related to SRs SPR-B1 and SPR-F1, cited concerns with the fragility selection for LOCA modeling in the Vogtle SPRA. Based on the discussion in the F&O and its resolution, it appeared to the NRC staff that the surrogate component(s) used to represent the fragility for LOCAs was changed subsequent to the peer review. Because using unrealistic fragilities can lead to incorrect categorization and, therefore, treatment of SSCs based on the SPRA results, the NRC staff requested the licensee to provide the technical justification for the selection of fragilities for modeling small break LOCAs and medium break LOCAs in the SPRA. In its February 21, 2018, response to RAI 7d, the licensee stated that the LOCA fragilities were based on plant-specific design basis calculations. The licensee further stated that the evaluation included multiple piping systems covering a range of piping sizes and systems, and that the RCP support was selected as the basis for the LOCA fragility because it was determined to have the lowest fragility. The licensee stated that since the failure of the RCP supports could cause small, medium, or large LOCAs, depending on the mode of failure, the same fragility was used for all three LOCA size ranges. The licensee cited a qualitative evaluation performed in response to RAI 5a to demonstrate that the uncertainty with the allocation of LOCA does not have a significant impact on this application. The NRC staff finds that the licensee has resolved F&O 16-5 for this application because it used plant-specific fragility values for modeling LOCAs in the Vogtle SPRA.

F&Os 16-4, 16-6, and 16-9, related to SRs SPR-B2, SPR-B1, and SPR-B4b, respectively, questioned the human reliability analysis (HRA) method employed in the SPRA model that was available for peer review. The resolutions of these F&Os stated that the Electric Power Research Institute (EPRI) guidance for HRA implementation for SPRAs was utilized and the "bins (breaking points) have been updated with additional breaking points ... to reflect seismic binning applicable" to the licensee. In RAI 7e, the NRC staff requested the licensee to justify the selection of the breaking points, including the "critical breaking point" (beyond which all HEPs are set to unity), with respect to the cited EPRI guidance and the fragilities of key SSCs in the SPRA. In its February 21, 2018, response to RAI 7e, the licensee stated that for the purposes of the HRA in the Vogtle SPRA, four damage-state bins were defined and provided a description of each bin along with the seismic acceleration range, termed breaking points. The licensee also provided the results of a sensitivity analysis to determine the impact of lowering the breaking points of each bin. Based on the results of the sensitivity, the licensee concluded that the breaking points used to define the HRA bins do not have a significant impact on the risk

metrics. The licensee also provided details on the calculation of the integrated performance shaping factors used in developing the HEPs for the Vogtle SPRA. The licensee stated that the only HRA sensitivity study that will be performed to support this application was the one recommended in NEI 00-04, Table 5-4. The licensee explained that the HRA sensitivity in NEI 00-04 revises the HEPs to the 95th and 5th percentiles which covered a large spectrum of potential HEPs and could provide adequate information to address the HRA related assumptions and uncertainties in the SPRA for this application. The NRC staff finds that the licensee resolved F&Os 16-4, 16-6, and 16-9 for this application because the licensee appropriately selected the breaking points for the HRA in the Vogtle SPRA and demonstrated, via a sensitivity, that the selected breaking points do not affect this application significantly.

Additionally, the NRC staff concludes that additional sensitivities related to the HRA in the Vogtle SPRA, beyond that in Table 5-4 of NEI 00-04, are not necessary for this application.

F&O 16-11, related to SR SPR-E2, stated that the review of the potential for additional HRA dependencies introduced by the SPRA model was missing. The resolution stated that the dependency analysis has been performed using the EPRI HRA Calculator which was implemented subsequent to the peer review. In RAI 7f, the NRC staff requested the licensee to clarify whether the use of the EPRI HRA Calculator in the SPRA was peer-reviewed and whether the use was expanded subsequent to the peer review. The NRC staff further requested the licensee to justify why the first time use and/or expansion of the use of the EPRI HRA Calculator in the SPRA should not be considered a PRA upgrade requiring a focused-scope peer review. In its February 21, 2018, response to RAI 7f, the licensee stated that the HEPs for the independent human failure events were estimated using the EPRI HRA calculator tool for the Vogtle SPRA and that those independent HEPs were peer reviewed. The licensee explained that the peer-reviewed SPRA model used a manual dependency analysis approach and the use of the EPRI HRA Calculator was expanded to include the HRA dependency analysis subsequent to the SPRA peer review. The licensee stated that the dependency tool in the EPRI HRA Calculator that was implemented subsequent to the SPRA peer review has a very similar set of rules when compared with the manual process. The licensee provided a description of the rules used for the manual dependency analysis as well as the dependency rules used in the EPRI HRA Calculator. The licensee further stated that the dependency rules and usage of the rules in the EPRI HRA Calculator was very similar to those used for manual analysis. The licensee also stated that the same dependency formulas were used for calculation in the manual and HRA Calculator based analysis. The licensee also provided a comparison of the value of the dependent HEPs using the manual analysis approach and the EPRI HRA Calculator for seven dependent failure event combinations that appeared in the detailed cutsets in the SPRA. The licensee described each dependency and the rationale for the differences in comparison values. The licensee also stated that the listed dependent HEP combinations were not significant to seismic CDF or LERF. The NRC staff finds that F&O 16-11 has been resolved for this application because the licensee demonstrated, through a comparison of the dependency rules and the values of the dependent HEPs, that the change from manual dependency analysis to the HRA Calculator dependency analysis does not adversely affect this application.

F&O 14-1, related to SR SFR-A2, stated that structural response factor used in all component fragilities reviewed by the peer review team is reported as 1.0. The F&O further stated that this factor would be greater than 1.0 because of the conservatism introduced in the demand through structural analysis, resulting in component and structural fragilities being biased low (i.e.,

conservative treatment). In RAI 7g, the NRC staff requested the licensee to provide the basis for the use of 1.0 as structural response factor as well as the corresponding uncertainty parameters. In its February 21, 2018, response to RAI 7g, the licensee stated that Vogtle is a

relatively soft soil site and the in-structure response spectra (ISRS) was driven by SSI effects.

The licensee further stated that for lower hazard levels (i.e., higher ground accelerations), the structural and soil damping increases, which counters the higher accelerations, thereby leading to similar ISRS amplitude as the ground motion response spectra level. The licensee explained that conservatisms in the response analysis identified by the peer review team were removed and a median response was computed by performing a realistic SSI analysis. The licensee also provided the basis for the randomness and uncertainty associated with each of the variables used in the fragility determination. The NRC staff finds that F&O 14-1 is resolved for this application because the licensee used realistic SSI analysis to develop the ISRS and provided the basis for the corresponding uncertainty parameters.

F&O 14-17, related to SR SFR-A2, stated that the reactor internal fragility evaluation in the licensee's SPRA determined the demand based on average spectra acceleration over the range of 2 to 3 Hertz, rather than using the peak acceleration in the range of the ISRS, and did not consider the contribution of higher modes. In RAI 7h, the NRC staff requested the licensee to justify that using the average spectral acceleration will not generate non-conservative fragilities and describe how the contribution of higher modes was considered. The NRC staff also requested the licensee to provide the basis for the fundamental frequency of 2-3 Hertz for the reactor internals. In its February 21, 2018, response to RAI 7h, the licensee stated that the fragility of the reactor internals was computed using the separation of variables approach and that median centered values were used for parameters that contributed to the demand. The licensee stated that a multiple-mode factor was applied to capture the contribution from higher modes per the guidance in EPRI TR-103959, "Methodology for Developing Seismic Fragilities."

The licensee indicated that the frequency range of interest was based on the seismic reliability proving tests for the reactors internals, control rod drive mechanism, rod cluster control, and fuel assemblies by Nuclear Power Engineering Test Center. The NRC staff finds that F&O 14-17 is resolved for this application because the licensee used a multiple-mode factor to address the higher modes and used tests to determine the frequency range of interest for the reactor internal components cited above.

F&O 14-7, related to SRs SFR-A2 and SFR-F4, stated that the fragility evaluation of the containment polar crane did not address the impact of variation in the fundamental frequency on the applicable seismic demand. The licensee's disposition of that F&O stated that the fragility evaluation was updated to address potential uncertainty in the fundamental frequency and contribution of higher modes. In RAI 7i, the NRC staff requested the licensee to describe how the fragility evaluation was updated to address the F&O. In its February 21, 2018, response to RAI 7i, the licensee stated that the polar crane was screened out of the SPRA model used to support this application because of the geometry of the crane and crane rails, and the tight fit within the containment. The licensee also stated that since the crane was not loaded during power operation, the stresses on the crane were low, and a collapse would be highly unlikely during an earthquake. The licensee further stated that the capacity of the crane due to vertical loading was updated to include a multi-mode factor consistent with the guidance in EPRI TR-103959 to capture the contribution of higher modes and the uncertainty associated with that factor. The NRC staff finds that the licensee's basis for screening-out the polar crane from the SPRA is acceptable, therefore, the NRC staff concludes that the resolution for F&O 14-7 does not impact this application.

F&O 14-8, related to SR SFA-F3, stated that the median capacity for two relays identified in the F&O was not realistic. The licensee's disposition of the F&O stated that the relay fragilities have been updated using the appropriate response and in-cabinet amplification factors. In RAI 7j, the NRC staff requested the licensee to describe how the relay evaluations were revised and cite

any applicable consensus approach used, including deviations from such approach. The NRC staff also requested the licensee to clarify whether a sensitivity analysis will be performed as part of the categorization process to address the uncertainty associated with this finding. In its February 21, 2018, response to RAI 7j, the licensee stated that the generic seismic ruggedness spectra (GERS) in EPRI NP-7147, "Seismic Ruggedness of Relays," contained seismic capacities of critical relays needed for safe shutdown of nuclear power plants. The licensee further explained that the electrical relay GERS were less generic and depended on details that varied with vintage and model number as compared with GERS for equipment classes. The licensee stated that the fragility analysis for relays followed the same methodology that is presented in EPRI TR-103959 for equipment evaluation based on testing. The licensee stated that the effective peak ISRS spectrum for relays was used and explained the approach for determining that parameter. The licensee stated that Cabinet Amplification Factors presented in EPRI TR-103959 express the most amplification generated at the worst location for a relay located in the specified cabinet types and that its approach is consistent with the methodology in EPRI TR-103959. The licensee also stated that additional sensitivity analysis will not be performed because the licensee's approach followed the standard industry approach without deviation. The NRC staff finds that the licensee has resolved F&O 14-8 for this application because the licensee used the Cabinet Amplification Factors approach to determine the relay fragilities.

F&O 14-9, related to SR SFA-A2, stated that the peer reviewers determined that certain valves, due to the corresponding valve operator heights and weights, would require further effort for resolution. The licensee's disposition stated that relevant valve operator heights and weights have been addressed through the corresponding fragility analysis. In RAI 7k, the NRC staff requested the licensee to describe how the relevant valve operator heights were taken into account in the fragility analysis and cite any applicable consensus approach used, including deviations from such approach and describe the impact of the deviations on this application. In its February 21, 2018, response to RAI 7k, the licensee provided details of the approach taken for the fragility analysis of the relevant valves, including any specific considerations and caveats. The licensee stated that a load based on Generic Implementation Procedure (GIP) for seismic verification of nuclear plant equipment was applied to the affected valve yoke's weakest direction. The licensee further stated that if the resulting yoke stresses are low and the relative deflections are small, then the caveat in GIP about the valve operator heights and weights was satisfied. The licensee also stated that if the GIP caveats were satisfied, the caveats in EPRI NP-6041, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," were satisfied as well because the EPRI NP-6041 caveats are based on the caveats in GIP. The NRC staff finds that F&O 14-9 is resolved for this application because the licensee used an acceptable approach for addressing the valves cited in the F&O, as described above.

In summary, the NRC staff concludes that the licensee resolved all of the finding level F&Os from the SPRA peer review for this application based on the potential impact of the findings on the categorization as well as the acceptability of the reported resolution for this application.

3.3.1.4 SPRA Acceptability Conclusion Pursuant to 10 CFR 50.69(c)(1 )(i), the categorization process must consider results and insights from a plant-specific PRA. The use of an SPRA to support categorization is endorsed by RG 1.201, Revision 1. The PRA must be acceptable to support the categorization process, and must be subjected to a peer review process assessed against a standard that is endorsed by the NRC. Revision 2 of RG 1.200 provides guidance for determining the acceptability of the PRA by comparing the PRA to the relevant parts of the ASME/ANS 2009 Standard using a peer

review process. The licensee has followed the guidance and submitted the results of the peer review.

The NRC staff reviewed the comparison between Addendum A and Addendum B of the ASME/ANS 2009 Standard, the peer review process and its results, and the licensee's resolution of the results and finds that the licensee's SPRA is acceptable to support the categorization of SSCs using the process endorsed by the NRC staff in RG 1.201.

3.3.2 Importance Measures, Integrated Importance Measures, and Sensitivity Studies Per Section 5.0 of NEI 00-04, Revision 0, the component safety significance assessment using PRA involves use of importance measures and sensitivity studies for both CDF and LERF.

First, the Fussell-Vesely (F-V) and Risk Achievement Worth (RAW) importance measures are obtained for each component and each hazard (i.e., separately for internal events, internal fire, and SPRAs) and compared to specified criteria. Then, sensitivity studies for each component and each hazard are performed. Last, integrated importance measures over all hazards are calculated per Section 5.6 of NEI 00-04.

3.3.2.1 Importance Measures and Integrated Importance Measures In the December 17, 2014, SE for Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2, the NRC staff found that the licensee had followed the guidance in NEI 00-04, Section 5.0. The current submittal does not request any changes to the licensee's component safety significance assessment. In RAI 9a, the NRC staff requested the licensee to describe how the SPRA importance measures are determined in the context of the "binning" approach employed in the licensee's SPRA for the seismic hazard and, therefore, the quantification of the SPRA. The NRC staff also requested the licensee to describe how the resulting importance measures are compared to the numerical criteria. In RAI 9b, the NRC staff requested information on the use of the SPRA importance measures in calculating the integrated importance measures. In its February 21, 2018, response to RAI 9a, the licensee stated that the F-V and RAW measures for a component for each seismic acceleration interval were calculated using a weighting approach and then the overall importance values (for F-V and RAW) for that component for the SPRA were determined by combining the importance values over all seismic acceleration intervals or "bins." The licensee stated that the F-V for a component from the SPRA was then combined with the F-V of the random failures for that component from the SPRA. In case of RAW, the licensee stated that "the maximum of the RAW for seismically induced failure and RAWs of random failures for that component is used to get a complete picture of the SPRA RAW importance measure." The licensee also stated that the resulting SPRA importance measures developed above would be compared to the F-V and RAW criteria in NEI 00-04 and if they were found to meet the criteria for classification as an HSS, then an integral assessment would be performed, as specified in NEI 00-04. The licensee further stated that if the integral assessment resulted in classification as an LSS, then that information, along with sensitivity information, would be given to the IDP for evaluation.

In its February 21, 2018, response to RAI 9b, the licensee stated that the formulae in NEI 00-04 for integrated F-V and RAW would be used to combine the seismic importance measures with the internal events and fire importance measures, and then compared to the NEI 00-04 criteria.

However, it was unclear from the licensee's response and the NEI 00-04 guidance how the integrated importance measures would be calculated for certain SPRA basic events that may not align with basic events in other PRA models. Examples of such SPRA basic events include SPRA basic events that are specific to the SPRA model or SPRA basic events that represent a

subcomponent modeled within the boundary of an internal events PRA component. Therefore, in RAI 9b-1, the NRC staff requested the licensee to describe and justify the calculation of integrated importance measures for SPRA basic events that may not align with basic events in other PRA models.

In its April 26, 2018, response to RAI 9b-1, the licensee stated that the importance evaluations performed in accordance with the process in NEI 00-04 were determined on a component basis.

The licensee provided details of integrated importance measure calculation approaches that would be employed for SPRA basic events that do not align with basic events in other PRA models. The licensee stated that subcomponents that were not directly modeled in other PRAs could be treated as another failure mode for the component to which it was associated. The licensee explained that the importance of such a subcomponent would be accounted for in the importance calculation for the corresponding component using the NEI 00-04 formulae for the integral assessment. The licensee further stated that the decision on the need to treat seismic basic events as representing subcomponents within the importance calculations for another modeled component would be made based on the modeling in each of the PRAs, as part of the PRA basic event-to-component mapping within the categorization process. For the case of SSCs that are unique to the SPRA and for which the seismic basic events were not explicitly modeled in the internal events or internal fire PRA, the licensee stated that if such SSCs are HSS based on the SPRA, then an integral assessment computation was not necessary and the safety significance would be presented to the IDP for their consideration in the decision-making process. The licensee provided examples to support the response. The NRC staff finds the licensee's approach for determining the SPRA-specific importance measures for basic events and calculating the corresponding integrated importance measures is acceptable for this application and follows the guidance of NEI 00-04.

As stated in the December 17, 2014, SE for Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2, in the internal events PRA, the component's importance measure includes its contribution to initiating events and its contribution to accident mitigation as described in NEI 00-04. In the SPRA, the fact that components cannot initiate seismic events is accounted for by only including the component's contribution to accident mitigation, which is consistent with the guidance in Section 5.3 of NEI 00-04 and, therefore, the NRC staff finds it to be acceptable.

In its February 6, 2018, response to RAI No. 12a, the licensee stated that a 10 percent margin to the importance measure thresholds in NEI 00-04 would be applied to the SPRA importance measures for categorization purposes. The NRC staff finds that the licensee's approach for the importance measures thresholds from the SPRA is consistent with the licensee's currently used process for categorization.

Section 5.1 of NEI 00-04 recommends that a truncation level of five orders of magnitude below the baseline CDF (or LERF) value should be used for calculating the F-V risk importance measures. The guidance also recommends that the truncation level used should be sufficient to identify all functions with RAW greater than 2. According to the information in Section 5.7.1 of the licensee's March 27, 2017, submittal in response to the March 12, 2012, 10 CFR 50.54(f) letter, the selected truncation limit for the "higher bins" of the Vogtle SPRA does not meet the guidance in NEI 00-04. In RAI 11 a, the NRC staff requested that the licensee demonstrate the impact of the selected truncation level for the "higher bins" in the SPRA on the importance measure criteria and the categorization. In its February 21 2018, response to RAI 11a, the licensee stated that only the highest three acceleration intervals in the SPRA did not meet the guidance on the truncation level in NEI 00-04. The licensee performed a sensitivity study by decreasing the truncation levels and recalculating the importance measures. The licensee stated that the truncation levels for the sensitivity study were made consistent with the

NEI 00-04 guidance except for the highest acceleration level for the seismic LERF calculation, which could not be reduced to the appropriate level due to computational limitations. Based on the results of the sensitivity study, the licensee stated that the change in the truncation did not make an appreciable difference for seismic CDF (SCDF) based importance measures, but did result in a change of the seismic LERF based RAW of several SSCs from below the corresponding threshold to above the threshold. The licensee stated that it plans to use five orders of magnitude lower truncation to the extent allowed by the software and hardware capabilities and continue to use the 10 percent margin for F-V and RAW thresholds when categorizing SSCs. The NRC staff finds that the licensee's use of the truncation levels is consistent with the NEI 00-04 guidance for this application because of the impact demonstrated by the sensitivity study and the use of the appropriate truncation level for the categorization process for this application.

According to the information in Section 4.4.1 of the licensee's March 27, 2017, submittal in response to the March 12, 2012, 10 CFR 50.54(f) letter, the value used for screening SSCs from the Vogtle SPRA "was adjusted until the maximum contribution was 2% of the final SCDF."

In RAI 11 b, the NRC staff requested that the licensee describe how the selected screening level in the SPRA maintains consistency with the importance measure criteria in NEI 00-04 and demonstrate the impact of the selected screening level in the SPRA on the importance measure criteria and the categorization. In its February 21, 2018, response to RAI 11.b, the licensee stated that the screening was conservative because no further refinement of the fragility was performed once the SSC fragility was evaluated to be above the screening value. The licensee further stated that many of the screened components would not lead directly to core damage or large early release and that the combination of failures needed to result in core damage or large early release would result in contribution lower than the screening threshold. The licensee also stated that SSCs that were judged to be important to CDF and LERF were included in the SPRA logic model even though their fragility was greater than the screening level. The licensee stated that a review of SSCs with fragilities greater than the screening level would be performed as part of the categorization to identify any SSC or correlated group of SSCs that could lead directly to core damage or large early release. The NRC staff finds the licensee's screening level in the SPRA to be acceptable for this application because it is not expected to adversely impact the categorization process.

Absolute importance measures are defined in Section 12 of NEI 00-04, endorsed by the NRC in RG 1.201, Revision 1, and are used when re-evaluating previously categorized SSCs with an updated PRA. In RAI 12b, the NRC staff requested that the licensee discuss any additional planned or anticipated changes to the licensee's categorization procedures related to use of absolute importance measures when re-evaluating previously categorized SSCs with an updated SPRA. In its February 6, 2018, response to RAI 12b, the licensee stated that, consistent with the licensee's current procedural guidance for internal events and fire PRAs, absolute importance will be used when previously categorized SSCs are re-evaluated with an updated SPRA. The NRC staff finds the licensee's approach for using absolute importance measures to be consistent with the licensee's currently used process for categorization that was approved in Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2.

The NRC staff finds the licensee's approach to determining the importance measures for the SPRA, and subsequently the integrated importance measures, to be consistent with the NRG-endorsed guidance in NEI 00-04 and the licensee's previously approved categorization program. Therefore, the NRC staff concludes that the calculation and use of importance measures from the SPRA by the licensee is acceptable for this application.

3.3.2.2 Sensitivity Studies Per the guidance in NEI 00-04, components not identified as HSS using the importance measure criteria above are further evaluated with sensitivity studies. The sensitivity studies are used to determine whether other conditions, such as the assumptions in the PRA, are masking the importance of an SSC. The importance measures obtained for each sensitivity study are compared to the screening criteria. The sensitivity studies determine the impact of the most uncertain parameters in the PRA and the impact of the different at-power operating configurations associated with online maintenance.

The methodology also requires that the evaluation of the acceptability of the licensee's PRA be evaluated to identify any other issues that should be addressed with a sensitivity study. The sensitivity studies required for SPRA in Table 5-4 of the NEI 00-04 guidance are as follows:

• Increase all human error basic events to their 95th percentile value

• Decrease all human error basic events to their 5th percentile value

• Increase all component common cause events to their 95th percentile value

• Decrease all component common cause events to their 5th percentile value

• Set all maintenance unavailability terms to 0.0

• Use correlated fragilities for all SSCs in an area

• Any applicable sensitivity studies identified in the characterization of PRA adequacy The licensee stated in Section 3.3.1.2 of the current submittal and clarified in response to RAI 6 that the sensitivity studies recommended in the NEI 00-04 guidance for the SPRA will be performed as part of the categorization process.

One of the sensitivity studies required for SPRA is to use correlated fragilities for all SSCs in an area. The NRC staff requested additional information on the licensee's treatment of correlated fragilities in RAI 6b. In RAI 6c, the NRC staff requested the licensee to clarify any applicable sensitivity studies that were identified in the characterization of PRA adequacy as part of the NRC staff's requests and the corresponding licensee responses. The NRC staff's review and conclusion on the licensee's responses to RAls 6b and 6c in the context of this application are provided in Section 3.3.1.2.3 of this SE.

In Section 3.1.3 of the current submittal, the licensee summarized its evaluation to determine if additional applicable sensitivity studies should be developed. Section 3.1.3 states that the SPRA assumptions and sources of uncertainty were reviewed to identify those that would be significant for the evaluation of this application. Section 3.1.3 further states that if the Vogtle SPRA model used a potentially non-conservative treatment or methods that are not commonly accepted, the underlying assumption or source of uncertainty was reviewed to determine its impact on this application. Finally, the licensee also stated in Section 3.1.3 that key Vogtle SPRA model specific assumptions and sources of uncertainty for this application have been identified and dispositioned within the Vogtle PRA documentation. The NRC staff requested additional information on the topic of SPRA assumptions and sources of uncertainty in RAls 3, 5, and 6. The NRC staff's review and conclusion on the licensee's responses to those RAls in the context of this application are provided in Section 3.3.1.2 of this SE.

3.3.3 Non-PRA Approaches In Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2, the NRC staff approved using the following non-PRA approaches in the licensee's categorization process:

• SMA to assess seismic risk;

• Screening during the IPEEE to assess risk from other external hazards (high winds, external floods);

• Shutdown Safety Plan to assess shutdown risk.

Seismic margin analysis is a screening method that does not quantify core damage frequency.

The current submittal incorporates the use of the Vogtle SPRA, which replaces SMA, into the licensee's 10 CFR 50.69 categorization process. Therefore, SMA will not be used in the licensee's 10 CFR 50.69 categorization process. Other previously approved non-PRA approaches are not changed by the submittal.

3.3.4 Component Safety Significance Assessment for Passive Components The submittal does not change any aspect of the passive categorization method approved in Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2. Therefore, the NRC staff did not review the licensee's safety significance assessment for passive components for this submittal.

3.3.5 Summary The NRC staff reviewed the licensee's SPRA-based approach proposed by the licensee for use in the licensee's previously approved 10 CFR 50.69 categorization process to assess the safety significance components and finds this method to be acceptable and consistent with RG 1.201 and the NRG-endorsed guidance in NEI 00-04. The NRC staff approves the use of the SPRA in the licensee's 10 CFR 50.69 categorization process. The use of other PRA and non-PRA approaches that were approved in Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2 for use in the 10 CFR 50.69 categorization process to assess the safety significance of active and passive components will remain unchanged.

The licensee's implementation of 10 CFR 50.69 is accomplished with the addition of a license condition. As stated in the license condition discussed in Section 2.1 of this SE, prior NRC approval is required for a change to the categorization process that is specified in the license condition.

3.4 Assessment of Defense-in-Depth and Safety Margin Assessment of defense-in-depth and safety margin is performed after the active and passive risk assessment and before performing the risk sensitivity studies. In the December 17, 2014, SE for Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2, the NRC staff found the licensee's assessment of defense-in-depth and safety margin to be consistent with the guidance provided in NEI 00-04, Section 6.0. The current submittal does not change the assessment of defense-in-depth and safety margin. Therefore, the NRC staff did not review the licensee's assessment of defense-in-depth and safety margin for this submittal.

3.5 Risk Sensitivity Study Consistent with Section 8.0 of NEI 00-04, an overall risk sensitivity study is performed for all the preliminary LSS components to confirm that the categorization process meets the acceptance guidelines of RG 1.174. This study is performed after the preliminary categorization is finalized and before review by the IDP. The NEI 00-04 guidance states that the unreliability of all preliminary LSS SSCs should be increased by a factor of 3 to 5. Separate sensitivity studies are to be performed for each system categorized, as well as a cumulative sensitivity study for all the SSCs categorized through the 10 CFR 50.69 process.

In RAI 8a, the NRC staff requested the licensee to describe and justify how the required risk sensitivity study outlined in Section 8.0 of NEI 00-04 will be performed for categorization using the SPRA. In its February 21, 2018, response to RAI 8a, the licensee stated that the implementation of the risk sensitivity study in the licensee's 10 CFR 50.69 program involves increasing the values for unreliability and unavailability for all SSCs that have been categorized as LSS by a factor of 3. The licensee stated that the factor used was not intended to be mechanistic, but instead was a test to determine that there continued to be adequate margin to the RG 1.174 acceptance guidelines, assuming that alternate treatments for LSS SSCs might result in a decrease in reliability. The licensee further stated that the sensitivity would be expanded to include the SPRA and that other changes in the risk sensitivity study were not required. The NRC staff found the licensee's proposed approach to performing the risk sensitivity study for categorization using the SPRA to be insufficient to ensure that an adequate margin to risk guidelines exists if seismic capacities are affected, specifically when components do not have associated random failures or seismic failures dominate the random failures.

Therefore, the NRC staff requested the licensee to describe how the risk sensitivity study approach addresses the potential impact on seismic capacities in RAI 8-1.

The licensee's April 26, 2018, response to RAI 8-1 expanded on the response to RAI 8b and stated that its programs and processes provided reasonable confidence that the seismic capacities of LSS components would not be impacted by alternative treatment. The licensee stated that its current design change process was being strengthened to formally require an SPRA impact assessment to determine if the change in risk is acceptable. The licensee further stated that for changes that could impact the seismic configuration, the SPRA impact assessment would require an analyst to determine the seismic capacity of the new SSC. The licensee explained that following the impact assessment, a recommendation would be made based on the results and overall impact to plant risk for the procurement to proceed or to be revised, or the SSC could be re-categorized for the SPRA. The licensee also stated that a program existed for periodic monitoring of degradation that could affect the seismic capacity of components. The licensee explained that if an identified degradation appeared to challenge an SPRA modeling aspect, an impact evaluation on the results of the SPRA would be performed to determine whether the original categorization remains valid. The licensee further stated that a rigorous configuration management program has been implemented to maintain the configuration of SSCs in the plant. The licensee explained that under the configuration management program, unless an item that was to be procured was equivalent to an existing item, an appropriate design change process was utilized to ensure that design requirements remain unchanged as required by the 10 CFR 50.69 rule. The NRC staff's review of the response to RAI 8-1 finds that the licensee's current programs and planned enhancements provide reasonable assurance that the seismic capacities of LSS components would not be significantly impacted. Because of the licensee's programs and processes, the exclusion of the seismic capacity of LSS components in the risk sensitivity study is acceptable for this application.

This sensitivity study, together with the periodic review process approved in Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2 and further reviewed in Section 3.8 of this SE in the context of this application, assure that the potential cumulative risk increase from the categorization is maintained acceptably low. The NRC staff concludes that the licensee performs the risk sensitivity study consistent with the guidance in Section 8.0 of NEI 00-04 and, therefore, that the proposed process is acceptable.

3.6 Review by the Integrated Decision-Making Panel In the December 17, 2014, SE for Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2, the NRC staff found the licensee's IDP process to be consistent with the guidance in NEI 00-04. The IDP systematically considers the quantitative and qualitative information available regarding the various modes of plant operation and initiating events, including PRA quantitative risk results and insights (e.g., CDF, LERF, and importance measures); deterministic engineering insights (e.g., defense-in-depth, safety margin, and containment integrity); and other pertinent information (e.g., generic and plant-specific operational and performance experience, feedback, and corrective actions program) in the categorization of SSCs.

The IDP members collectively have expertise in the areas of PRA, operations, safety analysis, design engineering, and system engineering. As stated in the SE for Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2, the licensee's procedures identify training requirements for members of the IDP, which include training on risk-informed defense-in-depth philosophy, PRA fundamentals, details on plant-specific PRA analyses (such as modeling scope and assumptions), interpretation of risk importance measures, role of sensitivity studies and changes in risk evaluations, and the categorization process. The training and qualification of each IDP member is documented and maintained as a quality assurance record for the life of the plant.

In RAI 12c, the NRC staff requested the licensee to discuss whether planned or anticipated changes will be made to the plant procedures that describe and document the process used by the IDP for the categorization of SSCs, including defining the duties and responsibilities of the IDP, its composition, and the qualifications and training for the IDP members. In its February 6, 2018, response to RAI 12c, the licensee stated that the existing qualifications and training process for the IDP members will continue to be used with the explanation that the Vogtle SPRA is being used for categorization. The licensee further stated that detailed training of IDP members in SPRA modeling is not planned, but the PRA IDP member will be familiar with the important SPRA approaches, results, modeling assumptions, and sources of uncertainty and sensitivity. The NRC staff finds that the licensee's approach to the SPRA by the IDP is acceptable because a PRA IDP member would be familiar with important aspects of the SPRA that contribute to the categorization.

As approved in Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2, components that are found to be HSS from one of the following types of evaluations cannot be categorized as LSS, although the IDP may request that further clarification and/or analysis be performed and brought back to the IDP:

• Internal Events PRA

• Non-PRA evaluations of seismic, other external events, or shutdown risk hazards

• Passive categorization

Components found to be HSS only from the seismic PRA perspective may be categorized as LSS if the integrated assessment of component risk importance over all PRA models shows that the integrated risk importance measures (discussed in Section 3.3.2 of this SE) meet the LSS criteria.

The NRG staff finds that the proposed approach does not change the licensee's approved IDP process and includes criteria allowing the IDP to change the categorization of components using SPRA insights, consistent with the NEI 00-04 guidance and is, therefore, acceptable.

3.7 Documentation In the December 17, 2014, SE for Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2, the NRG staff found the licensee's documentation to be in conformance with the requirements of 10 CFR 50.69(f) and is acceptable. The current submittal does not change the licensee's documentation of its categorization process. Therefore, the NRG staff did not review the licensee's documentation of the categorization process for this submittal.

3.8 Periodic Review As described in the December 14, 2017, SE for Amendment No. 173 for Unit 1 and Amendment No. 155 for Unit 2, the licensee has administrative controls in place to ensure that the PRA models used to support the categorization reflect the as-built, as-operated plant over time.

Section 3.3 of the current submittal states that the licensee would follow industry guidance and common practice in determining whether an update of the SPRA may be warranted due to availability of new consensus seismic hazard information. Section 3.1.2 of the current submittal describes the licensee's PRA maintenance and updates process and states that the process includes provisions for monitoring potential areas affecting the PRA models and for assessing the risk impact of unincorporated changes. Further, the licensee states that the assessment of the impact of the changes will be performed no longer than once every two refueling outages.

RAI 4 requested the licensee to justify the reliance on industry guidance and common practice, which will include non-site specific considerations and timeframes, instead of the licensee's periodic PRA maintenance process, which is site-specific, to identify and determine the incorporation of new information into the seismic hazard results for the Vogtle SPRA. In its February 21, 2018, response to RAI 4, the licensee stated that should site-specific seismic events occur or new site-specific information become available that would have a bearing on the licensee's SPRA, that information would be considered as part of the licensee's PRA maintenance process. The licensee's PRA maintenance process is designed to ensure that the impact of new site-specific hazard information, should it become available, would be addressed appropriately in the categorization of SSCs per 10 CFR 50.69. In the absence of new and significant information, the licensee would rely on updates to the seismic hazard catalog and ground motion models. The impact of the updates on the Vogtle SPRA will be evaluated within the PRA maintenance process. The NRG staff finds that the licensee adequately addressed the consideration of new site-specific seismic hazard information as it becomes available as part of its PRA maintenance process.

The regulation in 10 CFR 50.69(c)(1)(iv) requires that the categorization process include evaluations that provide reasonable confidence that for SSCs categorized as RISC-3, any potential increase in GDF and LERF resulting from changes in treatment are small. The regulations in 10 CFR 50.69(e)(2) and (3) require the licensee to monitor the performance of RISC-1 and RISC-2 SSCs and consider the data collected for RISC-3 SSCs and make

adjustments to the categorization or treatment processes so that the categorization process and results are maintained valid. In RAI 8b, the NRC staff requested that the licensee describe how it will be determined that the modeling inputs in the SPRA, as well as those used for the risk sensitivity study, and, therefore, the categorization process and results, continue to remain valid to ensure compliance with the requirements of 10 CFR 50.69(e). In its February 21, 2018, response to RAI 8b, the licensee stated that the validity of modeling inputs in the SPRA will be maintained by reviewing changes to the plant, operational experiences, applicable plant and industry operational experience, and periodic update of the seismic PRA model. The NRC staff requested supporting information in RAI 8-1. The NRC staff's review of the licensee's response to RAI 8-1 is discussed in Section 3.5 of this SE and the NRC staff's conclusions therein are also applicable in the context of the requirements of 10 CFR 50.69(e).

The NRC staff finds that the licensee's PRA update program and associated re-evaluation of component importance will appropriately consider the effects of changing failure probabilities and changing plant configuration on the component safety-significant categories. Because the update program and resulting revaluations provide assurance that feedback and process adjustment requirements in 10 CFR 50.69(e) will be met, the NRC staff concludes that the proposed process and program are acceptable for this application.

3.9 Conclusion The NRC staff reviewed the proposed change to incorporate the use of the Vogtle SPRA into the previously approved 10 CFR 50.69 categorization process. The use of SPRA in the licensee's process, in conjunction with the new license condition, is consistent with the NRC-endorsed NEI 00-04 guidance, and thus satisfies the requirements of 10 CFR 50.69(c).

The NRC staff finds that the licensee's change to the categorization process considers results and insights from a plant-specific SPRA that is of sufficient quality and level of detail to support the categorization process. The licensee's SPRA has been subjected to a peer review process assessed against a standard or set of acceptance criteria that is acceptable to the NRC staff.

Additionally, the NRC staff finds that the licensee's categorization process using SPRA continues to determine SSC functional importance using an integrated, systematic process that reasonably reflects the current as-built, as-operated plant configuration and operating practices, and applicable plant and industry operational experience. Furthermore, the NRC staff finds that the sensitivity studies to assess the seismic risk and the periodic reviews of the SSCs categorization are adequate for ensuring that the potential risk increase from implementing 10 CFR 50.69 is maintained acceptably small. Finally, the NRC staff finds that the licensee's current programs and enhancements to the design change process allow for the timely identification of changes to SSC failure characteristics and associated failure probabilities that may have an impact on the categorization using SPRA. Based on the above, the NRC staff concludes that the proposed change to the previously approved SSC categorization process is acceptable. Additionally, the NRC staff concludes that the proposed license condition implements 10 CFR 50.69 using approaches consistent with those in the applicable NRC-endorsed guidance and is, therefore, acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Georgia State official was notified of the proposed issuance of the amendments on July 2, 2018. The NRC staff confirmed that the State official had no comments on July 2, 2018.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on August 29, 2017 (82 FR 41072). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b ), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: S. Vasavada M. Reisi-Fard K. Hsu Date: August 10, 2018

ML18180A062 *via email OFFICE NRR/DORL/LPL2-1 /PM NRR/DORL/LPL2-1 /LA NRR/DRA/APLB/RILIT/TL NAME MOrenak KGoldstein MReisi-Fard DATE 7/9/18 7/6/18 & 7/24/18 5/29/18 OFFICE OGC (NLO)* NRR/DORL/LPL2-1 /BC NRR/DORL/LPL2-1 /PM NAME JWachutka MMarkley MOrenak DATE 7/24/18 8/10/18 8/10/18