ML24052A306

Issuance of Amendment Nos. 331 & 310, Regarding Adoption of Title 10 of Code of Federal Regulations Section 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Plants

ML24052A306
Person / Time
Site:	McGuire, Mcguire
Issue date:	04/08/2024
From:	John Klos Plant Licensing Branch II
To:	Pigott E Duke Energy Carolinas
Klos L
References
EPID L-2023-LLA-0022
Download: ML24052A306 (46)

Text

April 8, 2024

Edward Pigott Site Vice President Duke Energy Carolinas, LLC McGuire Nuclear Station 12700 Hagers Ferry Road Huntersville, NC 28078-8985

SUBJECT:

MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 331 AND 310, REGARDI NG ADOPTION OF TITLE 10 OF THE CODE OF FEDERAL REGULATIONS SECTION 50.69, RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS (EPID L-2023-LLA-0022)

Dear Edward Pigott:

The Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 331 to Renewed Facility Operating License NPF-9 and Amendment No. 310 to Renewed Facility Operating License NPF-17 for the McGuire Nuclear Station, Units 1 and 2, respectively. The amendments modify the licenses in response to your application dated February 17, 2023, as supplemented by letter dated November 2, 2023.

The amendments allow implementation of the provisions of Title 10 of the Code of Federal Regulations Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, and adds a license condition to Appendix B, Additional Conditions, of the Renewed Facility Operating Licenses for McGuire Nuclear Station, Units 1 and 2, respectively.

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

E. Pigott If you have any questions, please contact me at john.klos@nrc.gov or call me at 301-415-5136.

Sincerely,

/RA/

John Klos, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Docket Nos. 50-369 and 50-370

Enclosures:

1. Amendment No. 331 to NPF-9

2. Amendment No. 310 to NPF-17

3. Safety Evaluation

cc: Listserv

DUKE ENERGY CAROLINAS, LLC

DOCKET NO. 50-369

MCGUIRE NUCLEAR STATION, UNIT 1

AMENDMENT TO RENEWED FACI LITY OPERATING LICENSE

Amendment No. 331 Renewed License No. NPF-9

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

A. The application for amendment to the McGuire Nuclear Station, Unit 1 (the facility), Renewed Facility Operating License No. NPF-9, filed by the Duke Energy Carolinas, LLC (licensee), dated February 17, 2023, as supplemented by letter dated November 2, 2023, 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 Renewed Facility Operating License No. NPF-9 by adding the following text to Appendix B, Additional Conditions, page B-4 to read as follows:

(5) Additional Conditions

The Additional Conditions contained in Appendix B, as revised through Amendment 331, are hereby incorporated into this renewed operating license.

Duke Energy Carolinas, LLC 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 B, Additional Conditions, to Renewed Facility Operating License No. NPF-15 is hereby amended to include a new license condition to read as follows:

Duke Energy 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) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and high winds; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2, Class 3, and non-class SSCs and their associated supports; the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in the ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic, and the alternative seismic approach described in Duke Energys submittal letter RA-18-0090 dated February 17, 2023; as specified in License Amendment No. 331 dated April 8, 2024.

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

3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

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

Attachment:

Changes to License No. NPF-9 and the Technical Specifications

Date of Issuance: April 8, 2024 DUKE ENERGY CAROLINAS, LLC

DOCKET NO. 50-370

MCGUIRE NUCLEAR STATION, UNIT 2

AMENDMENT TO RENEWED FACI LITY OPERATING LICENSE

Amendment No. 310 Renewed License No. NPF-17

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

A. The application for amendment to the McGuire Nuclear Station, Unit 2 (the facility), Renewed Facility Operating License No. NPF-17, filed by the Duke Energy Carolinas, LLC (the licensee), dated February 17, 2023, as supplemented by letter dated November 2, 2023, 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 Renewed Facility Operating License No. NPF-17 by adding the following text to Appendix B, Additional Conditions, page B-4 to read as follows:

(6) Additional Conditions

The Additional Conditions contained in Appendix B, as revised through Amendment 310, are hereby incorporated into this renewed operating license.

Duke Energy Carolinas, LLC shall operate the facility in accordance with the Additional Conditions.

Duke Energy 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) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and high winds; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2, Class 3, and non-class SSCs and their associated supports; the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in the ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic, and the alternative seismic approach described in Duke Energys submittal letter RA-18-0090 dated February 17, 2023; as specified in License Amendment No. 310 dated April 8, 2024.

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

3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

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

Attachment:

Changes to License No. NPF-17 and the Technical Specifications

Date of Issuance: April 8, 2024 ATTACHMENT TO LICENSE AMENDMENT NO. 331

MCGUIRE NUCLEAR STATION, UNITS 1 AND 2

RENEWED FACILITY OPERATING LICENSE NO. NPF-9

DOCKET NO. 50-369

AND

LICENSE AMENDMENT NO. 310

RENEWED FACILITY OPERATING LICENSE NO. NPF-17

DOCKET NO. 50-370

Replace the following pages of the Renewed Facility Operating Licenses and the Appendix B, Additional Conditions, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

License Pages

Remove Insert

NPF-9, page 3 NPF-9, page 3 NPF-17, page 3 NPF-17, page 3

Additional Conditions

Remove Insert

NPF-9, page B-4 NPF-9, page B-4 NPF-17, page B-4 NPF-17, page B-4

(4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components;

(5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproducts and special nuclear materials as may be produced by the operation of McGuire Nuclear Station, Units 1 and 2, and;

(6) Pursuant to the Act and 10 CFR Parts 30 and 40, to receive, possess and process for release or transfer such byproduct material as may be produced by the Duke Training and Technology Center.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level

The licensee is authorized to operate the facility at a reactor core full steady state power level of 3469 megawatts thermal (100%).

(2) Technical Specifications

The Technical Specifications contained in Appendix A, as revised through Amendment No. 331, are hereby incorporated into this renewed operating license.

The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Updated Final Safety Analysis Report

The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation.

Duke shall complete these activities no later than June 12, 2021, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following issuance of this renewed operating license.

Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

Renewed License No. NPF-9 Amendment No. 331

(4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components;

(5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproducts and special nuclear materials as may be produced by the operation of McGuire Nuclear Station, Units 1 and 2, and;

(6) Pursuant to the Act and 10 CFR Parts 30 and 40, to receive, possess and process for release or transfer such byproduct material as may be produced by the Duke Training and Technology Center.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level

The licensee is authorized to operate the facility at a reactor core full steady state power level of 3469 megawatts thermal (100%).

(2) Technical Specifications

The Technical Specifications contained in Appendix A, as revised through Amendment No. 310, are hereby incorporated into this renewed operating license.

The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Updated Final Safety Analysis Report

The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation.

Duke shall complete these activities no later than March 3, 2023, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following issuance of this renewed operating license.

Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

Renewed License No. NPF-17 Amendment No. 310 APPENDIX B

ADDITIONAL CONDITIONS

FACILITY OPERATING LICENSE NO. NPF-9

Duke Energy Carolinas, LLC comply with the following condit ions on the schedules noted below:

Amendment Additional Conditions Implementation Number Date 331 Duke Energy is approved to implement 10 CFR 50.69 Upon using the processes for categorization of Risk-Informed implementation Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 of Amendment structures, systems, and components (SSCs) using: No. 331.

Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and high winds; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2, Class 3, and non-class SSCs and their associated supports; the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance p rocess identified in the ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic, and the alternative seismic approach described in Duke Energys submittal letter RA-18-0090 dated February 17, 2023; as specified in License Amendment No. 331 dated April 8, 2024.

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

330 For the Risk-Informed Completion Time (RICT) Upon calculations within the Risk-Informed Completion Time implementation Program, a singular approach for the high winds external of Amendment hazard will be specified and utilized for a given RICT. No. 330.

Either a high winds penalty or a high winds probabilistic risk assessment (PRA) will be utilized in the RICT Program calculations. A high winds PRA and high winds penalty shall not be used simultaneously to determine RICTs within the RICT Program.

-B4-

Renewed License No. NPF-9 Amendment No. 331 APPENDIX B

ADDITIONAL CONDITIONS

FACILITY OPERATING LICENSE NO. NPF-17

Duke Energy Carolinas, LLC comply with the follo wing conditions on the schedules noted below:

Amendment Additional Conditions Implementation Number Date 310 Duke Energy is approved to implement 10 CFR 50.69 Upon using the processes for categorization of Risk-Informed implementation Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 of Amendment structures, systems, and components (SSCs) using: No. 310.

Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and high winds; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2, Class 3, and non-class SSCs and their associated supports; the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance p rocess identified in the ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic, and the alternative seismic approach described in Duke Energys submittal letter RA-18-0090 dated February 17, 2023; as specified in License Amendment No. 310 dated April 8, 2024.

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

309 For the Risk-Informed Completion Time (RICT) Upon calculations within the Risk-Informed Completion Time implementation Program, a singular approach for the high winds external of Amendment hazard will be specified and utilized for a given RICT. No. 309.

Either a high winds penalty or a high winds probabilistic risk assessment (PRA) will be utilized in the RICT Program calculations. A high winds PRA and high winds penalty shall not be used simultaneously to determine RICTs within the RICT Program.

Renewed License No. NPF-17 Amendment No. 310

B-4 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

RELATED TO

AMENDMENT NO. 331 TO RENEWED FA CILITY OPERATING LICENSE NPF-9

AND

AMENDMENT NO. 310 TO RENEWED FACILITY OPERATING LICENSE NPF-17

DUKE ENERGY CAROLINAS, LLC

MCGUIRE NUCLEAR STATION, UNITS 1 AND 2

DOCKET NOS. 50-369 AND 50-370

1.0 INTRODUCTION

By letter dated February 17, 2023 (Reference 1), as supplemented by letter dated November 2, 2023 (Reference 2), Duke Energy Carolinas, LLC (the licensee, Duke Energy) submitted a license amendment request (LAR) for the McGuire Nuclear Station, Units 1 and 2 (McGuire) to the Nuclear Regulatory Commission (NRC, the Commission). Specifically, Duke Energy proposed to add a following license condition to Renewed Facility Operating License Nos.

NPF-9 and NPF-17, respectively, to allow the implementation of 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 reacto rs. Specifically, the license proposed to add a license condition to Appendix B, Additional Conditions, that would state:

Duke Energy 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) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire and high winds; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2, Class 3, and non-class SSCs and their associated supports; the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in the ASME/ANS PRA Standard RA -Sa-2009 for other external hazards except seismic, and the alternative seismic approach described in Duke Energys submittal letter RA 0090 dated February 17, 2023; as

Enclosure 3

specified in License Amendment Nos. 331 (Unit 1) and 310 (Unit 2) dated April 8, 2024.

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

The provisions of 10 CFR 50.69 allow adjustment of the scope of the SSCs subject to special treatment requirements (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation) based on an integrated and systematic risk-informed process that includes several approaches and methods for categorizing SSCs according to their safety significance.1

The licensee provided additional information in its supplement dated November 2, 2023 (Reference 2). The supplement provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on May 16, 2023 (88 FR 31285).

2.0 REGULATORY EVALUATION

2.1 Applicable Regulations

The provisions of 10 CFR 50.69 allow adjustment of the scope of SSCs subject to special treatment requirements. Special treatment refers to those requirements that provide increased assurance, beyond normal industry practices, that SSCs perform their design-basis functions.

For SSCs categorized to be of low safety significance (LSS), alternative treatment requirements may be implemented in accordance with the regulation. For SSCs determined to be of high safety significance (HSS), requirements may not be changed.

Section 50.69 of 10 CFR contains requirements regarding how a licensee categorizes SSCs using a risk-informed process; adjusts treatment requirements consistent with the relative significance of the SSC; and manages the process over the lifetime of the plant. A risk-informed categorization process is employed to determine the safety significance of SSCs and place the SSCs into one of four RISC categories.

SSC categorization does not allow for the elimination of functional requirements or allow equipment that is required by the deterministic design-basis to be removed from the facility.

Instead, 10 CFR 50.69 enables licensees to focus resources on SSCs that make a significant contribution to plant safety. For SSCs that are categorized as HSS, existing treatment requirements are maintained and may be enhanced. Conversely, for SSCs categorized as LSS that do not significantly contribute to plant safety on an individual basis, the regulation allows an alternative risk-informed approach to treatment that still provides a confidence that these SSCs will satisfy functional requirements. Implementation of 10 CFR 50.69 allows licensees to improve focus on HSS equipment.

1 Regulatory Guide (RG) 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, May 2006 (RG 1.201, Reference 3), describes the SSC categorization process in its entirety. It identifies a variety of acceptable approaches and methods, including PRA as well as some that do not use PRA.

2.2 Applicable Regulatory Guidance (RG)

The NRC staff considered the following regulatory guidance during its review of the LAR:

RG 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance (Reference 3)

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (RG 1.200, (Reference 4))

RG 1.174, Revision 2, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (RG 1.174, (Reference 5))

NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking (Reference 6)

NUREG-0800, Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition, Chapter 19, Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance (Reference 7)

2.3 Applicable NRC-Endorsed Guidance

The Nuclear Energy Institute (NEI) issued NEI 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline (Reference 8), as end orsed by the NRC in RG 1.201, Revision 1 for trial use, with clarifications. 2 NEI 00-04 describes a process that the NRC staff considers acceptable for complying with 10 CFR 50.69. This process determines the safety significance of SSCs and categorizes them into one of four RISC categories defined in 10 CFR 50.69.

Sections 2 through 10 of NEI 00- 04 describe the following steps/elements of the SSC categorization process for meeting the requirements of 10 CFR 50.69:

Sections 3.2 and 5.1 provide specific guidance corresponding to 10 CFR 50.69(c)(1)(i).

Sections 3, 4, 5, and 7 provide specific guidance corresponding to 10 CFR 50.69(c)(1)(ii).

Section 6 provides specific guidance corresponding to 10 CFR 50.69(c)(1)(iii).

Section 8 provides specific guidance corresponding to 10 CFR 50.69(c)(1)(iv).

Section 2 provides specific guidance corresponding to 10 CFR 50.69(c)(1)(v).

Sections 9 and 10 provide specific guidance corresponding to 10 CFR 50.69(c)(2).

2 All citations of NEI 00-04 refer to Revision 0. Unless otherwise noted, this means the industry guidance as clarified in RG 1.201, Revision 1, and endorsed by the NRC for trial use.

Section 11 of NEI 00-04 provides guidance on program documentation and change control related to the requirements of 10 CFR 50.69(f).

Section 12 of NEI 00-04 provides guidance on the periodic review related to the requirements in 10 CFR 50.69(e).

Maintaining change control and periodic review provides confidence that all aspects of the program reasonably reflect the current as-built, as-operated plant configuration and applicable plant and industry operational experience as required by 10 CFR 50.69(c)(1)(ii).

3.0 TECHNICAL EVALUATION

3.1 Method of NRC Staff Review

An acceptable approach for making risk-informed decisions about proposed Technical Specification (TS) changes, including both perm anent and temporary changes, is to show that the proposed licensing basis changes meet the five key principles stated in Section C of RG 1.174, Revision 2. These key principles are:

Principle 1: The proposed change meets the current regulations unless it is explicitly related to a requested exemption (i.e., a specific exemption under 10 CFR 50.12, Specific Exemptions).

Principle 2: The proposed change is consistent with the defense-in-depth philosophy.

Principle 3: The proposed change maintains sufficient safety margins.

Principle 4: When the proposed changes result in an increase in CDF [core damage frequency] or risk, the increase should be small and consistent with the intent of the Commissions policy statement on safety goals for the operations of nuclear power plants.

Principle 5: The impact of the proposed change should be monitored by using performance measures strategies.

3.2 Traditional Engineering Evaluation

The traditional engineering evaluation below addresses the first three key principles of RG 1.174, Revision 2 that are pertinent to: (1) meets the current regulations, (2) consistent with defense-in-depth, and (3) maintains safety margins.

3.2.1 Key Principle 1: Proposed Change Meets the Current Regulations

Paragraph 50.69(c) of 10 CFR requires licensees to use an integrated decision-making process to categorize safety-related and non-safety-related SSCs according to the safety significance of the functions they perform. All SSCs fall into one of the following four RISC categories, which are defined in 10 CFR 50.69(a):

RISC-1: Safety-related SSCs that perform safety significant functions 3

RISC-2: Non-safety-related SSCs that perform safety significant functions

RISC-3: Safety-related SSCs that perform low safety significant functions

RISC-4: Non-safety-related SSCs that perform low safety significant functions

The SSCs are classified as having either HSS functions (i.e., RISC -1 and RISC-2 categories) or LSS functions (i.e., RISC -3 and RISC-4 categories). For SSCs that are HSS, 10 CFR 50.69 maintains current regulatory requirements for special treatment (i.e., it does not remove any requirements from these SSCs). For SSCs that ar e LSS, licensees can implement alternative treatment requirements in accordance with 10 CFR 50.69(b)(1) and 10 CFR 50.69(d). For SSCs in RISC-3, licensees can replace special treatment with an alternative treatment. For SSCs in RISC-4, 10 CFR 50.69 does not impose new treatment requirements.

Paragraph 50.69(b)(3) of 10 CFR states that the Commission will approve a licensees implementation of this section by issuance of a license amendment if the Commission determines that the categorization process satisfies the requirements of 10 CFR 50.69(c).

As stated in 10 CFR 50.69(b), after the NRC approves an application for a license amendment, a licensee may voluntarily comply with 10 CFR 50.69, as an alternative to compliance with the following requirements for LSS SSCs:

(i) 10 CFR part 21 (ii) a portion of 10 CFR 50.46a(b)

(iii) 10 CFR 50.49 (iv) 10 CFR 50.55(e)

(v) specified requirements of 10 CFR 50.55a (vi) 10 CFR 50.65, except for paragraph (a)(4)

(vii) 10 CFR 50.72 (viii) 10 CFR 50.73 (ix) Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, to 10 CFR part 50 (x) specified requirements for containment leakage testing (xi) specified requirements of Appendix A, Seismic and Geologic Siting Criteria for Nuclear Power Plants, to 10 CFR part 100

The NRC staff reviewed the licensees SSC categor ization process against the categorization process described in NEI 00-04 as endors ed by the NRC in RG 1.201, Revision 1 (References 8 and 3, respectively), and the acceptability of the licensees PRA for use in the application of the 10 CFR 50.69 categorization process. The staffs review, as documented in this safety evaluation, used the framework pr ovided in RG 1.174, Revision 2, and NEI 00-04.

Section 2 of NEI 00-04, in part, states that the categorization process includes eight primary steps:

1. Assembly of Plant-Specific Inputs (Section 3 of NEI 00-04)

3 NEI 00-04 uses the term high-safety-significant to refer to SSCs that perform safety-significant functions. The NRC understands HSS to have the same meaning as safety-significant (i.e., SSCs that are categorized as RISC-1 or RISC-2), as defined in 10 CFR 50.69.

2. System Engineering Assessment (Section 4 of NEI 00-04)

3. Component Safety Significance Assessment (Section 5 of NEI 00-04)

4. Defense-In-Depth Assessment (Section 6 of NEI 00-04)

5. Preliminary Engineering Categorization of Functions (Section 7 of NEI 00-04)

6. Risk Sensitivity Study (Section 8 of NEI 00-04)

7. Integrated Decision-Making Panel Review and Approval (Section 9 of NEI 00-04)

8. SSC Categorization (Section 10 of NEI 00-04)

In Section 3.1.1, Overall Categorization Process, of the enclosure to the LAR, the licensee stated that it will implement the risk-informed categorization process in accordance with NEI 00-04, as endorsed by the NRC in RG 1.201, Revision 1. In Section 3.2.3, Seismic Hazards, of the enclosure to the LAR, the licensee proposed the use of an approach developed by the Electric Power Research Institute (EPRI) 3002017583 for sites with moderate seismic hazard or margin as an alternative method to assess the applicable hazard contribution(s). The NRC staff notes that use of these alternative methods is a deviation from the NEI 00-04 guidance as endorsed.

The regulatory requirements in 10 CFR 50.69 and 10 CFR Part 50, Appendix B, and the monitoring outlined in NEI 00-04, with clarifications in RG 1.201, Revision 1, ensure that the SSC categorization process is sufficient to ensure that the SSC functions continue to be met and that any performance deficiencies will be identified, and appropriate corrective actions taken. The licensees SSC categorization program includes the appropriate steps and elements prescribed in NEI 00-04 to ensure that SSCs specified are appropriately categorized consistent with 10 CFR 50.69. The NRC staff performed a more detailed review of specific steps and elements of the licensees SSC categorization proc ess where necessary to confirm that they are consistent with NEI 00-04 guidance. Based on the above, the NRC staff concludes that the proposed 10 CFR 50.69 program meets the first key principle for risk-informed decision-making prescribed in RG 1.174, Revision 2.

3.2.2 Key Principle 2: Proposed Change is Consistent with the Defense-in-Depth Philosophy

In RG 1.174, Revision 2, the NRC identified the following considerations used for evaluating how the licensing basis change is maintained for the philosophy of defense-in-depth:

A reasonable balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation.

Over-reliance on programmatic activities as compensatory measures associated with the change in the LB [licensing basis] is avoided.

System redundancy, independence, and diversity are preserved commensurate with the expected frequency, consequences of challenges to the system, and uncertainties (e.g.,

no risk outliers).

Defenses against potential common-cause failures are preserved, and the potential for the introduction of new common-cause failure mechanisms is assessed.

Independence of barriers is not degraded.

Defenses against human errors are preserved.

The intent of the plants design criteria is maintained.

RG 1.201, Revision 1, endorses the guidance in Section 6 of NEI 00-04 but notes that the containment isolation criteria in this section of the guidance are separate and distinct from those set forth in 10 CFR 50.69(b)(1)(x). The criteria in 10 CFR 50.69(b)(1)(x) are to be used in determining which containment penetrations and valves may be exempted from the Type B and Type C leakage testing requirements in both Options A and B of Appendix J to 10 CFR Part 50.

The criteria provided in 10 CFR 50.69(b)(1)(x) are not to determine the proper RISC category for containment isolation valves or penetrations.

In Section 3.1.1 of its letter dated February 17, 2023 (Reference 1), the licensee clarified that it will require an SSC to be categorized as HSS based on the defense-in-depth assessment performed in accordance with NEI 00-04. Based on the above, the NRC staff concludes that the proposed change is consistent with the defense-in-depth philosophy and, therefore, satisfies the second key principle for risk-informed decision-making prescribed in RG 1.174, Revision 2. The NRC staff finds that the licensee's process is consistent with the NRC-endorsed guidance in NEI 00-04 and would meet the 10 CFR 50.69(c)(1)(iii) criterion that requires defense-in-depth to be maintained.

3.2.3 Key Principle 3: Proposed Change Maintains Sufficient Safety Margins

The engineering evaluation that will be conducted by the licensee under 10 CFR 50.69 for SSC categorization will assess the design function(s) and risk significance of the SSC to assure that sufficient safety margins are maintained. The guidelines used for making that assessment will include ensuring that the categorization of the SSC does not adversely affect any assumptions or inputs to the safety analysis; or, if such inputs are affected, justification is provided to ensure that sufficient safety margin are maintained. With sufficient safety margins, (1) the codes and standards or their alternatives approved for use by the NRC are met and (2) safety analysis acceptance criteria in the LB (e.g., FSAR, supporting analyses) are met or proposed revisions provide sufficient margin to account for uncertainty in the analysis and data.

The SSCs design-basis function as described in the plants licensing basis, including the Updated Final Safety Analysis Report and TS Bases, do not change and should continue to be met. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. Based on the above, the NRC sta ff concludes that the licensee has established a program to ensure that sufficient safety margins are maintained in accordance with the third key principle of RG 1.174, Revision 2, and would, therefore, meet 10 CFR 50.69(c)(1)(iv).

3.3 Risk-Informed Evaluation

3.3.1 Key Principle 4: Change in Risk is Consistent with the Safety Goal Policy Statement

The risk-informed considerations prescribed in NEI 00-04 address the fourth key principle of the standards for risk-informed decision-making. These pertain to the assessment for change in risk.

A summary of how the licensees SSC categorizat ion process is consistent with the guidance and methodology in NEI 00-04, Revision 0, and RG 1.201, Revision 1, is provided in the sections below.

In Section 3.1.1 of its letter dated February 17, 2023 (Reference 1) and in the supplement dated November 2, 2023 (Reference 2), the licensee stated that the categorization process uses PRA-modeled hazards to assess risks for internal events, internal flooding, internal fires, and high winds. These are consistent with the approaches and methods included in NEI 00 -04. For the other risk contributors, the licensees process uses non-PRA methods to characterize the risk:

Seismic Hazard: An alternative seismic approach developed by EPRI is used. The approach for moderate seismic hazard/moderate seismic margin sites (Tier 2) is presented in EPRI Technical Update 3002017583, (Reference 9), hereafter referred to as the EPRI report.

Other External Hazards: A screening analysis performed for the Individual Plant Examination of External Events (IPEEE), (Reference 10) was updated using criteria from ASME/ANS RA-Sa-2009, (the ASME/ANS PRA Standard) (Reference 11) as endorsed by the NRC. More specifically, the update us ed criteria in Part 6, Requirements for Screening and Conservative Analysis of Other External Hazards At-Power.

Low Power and Shutdown Hazard: A Safe Shutdown Risk Management program is used consistent with NUMARC 91-06 (Reference 12).

Passive Components: A plant-specific method for the risk-informed categorization and treatment of passive components is proposed us ing the Arkansas Nuclear One, Unit 2 (ANO-2) plant-specific precedent (Reference 13). The ANO-2 plant-specific alternative was authorized for repair/replacement activi ties in ASME Boiler and Pressure Vessel Code Class 2 and 3 moderate-and high-energy systems.

The NRC staff evaluation of the proposed plant-specific use of the ANO -2 precedent for McGuire in the passive SSC categorization of passive components is provided in Section 3.3.1.2 of this safety evaluation.

3.3.1.1.1 Scope of the PRA

The McGuire PRA comprises a full-power internal events PRA, internal flooding PRA, fire PRA, and high-winds PRA. Each one of these assessments evaluates risk metrics of CDF and large early release frequency (LERF).

The NRC staff evaluated the scope of the PRA including: (1) peer-review history and results, (2) the independent assessment process including fact and observation (F&O) closure,

(3) credit for Diverse and Flexible Coping Strategies (FLEX) in the PRA, and (4) assessment of assumptions and approximations.

The NRC staff finds that the LAR, as supplement ed, provided the information necessary to support the staff review of the internal events PRA, internal flooding PRA, fire PRA, and high-winds PRA for technical acceptability, and therefore, meets the requirements set forth in 10 CFR 50.69(b)(2)(iii).

Peer-Review History for the Internal Events PRA and Internal Flooding PRA

In Section 3.3, PRA Review Process Results (10 CFR 50.69(b)(2)(iii)), of the enclosure to its letter dated February 17, 2023 (Reference 1), the licensee stated that the internal events PRA model was assessed in June 2015 against the ASME/ANS PRA Standard, RG 1.200, Revision 2, and NEI 05-04, Revision 3, (Reference 14). Resolved findings were reviewed and closed using the process in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, Close-out of Facts and Observations (F&Os). This was re-performed for the internal events PRA in May 2019 according to the process documented in the NEI letter to the NRC (Reference 15, Appendix X).

A subsequent finding closure review was conducted in November 2021 where resolved findings were reviewed and closed using the process documented in NEI 17 -07, Revision 2, (Reference 16).

The internal flooding PRA was assessed in September 2011 against the ASME/ANS PRA Standard, RG 1.200, Revision 2, and NEI 05-04. Resolved findings were reviewed and closed and for the internal flooding PRA in November 2018 using Appendix X. A subsequent finding closure review was conducted in June 2022 where resolved findings were reviewed and closed using the process documented in NEI 17 -07.

For both the internal events PRA and the internal flooding PRA, the licensee states in its letter dated February 17, 2023 (Reference 1), Enclosure, Section 3.2 Technical Adequacy Evaluation (10 CFR 50.69(b)(2)(ii)), and in Section 3.3 that there are no PRA upgrades that have not been peer reviewed. In its review, the NRC staff determined that all F&Os were assessed appropriately by the independent assessment team to assure that no newly developed methods or upgrades were inadvertently incorporated into the internal events PRA without a peer review in accordance with the ASME/ANS PRA standard as endorsed by the NRC.

Based on the above, the NRC staff concludes that the McGuire internal events PRA and internal flooding PRA were peer reviewed appropriately, consistent with RG 1.200, Revision 2, and that the F&Os were adequately closed.

Peer-Review History for the Internal Fire PRA

The licensees fire PRA underwent peer review in 2010. It was reviewed against the ASME/ANS PRA Standard, RG 1.200, Revision 2 and NEI 07-12, Revision 1 (Reference 17). Subsequent focused-scope peer reviews were performed to address various PRA upgrades. All of these were performed against the ASME/ANS PRA Standard, RG 1.200, Revision 2 and NEI 07-12 or NEI 17-07.

Independent assessment for closure of F&Os using Appendix X was performed in January 2019, December 2020, November 2021, and September 2022, which resulted in closure of all finding-level F&Os. Hence, the LAR does not identify any open finding-level F&Os.

The NRC staff has reviewed the fire PRA peer-review results and the licensee's resolution of the results and concludes that the McGuire fire PRA was peer reviewed appropriately, consistent with RG 1.200, Revision 2, and the F&Os were adequately closed.

Peer-Review History for the High-Winds PRA

In its letter dated November 2, 2023 (Reference 2), the licensee revised and updated the 10 CFR 50.69 LARs initial license condition to state that the high-winds PRA will be utilized in its 10 CFR 50.69 program. The licensee confirmed that the McGuire high-wind events PRA model received a full-scope peer review in October 2014 using the PRA Standard ASME/ANS RA-Sb-2013 (Reference 18). Although this version of PRA standard is not endorsed by RG 1.200, Revision 2, it was cited in the peer review report as part of this application. The issue of endorsement was resolved via a McGuire amendment issuance (Reference 19), which concluded that for high-winds PRA, there are no substantive differences between ASME/ANS RA-Sa-2009 and ASME/ANS RA-Sb-2013 for McGuire. Based on the above, the NRC staff finds that the later revision of the standard is acceptable for this application because for the licensees peer-review process, ASME/ANS RA-Sb-2013 also determined whether:

The methods used to develop the PRA are implemented correctly, The PRA represents the as-built and as-operated plant, The PRA assumptions and approximations are reasonable, and The licensee has procedures or guidelines in place for updating the PRA to reflect changes in plant design, operation, or experience.

Subsequent to the October 2014 peer review, an independent assessment for closure of F&Os was performed in December 2021, using the process documented in NEI 17 -07, which resulted in closure of all finding-level F&Os. Hence, the LAR does not identify any open finding-level F&Os. All PRA upgrades have been peer reviewed. In addition, there were no model assumptions or sources of uncertainty identified as key for high winds with respect to the 10 CFR 50.69 program.

Based on the above, the NRC staff finds that the McGuire high-winds PRA has been appropriately peer reviewed appropriately, the F&Os were closed using an NRC-approved approach, and the high-winds PRA model maintenance is acceptable. Therefore, the NRC staff concludes that the McGuire high-winds PRA is acceptable for use in the McGuire 10 CFR 50.69 program.

Appendix X, Independent Assessment Process for F&O Closure

Section X.1.3 of Appendix X to NEI 05-04/07-12/12-13 (Reference 15) provides guidance to perform an independent assessment for the closure of F&O identified from a full-scope or focused-scope peer review.

Based on its review of the LAR, as supplement ed, the NRC staff concluded that all F&Os were assessed appropriately by the independent assessment team to assure that no newly developed methods or upgrades were inadvertently incorporated into the internal events, internal flooding, fire, or high-winds PRA without a peer review in accordance with the ASME/ANS PRA standard as endorsed by the NRC in with RG 1.200, Revision 2. Therefore, the NRC staff finds that the McGuire internal even ts PRA, internal flooding PRA, fire PRA, and high-winds PRA were appropriately peer reviewed consistent with RG 1.200, Revision 2, and meet the requirements set forth in 10 CFR 50.69(c)(1)(i).

Assessment of Assumptions and Approximations/Identification of Key Assumptions and Sources of Uncertainty

In Section 3.2.7, PRA Uncertainty Evaluations, in its letter dated February 17, 2023 (Reference 1), the licensee states that NUREG -1855, Revision 1, was used to identify, screen, and characterize those sources of model uncertainty and related assumptions in the base PRA that are relevant to this application. Substep E -1.4, Qualitative Screening of Sources of Model Uncertainty and Related Assumptions, provides guidance on a qualitative screening process that involves identifying and validating whether consensus models were used in the PRA to evaluate identified model uncertainties. 4 The licensee confirmed that, for the McGuire uncertainty analysis, some uncertainties and assumptions were screened based on the use of a consensus method. The NRC staff finds that the assessment performed to identify the key assumptions/sources of uncertainty is c onsistent with the guidance provided in NUREG -1855.

Treatment of the Key Assumptions and Sources of Uncertainty

NUREG-1855, provides guidance regarding how to address PRA uncertainties to assure the risk-informed decision is in the context of the application for the decision under consideration. In Section 3.1.1 of its letter dated February 17, 2023 (Reference 1), the licensee confirmed that sensitivity studies will be performed consistent with NEI 00-04. In accordance with the guidance, the results of the sensitivity studies are given to the integrated decision-making panel (IDP) for consideration in the final risk characterization for components initially classified as LSS that may be reclassified to HSS. The NRC staff finds that t he licensee plans to perform sensitivity studies consistent with NEI 00-04 guidance to address the identified key assumptions and sources of uncertainty and to address PRA uncertainties in accordance with NUREG-1855 is acceptable.

The NRC staff recognizes that the licensee will perform routine PRA changes and updates, as stated in Reference 1, Section 3.2.6, PRA Maintenance and Updates, (), to ensure that the PRA continually reflects the as-built, as-operated plant, in addition to changes made to the PRA to support the context of the analysis being performed (i.e., sensitivities). Paragraphs 50.69(e) and (f) stipulate the process for feedback and adjustment to assure configuration control is maintained for these routine changes and updates to the PRA(s).

PRA Importance Measures and Integrated Importance Measures

The scope of modeled hazards for McGuire includes the internal events PRA, internal flooding PRA, fire PRA, and high-winds PRA. The NRC staff finds that the licensees use and treatment of importance measures is consistent with the guidance in NEI 00-04.

Credit for FLEX Equipment

The NRC memorandum dated May 6, 2022, Assessment of the Nuclear Energy Institute 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, Guidance for Risk-Informed Changes to Plants Licensing Basis (Reference 20), provides the NRC staffs assessment of challenges to incorporating FLEX into a PRA mo del in support of risk-informed decision-making in accordance with the guidance of RG 1.200, Revision 2.

4 Per NUREG-1855 (Reference 6) a consensus model is a model that has a publicly available published basis and has been peer reviewed and widely adopted by an appropriate stakeholder group.

In Section 3.0, Technical Evaluation, of its letter dated February 17, 2023 (Reference 1), the licensee states that PRA models used are the same as those described in its letter dated February 16, 2023 (Reference 21), for application to implement risk-informed completion times.

The licensee stated that, FLEX equipment is not incorporated into the [McGuire] PRA models.

Based on the above, the NRC staff finds that the McGuire internal events PRA, internal flooding PRA, fire PRA, and high-winds PRA do not credit FLEX equipment for the SSC categorization process.

PRA Acceptability Conclusions

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 PRA for internal events, internal flooding, fire, and high winds to support SSC categorization is endorsed by RG 1.201, Revision 1. The PRAs 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 acceptabili ty of the PRA by comparing the PRA to the relevant parts of the ASME/ANS PRA Standard using a peer-review process.

The licensee has subjected the PRAs for internal events, internal flooding, fire, and high winds to the peer-review processes and submitted the results of the peer review. The NRC staff reviewed the peer-review history (which incl uded the results and findings), the licensee's resolution of peer-review findings, and the identification and disposition of key assumptions and sources of uncertainty. The staff concludes that (1) the licensee's PRAs for internal events, internal flooding, fire, and high winds are acceptable to support the categorization of SSCs using the process endorsed by the NRC staff in RG 1.201, Revision 1, and (2) the key assumptions for the PRAs have been identified consistent with the guidance in RG 1.200, Revision 2, and NUREG-1855, Revision 1, and are addressed appropriately for this application.

The NRC staff finds the licensee provided the required information, and the internal events PRA, internal flooding PRA, fire PRA, and high-winds PRA are acceptable and, therefore, meet the requirements of10 CFR 50.69(c)(1)(i) and (ii).

3.3.1.2 Evaluation of the Use of Non-PRA Methods in SSC Categorization

The licensees categorization process uses the following non-PRA methods:

An alternative approach for moderate seismic hazard/moderate seismic margin sites (Tier 2), as described in the EPRI report. (Note: Reference 9 defines as Tier 2 plant as Plants where the [ground-motion response spectra] GMRS to [safe shutdown earthquake] SSE comparison between 1.0 Hz and 10 Hz is greater than in Tier 1 but not high enough to be treated as Tier 3. At these sites, the unique seismic categorization insights are expected to be limited

Screening analysis performed for the IPEEE (R eference 10) for other external hazards (e.g., external flood), updated using the external hazard screening significance process identified in the ASME/ANS PRA Standard (Reference 11).

Safe Shutdown Risk Management program consistent with NUMARC 91 -06 (Reference 12).

Passive Components: ANO-2 passive categorization (Reference 13).

The NRC staff's review of these methods is discussed below.

Alternative Seismic Approach

The licensee proposed using an alternative seismic approach for this LAR, (Section 3.2.3) which, for Tier 2 plants, has two important bases: (1) the impact of seismic risk in the categorization process due to the high relative cont ribution of seismic risk to the overall plant risk and (2) the conclusions from the case studies in the EPRI report (Reference 9).

In 3.2.3 of its letter dated February 17, 2023 (Reference 1), the licensee stated that its basis for McGuires classification as a Tier 2 plant is that consideration of the full range of the seismic hazard produces limited unique insights to the categorization process. The licensee stated that, based on the EPRI Report (Reference 9), the statements apply to McGuire:

At Tier 2 sites, there may be a limited number of unique seismic insights, most likely attributed to the possibility of seis mically correlated failures, appropriate for consideration in determining HSS SSCs. The special seismic risk evaluation process recommended using a Common Cause impact approach in the FPIE

[full-power internal events] PRA can identify the appropriate seismic insights to be considered with the other categorization insights by the Integrated Decision-making Panel for the final HSS determinations.

The licensee explained that the basis for using the proposed alternative seismic approach is that the special seismic risk evaluation process for the proposed approach can identify the appropriate seismic insights to be considered with the other categorization insights by the IDP for the final HSS determinations.

The licensee also stated that the proposed alternative seismic approach follows the same approach approved for LaSalle County Station (LaSalle) with two exceptions in that McGuire cites EPRI Report 3002017583 (Reference 9) which was an update to the prior EPRI Report that LaSalle cited with updates to bring their categorization program up to that in Reference 9 and McGuire did not include an unrelated request for additional information (RAI) response from the LaSalle 10 CFR 50.69 LAR RAI responses. The detailed technical review of this license amendments alternative seismic approach is detailed in Section 3.3.1.2 of this safety evaluation.

The licensee further stated that its proposed approach is specified in the EPRI report with the revision markups. These were included in LaSalles responses to requests for additional information supporting its LAR to implement 10 CFR 50.69 dated October 16, 2020, and January 22, 2021 (References 22 and 23, respectively), which LaSalle incorporated by reference into its application to implement the alternative seismic approach (Reference 24).

In Section 3.2.3 of its letter dated February 17, 2023 (Reference 1), the licensee stated that EPRI 3002017583 (Reference 9) is an update to EPRI 3002012988 (Reference 25). The latter report was referenced in Calvert Cliffs Nuclear Power Plants amendment issuance for 10 CFR 50.69 (Reference 26) which reflects, by reference, a Clinton Power Stations RAI response dated November 24, 2020 (Reference 27) which, in particular, addresses an assessment of the differences between EPRI 3002012988 and EPRI 3002017583.

To capture the potential impact of seismic risk in the categorization process, the licensees alternative seismic approach includes both quantitative and qualitative assessments of plant

SSC-specific seismic insights and their presentation to the IDP as part of its decision making.

The proposed approach includes focused walkdowns and quantification of PRA importance measures, based on a surrogate sensitivity study, for SSCs selected using the licensees internal events PRA. The proposed approach also includes consideration of seismic risk through insights from plant-specific seismic information.

Summary of Case Studies in the EPRI Alternative Seismic Method Report

The EPRI report (Reference 9) includes the results from case studies performed to determine the extent and type of unique HSS SSCs from seismic PRAs. The case studies were performed for four plants, designated as Plants A through D in the report. Description and evaluation of these case studies were documented in the NRC issuance of LaSalles 50.69 license amendment (Reference 28), which is incorporated by reference in the LAR supplement for this application.

Evaluation of the Information Provided for the Proposed Alternative Seismic Approach

In Section 3.2.3 of its letter dated February 17, 2023 (Reference 1), and in the supplement dated November 2, 2023 (Reference 2), the licensee provided (1) a description of its proposed Tier 2 alternative seismic approach for considering seismic risk in the categorization process and (2) how the proposed alternative seismic approach would be used in the categorization process. The licensee cited LaSalles 10 CFR 50.69 amendment issuance for its proposed alternative seismic approach (Reference 28). In addition, the licensee based the acceptability of its proposed alternative seismic approach on the conclusions drawn from case studies performed in the EPRI report (Reference 9) and, therefore, indirectly, on the acceptability of the PRAs used for the case studies.

The information on the proposed alternative seismic alternative - presented in the LARs Enclosure, in the supplement, and the EPRI report (Reference 9) - provides sufficient detail for how that information would be used in the categorization process and demonstrates that the quality and level of detail for that SSC categor ization is adequate. Based on the above, the NRC staff finds that the requirements in 10 CFR 50.69 (b)(2)(ii) are met for the licensees proposed alternative seismic approach.

The licensee submitted in its letter dated February 17, 2023 (Reference 1), as supplemented (Reference 2), along with the documents incorporated by reference, provides sufficient description and basis for evaluating acceptability to satisfy 10 CFR 50.69(b)(1)(iv). Based on the above, the NRC staff finds that the requirements in 10 CFR 50.69(b)(2)(iv) are met for the proposed alternative seismic approach at McGuire (i.e., the EPRI alternative seismic approach for Tier 2 plants).

Evaluation of Technical Acceptability of the PRAs Used for Case Studies Supporting the Proposed Alternative Seismic Approach

In the enclosure to of its letter dated February 17, 2023 (Reference 1), the licensee provided information concerning the case studies from the EPRI report (Reference 9), mapping approach, and conclusions on the determination of unique HSS SSCs used by the licensee to support its proposed alternative seismic approach. The key categorization conclusion from the Plants A, C, and D case studies is that the only SSCs identified as HSS in the seismic PRA (that were not also HSS from the internal events or fire PRA) were from unique, seismically induced failure modes. The remainder of HSS SSCs from the seismic PRA were captured by the

corresponding internal events or fire PRAs or other aspects of the categorization process in NEI 00- 04.

The licensee also stated that it used the case study information from the EPRI report (Reference 9) (termed test cases by the licensee). The licensee also incorporated by reference information related to the technical acceptability of the PRAs used (and the technical adequacy of certain technical details of the conduct of case studies) for case study Plants A, C, and D. The NRC staff reviewed and evaluated the technical acceptability of the PRAs used in the case studies for Plant A, C, and D for the McGuire LAR. The NRC staff also evaluated the peer-review process and resolution of peer-review findings. Finally, the NRC staff evaluated key assumptions and sources of uncertainty for Plants A, C, and D, which were incorporated by reference by the licensee.

The NRC staff finds that the technical acceptability of PRAs used for the Plant A, C, and D case studies in the EPRI report (Reference 9), the mapping approach used in those case studies, and the conclusions on the determination of unique HSS SSCs from the case studies are applicable to this licensees proposed plant-specific alternative seismic approach. Therefore, the staff concludes that the PRAs for Plants A, C, and D were technically acceptable and applicable for use in support of the licensees propos ed alternative seismic approach. The NRC staff determined the mapping of SSCs between the seismic PRA, the full-power internal events PRA (and, as applicable, the fire PRA) for the Plant A, C, and D case studies was appropriate. The NRC staff finds that the licensees plant-specif ic evaluation is sufficient to determine unique HSS SSCs from seismic PRAs equivalently to the same process utilized in the Plant A, C, and D case studies from the EPRI report.

Evaluation of the Applicability of the Criteria for the Proposed Alternative Seismic Approach

In response to the 10 CFR 50.54(f) letter associated with post-Fukushima Near-Term Task Force (NTTF) Recommendation 2.1 (Reference 29), the licensee submitted its seismic hazard screening report in a letter dated March 20, 2014 (Reference 30). The NRC staff evaluation of the licensees submittal (Reference 31) included confirmatory analysis of the seismic hazard and concluded that the licensees seismic hazard screening report was responsive to the 10 CFR 50.54(f) request and that the GMRS determined by the licensee adequately characterizes the reevaluated hazard for the McGuire site.

In its letter dated February 17, 2023 (Reference 1), the licensee provides its basis for McGuire being a Tier 2 plant as states, in part, that,

As defined in Reference 17, McGuire meets the Tier 2 criteria for a "Moderate Seismic Hazard/Moderate Seismic Margin" site. The Tier 2 criteria are as follows:

"Tier 2: Plants where the GMRS to SSE comparison between 1.0 Hz and 10 Hz is greater than in Tier 1 but not high enough to be treated as Tier 3.

At these sites, the unique seismic categorization insights are expected to be limited."

Note: Reference 17 applies to the Tier 2 sites in its entirety except for Sections 2.2 (Tier 1 sites) and 2.4 (Tier 3 sites).

For comparison, Tier 1 plants are defined as having a GMRS peak acceleration at or below approximately 0.2g or where the GMRS is below or approximately

equal to the SSE between 1.0 Hz and 10 Hz. Tier 3 plants are defined where the GMRS to SSE comparison between 1.0 Hz and 10 Hz is high enough that the NRC required the plant to perform a seismic PRA (SPRA) to respond to the Fukushima 10 CFR 50.54(f) letter (Reference 20). The NRC did not require McGuire to perform an SPRA as stated in its revised seismic screening and prioritization letter dated December 22, 2016 (Reference 35):

"the NRC has determined that seismic probabilistic risk assessments (SPRAs) for Catawba and McGuire are no longer necessary to fulfill the March 12, 2012, request for information pursuant to Title 10 of the Code of Federal Regulations, Part 50, Section 50.54(f) (ML12053A340)."

In Section 3.2.3, Seismic Hazards, of its letter dated February 17, 2023 (Reference 1), the licensee stated that the licensee compared the reevaluated GMRS with the sites design-basis SSE to demonstrate that the plant meets the criteria for application of the proposed alternative seismic approach.

Since the same hazard is used for comparison against the criteria for use of the proposed alternative seismic approach, the NRC staffs pr evious assessment on the reevaluated hazard is applicable to this review. The NRC staff finds that the plants GMRS is above the Tier 1 criteria.

Further, the NRC staffs review confirmed that the licensee did not perform a seismic PRA as part of the NRCs post-Fukushima actions.

In summary, the NRC staff finds that the licen sees basis for applying the proposed alternative seismic approach to its site is acceptable because the licensee meets the Tier 2 criteria for use of the proposed alternative seismic approach based on its seismic hazard screening report.

Evaluation of the Implementation of the Proposed Alternative Seismic Approach

The categorization conclusions from the case studies in the EPRI report (Reference 9) indicate that seismic-specific failure modes resulted in HSS categorization uniquely from seismic PRAs.

Therefore, such seismic-specific failure modes, such as correlated failures, interaction failures, relay chatter, and passive component structural failure modes can influence the categorization process. The licensee discussed the implementation of its alternative seismic approach in Reference 2, Section 3.2.3, Seismic Hazards. The NRC staff reviewed this information to evaluate whether the categorization-related to conclusions from the EPRI report were appropriately included and implemented.

The proposed alternative seismic approach includes a combination of qualitative and quantitative considerations of the mitigation capabilities as well as seismic failure modes of SSCs in the categorization process. These considerations are based on plant-specific walkdowns for the SSCs undergoing categorization, quantification of the impact of seismic failure of SSCs subject to correlated or interaction failures, and insights obtained from prior seismic evaluations performed for McGuire.

Qualitative Evaluation for the Alternative Seismic Approach

In its letter dated February 17, 2023 (Reference 1), Section 3.1.1, the licensee stated that in applying the alternative seismic approach, a categorization team will prepare a system categorization document (SCD). At several steps of the process, the categorization team will consider the available seismic insights relative to the system being categorized and document

their conclusions in the SCD. When the SCD is presented to the IDP, a section of the document will provide the basis for the proposed alternative seismic approach including the seismic hazard for the plant and the criteria for use of the proposed alternative seismic approach. The IDP will be informed of plant SSC-specific seismic in sights that it may choose to consider in its deliberations on the categorization of SSCs as LSS or HSS. The licensee stated in Reference 2, Section 3.1.1, Overall Categorization Process, that the categorization evaluation for seismic hazard would be performed at the function level, the component level, or both using the alternative seismic approach.

In its letter dated February 17, 2023 (Reference 1), the licensee states, in part, that.

The categorization team will review available MNS [McGuire] plant-specific seismic reviews and other resources such as those identified above. The objective of the seismic review is to identify plant-specific seismic insights that might include potentially important impacts such as:

Impact of relay chatter

Implications related to potential seismic interactions such as with block walls

Seismic failures of passive SSCs such as tanks and heat exchangers

Any known structural or anchorage issues with a particular SSC

Components implicitly part of PRA-modeled functions (including relays).

For each system categorized, the categorization team will evaluate correlated seismic failures and seismic interactions between SSCs.

The licensee further explained that these insights would provide the IDP a means to consider potential impacts of seismic events in the categorization process. The licensee stated, in part,

that,

Seismic impacts would be compiled on an SSC basis. As each system is categorized, the system-specific seismic insights will be documented in the categorization report and provided to the IDP for consideration as part of the IDP review process. The IDP cannot challenge any candidate HSS recommendation for any SSC from a seismic perspective if they believe there is a basis, except for certain conditions identified in Step 10 of Section 2.3.1 of Reference 17. Any decision by the IDP to downgrade preliminary HSS components to LSS will consider the applicable seismic insights in that decision.

The licensee explained that sources of the insights related to seismic events would be prior plant-specific seismic evaluations, including those performed in response to recommendations by the post-Fukushima Near-Term Task Force: the seismic hazard screening, spent fuel pool assessment, expedited seismic evaluation process, and seismic high frequency evaluation performed for NTTF Recommendation 2.1; seismic walkdowns performed for NTTF Recommendation 2.3; and the seismic mitigation strategy assessment performed for NTTF Recommendation 4.2 (Reference 32).

In Section 3.2.3 of its letter dated February 17, 2023 (Reference 1), the licensee states, in part,

that,

For components that are HSS due to fire PRA but not HSS due to internal events PRA, the categorization team will review design-basis functions during seismic events or functions credited for mitigation and prevention of severe accidents caused by seismic events and characterize these for presentation to the IDP as additional qualitative inputs, which will also be described in the SCD.

Based on its review of the qualitative evaluations for seismic risk in the licensees proposed alternative seismic approach, the NRC staff finds t hat: (1) the evaluations will include potentially important seismically induced failure modes, as well as mitigation capabilities of SSCs during seismically induced design-basis and severe accident events, consistent with the conclusions on the determination of unique HSS SSCs from seismic PRAs in the EPRI report (Reference 9),

(2) the licensee will provide system-specific qualitative seismic insights to the IDP for consideration as part of the IDP review process as each system is categorized, (3) the insights will use plant-specific prior seismic evaluations, which, in conjunction with the performance monitoring for the proposed alternative seismic approach, reasonably reflect the current plant configuration, and (4) the qualitative evaluation will complement focused walkdowns and quantitative evaluations identified for the SSCs. Further, the recommendation for categorizing civil structures in the proposed alternative seismic approach as discussed in the EPRI report provides appropriate consideration of such failures from a seismic event and that consideration is evaluated by the staff in the section titled, Quantitative Evaluation for the Alternative Seismic Approach.

Focused Walkdowns for the Alternative Seismic Approach

In Section 3.2.3 of its letter dated February 17, 2023 (Reference 1), the licensee stated that the proposed alternative seismic approach includes focused seismic walkdowns of SSCs undergoing categorization. The purpose of the walkdowns is to identify, for the SSCs that are being categorized, the conditions for occurrence of correlated failures, failure of more than one SSC due to interactions with other SSCs, and single component failures.

The NRC staff evaluated the focused walkdowns for the proposed alternative seismic approach, as described in the enclosure to the LAR and in the EPRI report (Reference 9), including the revision markups in the LaSalle 10 CFR 50.69 LAR supplements (References 22, 23, and 30) that were incorporated by reference in this LAR. The licensee cited the LaSalle 10 CFR 50.69 approval and stated that its proposed approach followed that precedent with two exceptions related to the EPRI report and plant-specific responses to requests for additional information.

The NRC staffs review of the supplement to the LAR and the documents incorporated by reference in the LAR determined that the staffs previous evaluation documented in the LaSalle 10 CFR 50.69 safety evaluation is applicable to McGuire.

The NRC staffs review of the focused walkdowns in the proposed alternative seismic approach described in the EPRI report (Reference 9), including the revision markups in the LaSalle 10 CFR 50.69 LAR supplements that are incorporated by McGuire in this LAR, finds the following:

The licensees focused walkdown in the proposed alternative seismic approach:

(i) includes consideration of seismically induced, seismically correlated, and seismic interaction failures that fail more than one SSC as well as single component failures;

(ii) includes evaluations of the direct and indirect impacts of seismically induced, seismically correlated, and seismic interaction failure of an SSC; (iii) reflects the insights from the case studies in the EPRI report (Reference 9) in addressing these failure modes; and (iv) confirms that the modifications to the proposed alternative seismic approach through changes to the EPRI report appropriately reflect the evaluation of such direct and indirect impacts.

The qualification of personnel performing the walkdowns as well as documentation and retention of the walkdown results is acceptable for the proposed alternative seismic approach. The qualification of personnel performing the walkdowns for the proposed alternative seismic approach is consistent with the state-of-practice for development and peer review of contemporary seismic PRAs and the documentation. Additionally, retention of walkdown information for the proposed alternative seismic approach is consistent with state-of-practice seismic PRAs, and the guidance in NEI 00 -04 will result in appropriate information being presented to the IDP for categorization decisions.

(These modifications are addressed in the quantitative evaluation section below.)

The licensees approach for selecting the screening criterion is consistent with the state-of-practice seismic PRAs and that SSCs screened out based on the criterion are not expected to result in HSS components within the 10 CFR 50.69 categorization process.

The fragility approaches proposed by the licensee for development of fragility values in Step 5b are acceptable for the proposed alternative seismic approach because (i) they represent state-of-practice approaches consistent with those used in contemporary seismic PRAs reviewed by the NRC staff, and (ii) no unreviewed methods would be used for fragility calculations.

The personnel performing fragility evaluations for the proposed alternative seismic approach will have experience or background consistent with that used for state-of-practice seismic PRAs as well as the guidance in NEI 00 -04 on personnel qualifications and the use of such personnel is, therefore, acceptable for the proposed alternative seismic approach. In addition, the NRC staff review determined that the documentation of the fragility evaluations will be consistent with the documentation used for other categorization processes and is, therefore, acceptable for the proposed alternative seismic approach.

The proposed alternative seismic approach will result in consideration of relays as implicitly modeled components and of insights related to the impact of seismically induced relay chatter for the function achieved by the SSC during the categorization.

The focused walkdowns of SSCs undergoing categorization will identify seismic interaction and correlated failures, including those resulting from potential failures of passive components, as well as structural and anchorage issues. Further, the NRC staff concludes that insights from available plant-specific seismic reviews will also provide categorization-related insights from a seismic failure modes perspective.

Quantitative Evaluation for the Alternative Seismic Approach

In Section 3.2.3 of its letter dated February 17, 2023 (Reference 1), the licensee explained that SSCs identified as being subject to seismic correlation or interaction failure modes based on the walkdown would be subjected to a quantitative evaluation using the licensees internal events

PRA to determine the impact of seismic events on the categorization. The licensee further explained that the quantitative evaluation would be performed through a sensitivity study, including a sensitivity to surrogate events, us ing the licensees internal events PRA. The NRC staff notes that further details on the surrogate sensitivity are provided in Section 2.3.1 of the EPRI report (Reference 9), including the revision markups in the LaSalle 10 CFR 50.69 LAR supplements that are incorporated by reference in McGuires LAR.

The surrogate sensitivity would be performed by selecting PRA basic events, termed surrogate events, in the licensees internal events PRA. These events would be inserted at appropriate locations to reflect seismically induced or correlated failure, or seismic interaction failure, of single or multiple SSCs. Subsequently, the modified internal events PRA with the surrogate events would be quantified for the loss-of-offsite power (LOOP) and small break loss-of-coolant accident (LOCA) (hereafter referred to as sm all LOCA) initiators and importance measures would be derived. The importance measures for the surrogate events derived from this sensitivity study would be used to identify the SSCs that should be HSS due to seismically correlated failures or seismic interaction related failures. The licensee further stated that the quantitative evaluation to determine the importance of SSCs on a system basis in the proposed alternative seismic approach is detailed in Section 2.3.1 of the EPRI report (Reference 9).

The NRC staff reviewed the quantitative evaluation for the alternative seismic approach described in the LAR, its supplement, and the EPRI report (Reference 9), including the revision markups in the LaSalle 10 CFR 50.69 LAR supplements (References 22, 23, and 33). the NRC staff's evaluation determined that the licensee has made a sufficient plant-specific case for McGuire to adopt applicable portions of the technical evaluation of the LaSalle 10 CFR 50.69 safety evaluation (Reference 28).

The NRC staff determined that seismically induced LOOP and small LOCA occurrence frequencies are representative for McGuire based on the three seismic PRAs in the case studies in the EPRI report (Reference 9) and the fact that the seismic hazard at the licensees site is lower than the hazard for the seismic PRAs for Plants A, C, and D. Therefore, the NRC staff concludes that the proposed occurrence frequency for the seismically induced LOOP event of 1.0 per year, the proposed occurrence frequency for the seismically induced small LOCA event of 10-2 per year, and the proposed surrogate event failure probability of 10 -4 are acceptable for use in the licensees alter native seismic approach. Further, the NRC staff determined that changing the occurrence frequency and surrogate event failure probability in the surrogate sensitivity analysis is acceptable for the licensees alternative seismic approach because: (1) it is necessary for developing the importance measures for comparison against the corresponding thresholds in NEI 00 -04, and (2) it does not alter the basis for the proposed values. Based on the above, the NRC staff finds that there is reasonable assurance that the categorization outcome from the licensees proposed alternative seismic approach will be comparable to those from seismic PRAs.

Implementation of the Alternative Seismic Approach

Based on its review of the licensees proposed alternative seismic approach described in letter dated February 17, 2023 (Reference 1), as supplemented by letter dated November 2, 2023 (Reference 2) the NRC staff finds that the use of the proposed alternative seismic approach will ensure the acceptability of the evaluations required by 10 CFR 50.69(c)(1)(ii) and (iv) because:

The approach includes qualitative consideration of seismic events at several steps of the categorization process, including documentation of the information for presentation to the IDP as part of the integrated, systematic process for categorization.

The approach includes focused walkdown(s) which evaluate(s) the direct and indirect impacts of seismically induced or correlated failures, seismic interaction failures, and single component failures in a system under categorization.

The approach includes a quantitative evaluation, with justified failure probability and initiating event frequencies, that provides reasonable assurance that the categorization results from the licensees proposed alternative seismic approach will be similar to those from seismic PRAs.

Personnel performing necessary walkdowns and analyses will have qualifications consistent with the state-of-practice for seismic PRAs and the guidance in NEI 00 -04.

The documentation of these walkdowns and analyses will be consistent with the state-of-practice for seismic PRAs and the guidance in NEI 00 -04.

The quantitative and qualitative insights presented to the IDP include potentially important seismically induced failure modes as well as mitigation capabilities of SSCs during seismically induced design-basis and severe accident events, consistent with the conclusions on the determination of unique HSS SSCs from seismic PRAs in the EPRI report (Reference 9) and with the markups provided in the LaSalle 10 CFR 50.69 LAR supplements (References 22, 23, and 30), which were incorporated by reference by the licensee in this LAR. The quantification will use the licensees internal events PRA and the insights will use prior plant-specific seismic evaluations. Therefore, in conjunction with performance monitoring for the proposed alternative seismic approach, the proposed alternative seismic approach will reasonably reflect the current plant configuration.

The approach presents system-specific insights and categorization results from a seismic risk perspective to the IDP for consideration as part of the IDP review process, thereby providing the IDP with a means to consider potential impacts of seismic events in the categorization process. This process includes a categorization team that per its letter dated February 17, 2023 (Reference 1), Section 3.1.1, the licensee states that the team will,

provide preliminary assessments of the seven considerations for the IDPs,

[and] that [if] one or more of the seven considerations cannot be confirmed, then that function is presented to the IDP as preliminary HSS.

Conversely, if all the seven considerati ons are confirmed, then the function is presented to the IDP as preliminary LSS.

The approach presents the IDP with the basis for using the proposed alternative seismic approach while considering that McGuire is a seismic Tier 2 plant ((aka, a moderate seismic hazard, (Ref. 1, Sec. 3.1.1, page 7) plant and the criteria for use of the proposed alternative seismic approach.

Evaluation for Performance Monitoring for the Alternative Seismic Approach

In Section 3.5, Feedback and Adjustment Process of its letter dated February 17, 2023 (Reference 1), the licensee stated that its configuration control process ensures that changes to the plant, including a physical change and changes to documents, are evaluated to determine the impact on drawings, design bases, licensing documents, programs, procedures, and training.

The NRC staff evaluated the licensees discussion of its performance monitoring program for the proposed alternative seismic approach to ensure: (1) the continued validity of the plant-specific information that is developed for each SSC categorized, (2) that any changes to the plant, including the seismic hazard, are captured and appropriately addressed as part of the 10 CFR 50.69 program, and (3) that the requirements in 10 CFR 50.69(e) were met for the proposed alternative seismic approach.

In Section 3.5 of the enclosure to the LAR, the licensee stated that its performance monitoring process requires periodic review to assess changes that could impact the categorization results and to provide the IDP with an opportunity to recommend categorization and treatment adjustments due to such changes. The licensee explained that its configuration control program had been updated to have a checklist related to the impact of seismic events on categorization.

The licensee identified some of the items in the checklist in Section 3.5 of the enclosure to the LAR.

In Section 3.5 of its letter dated February 17, 2023 (Reference 1), the licensee states, in part,

that,

The Duke Energy 10 CFR 50.69 program requires that SCDs cannot be approved by the IDP until the panels comments have been resolved to the satisfaction of the IDP. This includes issues related to system-specific seismic insights considered by the IDP during categorization.

Scheduled periodic reviews are completed at least once every two refueling cycles in accordance with Duke Energy procedures and will evaluate new insights resulting from available risk information (i.e., PRA model or other analysis used in the categorizati on) changes, design changes, operational changes, and SSC performance. If it is determined that these changes have affected the risk information or other elements of the categorization process such that the categorization results are more than minimally affected, then the risk information and the categorization process will be updated. This review will include:

A review of plant modifications since the last review that could impact the SSC categorization A review of plant-specific operating experience that could impact the SSC categorization A review of the impact of the updated risk information on the categorization process results A review of the importance measures used for screening in the categorization process An update of the risk sensitivity study performed for the categorization

Input from Regulatory Affairs and Operations regarding changes that may affect the bases for the categorization results.

In addition to the normally scheduled periodic reviews, if a PRA model or other risk information is upgraded, a review of the SSC categorization will be performed.

The NRC staff recognizes that the seismic hazard at any site could potentially increase such that the categorization process may be impacted from a seismic risk perspective, either solely due to the seismic risk or via the integrated importance measure determination. In Section 3.2.3 of the enclosure to the LAR, the licensee stated that if the McGuire seismic hazard changed at some future time, and if its feedback process determines that a process different from the proposed alternative seismic approach is warranted for seismic risk consideration in categorization under 10 CFR 50.69, it will seek prior NRC approval for use of such an approach.

The NRC staff notes that seeking prior NRC approval for the use of a process different from the proposed alternative seismic approach and the previously approved seismic margin assessment is consistent with the new license condition proposed by the licensee, as stated by the licensee in Reference 2, Section 3.2.3. The licensee further stated that, after receiving NRC approval, it will follow its categorization review and adjustment process to review the changes to the plant and update, as appropriate, the SSC categorization in accordance with 10 CFR 50.69(e).

Based on the above, the NRC staff finds that the licensees configuration control program includes consideration of seismic issues as well as failure modes, such as interaction between components, and a review of seismic loading and seismic dynamic qualification. Further, the licensees performance monitoring program asse sses changes that impact the categorization results and provides the IDP with an opportunity to recommend categorization and treatment adjustments due to such changes. Therefore, the NRC staff finds that the licensees performance monitoring and configuration control process addresses plant-specific seismic evaluation, thereby ensuring that the corresponding impacts on SSC categorization continues to remain valid and if necessary, are presented to the IDP for consideration of categorization changes.

During its review, the NRC staff noted that the licensees performance monitoring program for 10 CFR 50.69 has the capability to identify significant changes to the plant risk profile as well as instances in which a RISC -3 or RISC-4 SSC may fail to perform a safety significant function, resulting in an immediate evaluation and review for such instances. Based on its review, the staff finds that the requirements in 10 CFR 50.69(e) are met for the proposed alternative seismic approach.

Conclusion for Proposed Alternative Seismic Approach

Based on the above, the NRC staff concludes that the licensees proposed alternative seismic approach for McGuire, as described in the licensees LAR, as supplemented, is acceptable for considering seismic risk in the licensees categorization process under 10 CFR 50.69.

High-Winds Risk

In Attachment 4 of the enclosure to the LAR, the licensee screened the extreme winds and tornadoes based on the criteria of PS4 (the bounding mean CDF is < 10 -6/year) and C4 (the event is included in the definition of another event, per Reference 11 which is used only to include within another event). During the audi t, the NRC staff requested that the licensee demonstrate that all HSS SSCs identified from high-winds PRA are covered by using the

full-power internal events, internal flooding, or fire PRAs. This confirmed that no additional HSS SSCs can be identified using the high-winds PRA model. In the supplement (Reference 2), the licensee revised the 10 CFR 50.69 LAR to include provision for the high-winds PRA to be used in its 10 CFR 50.69 program. This is reflected in the wording of the first paragraph of the proposed license condition that approves the implementation of 10 CFR 50.69.

Other External Hazards

This hazard category includes all non-seismic external hazards such as external floods, transportation, and nearby facility accidents. In Section 3.2.4, Other External Hazards, of its letter dated November 2, 2023 (Reference 2), the licensee stated that all external hazards, except for seismic and high winds, were screened for applicability to McGuire per a plant-specific evaluation. This evaluation was conducted in accordance with Generic Letter (GL) 88-20, Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f), Supplement 4 (Reference 10). It was updated to use the criteria in the ASME/ANS PRA Standard (Reference 11).

In Attachment 4 of the enclosure to the LAR, the licensee stated that the doors and barriers, which are credited to allow the external flooding and local intense precipitation hazards to be screened, will be considered HSS in accordance with NRC-approved guidance. The NRC staff finds that the licensee external flooding screening is acceptable because it has an insignificant impact and the credited SSCs were categorized as HSS, consistent with NEI 00 -04 (Reference 8).

Based on the above, the NRC staff finds that the licensees assessment of all other external hazards is consistent with Section 5 of NEI 00 -04 and is, therefore, acceptable for this application and meets the requirements of 10 CFR 50.69(c)(1)(ii).

Shutdown Risk

Consistent with the guidance in NEI 00-04, the licensee proposed using a shutdown safety assessment based on NUMARC 91 -06 (Reference 12). NUMARC 91 -06 provides considerations for maintaining defense-in-depth for the five key safety functions during shutdown, namely, decay heat removal capability, inventory control, power availability, reactivity control, and containment-primary/secondary. NUMARC 91-06 also specifies that a defense-in-depth approach should be used with respect to each defined shutdown key safety function. This is accomplished by designating a running and an alternative system/train to accomplish the given key safety function.

The use of NUMARC 91 -06, as described by the licensee in its submittal, is consistent with the guidance in NEI 00-04 (Reference 8) as endorsed by the NRC in RG 1.201 (Reference 3). The approach uses an integrated and systematic proce ss to identify HSS components, consistent with the shutdown evaluation process. Based on the above, the NRC staff finds that the licensee's use of NUMARC 91 -06 is acceptable, and meets the requirements set forth in 10 CFR 50.69(c)(1)(ii).

Component Safety Significance Assessment for Passive Components

Passive components are not modeled in the PRA; therefore, a different assessment method is necessary to assess the safety significance of these components. For the purposes of 10 CFR 50.69 categorization, passive components are those components having only a

pressure retaining function. This process also addresses the passive function of active components such as the pressure/liquid retention of the body of a motor-operated valve.

In Section 3.1.2, Passive Categorization Process, of its letter dated February 17, 2023 (Reference 1), the licensee proposed using a plant-specific method for the risk-informed categorization precedent and treatment of passive components that was not cited in NEI 00-04.

However, this plant-specific authorization was authorized as an alternative for Arkansas Nuclear One, Unit 2 (ANO-2) (Reference 13). The ANO -2 plant-specific alternative for repair/replacement activities in ASME Class 2 a nd 3 pressure retaining items and their supports (exclusive of Class CC and MC items), using a modification of the ASME Code Case N -660 (Reference 34). The ANO-2 plant-specific authorization relies on the conditional core damage and large early release probabilities associated with pipe ruptures. Safety significance of an SSC is generally measured by the frequency and the consequence of its failure, in this case, pipe ruptures. Treatment requirements (includi ng repair/replacement) only affect the frequency of passive component failure. Categorizing solely based on consequences, which measures the safety significance of the pipe given that it ruptures, is conservative compared to including the rupture frequency in the categorization. The categorization will not be affected by changes in frequency arising from changes to the treatment. Based on its review of the McGuire plant-specific request and applicability, the NRC staff finds that the use of the ANO-2 repair/replacement precedent is acceptable for categorization of passive ASME Boiler and Pressure Vessel Code Class 2 and 3 moderate-and high-energy systems.

In Section 3.1.2 of the enclosure to the LAR, the licensee states, in part, that,

The passive categorization process can apply the same risk-informed process accepted by the NRC in Reference 5 for the passive categorization of Class 2, 3, and non-class components. This is the same passive SSC scope the NRC has conditionally endorsed in ASME Code Cases N-660 and N-662 as published in Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Revision 15. Both code cases employ a similar risk-informed safety classification of SSCs in order to change the repair/replacement requirements of the affected LSS components. All ASME Code Class 1 SSCs with a pressure retaining function, as well as supports, will be assigned HSS for passive categorization which will result in HSS for its risk-informed safety classification and cannot be changed by the IDP.

Based on the above, the N RC staff finds the licensees proposed plant-specific approach for McGuire for the categorization and treatment of passive Class 2 and 3 SSCs to satisfy the 10 CFR 50.69 SSC categorization process.

Risk Sensitivity Study (NEI 00-04, Section 8)

In Section 3.1.1,Overall Categorization Process, of its letter dated February 17, 2023 (Reference 1), the licensee states that an unreliability factor of three will be used for the sensitivity studies described in Section 8, Risk Sensitivity Study, of NEI 00-04. Section 3.2.7, PRA Uncertainty Evaluations, of the LAR further confirms that a cumulative sensitivity study will be performed where the failure probabilities (unreliability and unavailability, as appropriate) of all LSS components modeled in PRAs for all systems that have been categorized are increased by a factor of three. The NRC staff finds the application of a factor of three for the sensitivities is consistent with the guidance in NEI 00-04.

In Section 3.1.1 of its letter dated February 17, 2023 (Reference 1) the licensee cites NEI 00-04 and quoted RG 1.201 (Reference 3), which states that the implementation of all processes described in NEI 00-04 ([Reference 8,] i.e., Sections 2 through 12) is integral to providing reasonable confidence and that all aspects of NEI 00-04 must be followed to achieve reasonable confidence in the evaluations required by 10 CFR 50.69(c)(1)(iv). This sensitivity study, together with the periodic review process assures that the potential cumulative risk increase from the categorization is maintained acceptably low. The performance monitoring process monitors the component performance to ensure that potential increases in failure rates of categorized components are detected and addressed before reaching the rate assumed in the sensitivity study.

The NRC staff finds that the licensee will perform the risk sensitivity study consistent with the guidance in Section 8 of NEI 00-04, and, therefore, will assure that the potential cumulative risk increase from the categorization is maintained acceptably low, as required by 10 CFR 50.69(c)(1)(iv).

Integrated Decision-Making

NRC Standard Review Plan (NUREG-0800), Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance, provides guidance to the NRC staff for the revi ew of risk-informed changes to the licensing basis including an integrated decision-making process. The Appendix states, in part, that

[r]isk-informed applications are expected to require a process to integrate traditional engineering and probabilistic considerations to form the basis for acceptance. Guidance in NEI 00-04 identifies two steps in the categorization process: (1) Preliminary Engineering Categorization of Function and (2) IDP Review and Approval that are responsible for the integrated assessment of the traditional engineer ing analyses and the risk results from the PRA and non-PRA assessments that are performed to make a determination and approval of the safety significance of the SSC for categorizat ion. The NRC staff reviewed these two steps to ensure that the processes are well-defined, systematic, repeatable, and scrutable.

Preliminary Engineering Categorization of Function (NEI 00-04, Section 7)

In Section 3.1.1 of its letter dated February 17, 2023 (Reference 1), the licensee stated, in part, that if any SSC is identified as HSS from either the integrated PRA component safety significance assessment (Section 5 of NEI 00-04) or the defense-in-depth assessment (Section 6), the associated system function(s) would be identified as HSS. The licensee also stated that Once a system function is identified as HSS, then all the components that support that function are preliminary HSS.

Based on the above, the NRC staff finds that the preliminary categorization of functions is consistent with NEI 00-04 (Reference 8), as endorsed by the NRC in RG 1.201 (Reference 3),

and is, therefore, acceptable.

IDP Review and Approval (NEI 00-04, Sections 9 and 10)

In Section 3.1.1 of its letter dated February 17, 2023 (Reference 1), the licensee states that,

The IDP will be composed of a group of at least five experts who collectively have expertise in plant operation, design (mechanical and electrical) engineering, system engineering, safety analysis, and PRA. At least three members of the IDP

will have a minimum of five years of experience at the plant, and there will be at least one member of the IDP who has a minimum of three years of experience in the modeling and updating of the plant-specific PRA."

The IDP will be trained in the specific technical aspects and requirements related to the categorization process. Training will address at a minimum the purpose of the categorization; present treatment requirements for SSCs including requirements for design basis events; PRA fundamentals; details of the plant-specific PRA including the modeling, scope, and assumptions, the interpretation of risk importance measures, and the role of sensitivity studies and the change-in-risk evaluations; and the defense-in-depth philosophy and requirements to maintain this philosophy.

The guidance in NEI 00-04, as endorsed by the NRC in RG 1.201, provides confidence that the IDP expertise is sufficient to perform the categorization and that the results of the different evaluations (PRA and non-PRA) are used in an integrated, systematic process as required by 10 CFR 50.69(c)(1)(ii). Based on the above, the NRC staff finds that the licensee's proposed IDP would have the expertise to meet the requirements in 10 CFR 50.69(c)(2) and that the additional descriptions of the IDP characteristics, training, processes, and decision guidelines are consistent with NEI 00-04 as endorsed by the NRC in RG 1.201.

As discussed in NEI 00-04, the only requirements that are relaxed for RISC -3 SSCs are those related to treatment, not design or capability. Also, 10 CFR 50.69(d)(2)(i) requires the licensee to conduct periodic inspection and testing activities to ensure, with reasonable confidence, that RISC-3 SSCs remain capable of performing their safety-related functions under design-basis conditions. Therefore, the NRC staff finds that the IDP for the McGuire categorization process is consistent with the endorsed guidance in NEI 00-04 and, therefore, fulfills the requirements of 10 CFR 50.69(c)(1)(iv).

Based on the above, the NRC staff review finds that for: (1) internal events PRA, internal flooding PRA, fire PRA, and high-winds PRA acceptability, (2) PRA importance measures and integrated importance, (3) evaluation of the use of non-PRA methods, (4) risk sensitivity study, and (5) integrated decision-making, that the proposed change satisfies the fourth key principle for risk-informed decision-making prescribed in RG 1.174 (Reference 5).

3.3.2 Key Principle 5: The impact of the proposed change should be monitored by using performance measurement strategies.

NEI 00-04 (Reference 8) provides guidance that includes programmatic configuration control and a periodic review to ensure that all aspects of the 10 CFR 50.69 program (i.e., includes traditional engineering analyses) and PRA models used to perform the risk assessment continue to reflect the as-built, as-operated plant and that plant modifications and updates to the PRA over time are continually incorporated.

Programmatic Configuration Control (NEI 00-04, Sections 11 and 12)

Sections 11 and 12 of NEI 00-04, include discussion on periodic review; and program documentation and change control. Maintaining change control and periodic review will also maintain confidence, that all aspects of the 10 CFR 50.69 program and risk categorization for SSCs, continually reflect the McGuire as-built, as-operated plant. A more detailed NRC staff review is provided as follows.

Periodic Review (NEI 00-04, Section 12)

Section 50.69(e), Feedback and process adjustment, of 10 CFR requires that periodic updates to the licensees PRA and SSC categorization must be performed. Changes over time to the PRA and to the SSC reliabilities are inevitable and such changes are recognized by the 10 CFR 50.69(e) requirement for periodic updates.

In Section 3.2.6, PRA Maintenance and Updates, of its letter dated February 17, 2023 (Reference 1), the licensee described the process for maintaining and updating the McGuire PRA models used for the 10 CFR 50.69 categorization process and states, in part, that,

The Duke Energy risk management process ensures that the applicable PRA models used in this application continues to reflect the as-built and as-operated plant for each of the MNS [McGuire] units. The process delineates the responsibilities and guidelines for updating the PRA models and includes criteria for both regularly scheduled and interim PRA model updates. The process includes provisions for monitoring potential areas affecting the PRA models (e.g.,

due to changes in the plant, errors or limitations identified in the model, and industry operational experience) for assessing the risk impact of unincorporated changes, and for controlling the model and associated computer files. The process will assess the impact of these changes on the plant PRA model in a timely manner but no longer than once every two refueling outages. If there is a significant impact on the PRA model, the SSC categorization will be re-evaluated.

In addition, Duke Energy will implement a process that addresses the requirements in NEI 00-04, Section 11, Program Documentation and Change Control. The process will review the results of periodic and interim updates of the plant PRA that may affect the results of the categorization process. If the results are affected, adjustments will be made as necessary to the categorization or treatment processes to maintain the validity of the processes. In addition, any PRA model upgrades will be peer reviewed prior to implementing those changes in the PRA model used for categorization.

In Section 3.4, Risk Evaluation (10 CFR 50.69(b)(2)(iv)), of its letter dated February 17, 2023 (Reference 1), the licensee described the overall risk evaluation process:

The MNS [McGuire] 10 CFR 50.69 categorization process will implement the guidance in NEI 00-04. The overall risk evaluation process described in the NEI guidance addresses both known degradation mechanisms and common-cause interactions and meets the requirements of §50.69(b)(2)(iv). Sensitivity studies described in NEI 00-04 Section 8 will be used to confirm that the categorization process results in acceptably small increases to core damage frequency (CDF) and LERF. The failure rates for equipment and initiating event frequencies used in the PRA include the quantifiable impacts from known degradation mechanisms, as well as other mechanisms (e.g., design errors, manufacturing deficiencies, and human errors). Subsequent performance monitoring and PRA updates required by the rule will continue to capture this data and provide timely insights into the need to account for any important new degradation mechanisms.

The NRC staff finds the risk management process described by the licensee in the LAR is consistent with Section 12 of NEI 00-04 guidance. Based on the above, the NRC staff has determined that the proposed change satisfies the fifth key principle for risk-informed decision-making prescribed in RG 1.174 and would meet the requirements in 10 CFR 50.69(e).

Program Documentation and Change Control (NEI 00-04, Section 11)

Section 50.69(f) of 10 CFR requires, in part, program documentation, change control, and records. As noted above, the licensee stated that it will implement a process that addresses the requirements in Section 11 of NEI 00-04, pertaining to program documentation and change control records. In Section 3.1.1 of its letter dated February 17, 2023 (Reference 1) states, in part, that, the SSC categorization process documentation will include the following 10 elements:

1. Program procedures used in the categorization

2. System functions, identified and categorized with the associated bases

3. Mapping of components to support function(s)

4. PRA model results, including sensitivity studies

5. Hazards analyses, as applicable

6. Passive categorization results and bases

7. Categorization results including all asso ciated bases and RISC classifications

8. Component critical attributes for HSS SSCs

9. Results of periodic reviews and SSC performance evaluations

10. IDP meeting minutes and qualification/training records for the IDP members

The NRC staff also recognizes that under 10 CFR Part 50, Appendix B, criterion VI, Document Control, procedures are considered formal plant documents requiring that [m]easures shall be established to control the issuance of documents, such as instructions, procedures, and drawings, including changes thereto, which pr escribe all activities affecting quality. The elements provided in Section 3.1.1 of the enclosure to its letter dated February 17, 2023 (Reference 1), in addition to the list of other categorization prerequisites provided in of that enclosure and supplemented by the enclosure to the letter dated November 2, 2023 (Reference 2), will ensure that the McGuire 10 CFR 50.69 categorization process will be documented in formal licensee procedures. This is consistent with Section 11 of NEI 00-04. Based on the above, the NRC staff finds that the procedures will be sufficient for meeting the 10 CFR 50.69(f) requirement for program documentation, change control, and records.

4.0 CHANGES TO THE OPERATING LICENSE

Based on the NRC staffs review of its letter dated February 17, 2023 (Reference 1), as supplemented by letter dated November 2, 2023 (Reference 2), the NRC staff identified specific actions, as described below, that are identified as being necessary to support the NRC staffs

conclusion that the proposed program meets the requirements in 10 CFR 50.69 as well as the guidance in NEI 00-04 and RG 1.201. Additional actions (e.g., final procedures and proposed alternative treatment) have not been submitted by the licensee or reviewed by the NRC staff for issuance of this safety evaluation (Reference 1 Section 3.1.1), but will be completed before implementation of the program as specified in the 10 CFR 50.69 rule.

The NRC staffs finding on the acceptability of the SSC categorization process and PRA evaluation in the licensees proposed 10 CFR 50.69 program is conditioned upon the license condition provided below. The licensee proposed to add the following license condition to Appendix B, Additional Conditions, of the Renewed Facility Operating Licenses for McGuire Nuclear Station, Units 1 and 2:

Duke Energy 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) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and high winds; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2, Class 3, and non-class SSCs and their associated supports; the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in the ASME/ANS PRA Standard RA -Sa-2009 for other external hazards except seismic, and the alternative seismic approach described in Duke Energys submittal letter RA-18- 0090 dated February 17, 2023; as specified in License Amendment Nos. 331 (Unit 1) and 310 (Unit 2) dated April 8, 2024.

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

The NRC staff finds that the proposed license co ndition is acceptable because it adequately implements 10 CFR 50.69 using models, methods, and approaches consistent with the applicable guidance that has previously been endorsed by the NRC. The NRC also finds the non-PRA methods for assessing risk for internal fires, seismic, and passive components, which are deviations from NEI 00-04, to be acceptable.

The NRC staff notes that the guidance for implementing 10 CFR 50.69 provided by the Commission in the Federal Register notice dated November 22, 2004, 5 Section III.4.10.2, Section 50.36 Technical Specifications, stated that the 10 CFR 50.69 rule does not include 10 CFR 50.36 in the list of special treatment requirements that may be replaced by the alternative 10 CFR 50.69 requirements for RISC-3 and RISC-4 SSCs when implementing a 10 CFR 50.69 license amendment. As a result, the NRC staff does not consider the TS (including Improved TS) and the associated technical requirements manual to be part of the

5 Federal Register Notice (69 FR 68008, 68028-68029; November 22, 2004), related to Risk-Informed Categorization and Treatment of Structure, Systems and Components for Nuclear Power Reactors.

10 CFR 50.69 rule. Therefore, the licensee must address proposed changes to its TS separately.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the North Carolina State official was notified of the proposed issuance of the amendments on February 17, 2024. On February 19, 2024, the State official confirmed that the State of North Carolina had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change the requirement 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 published in the Federal Register Federal Register on May 16, 2023 (88 FR 31285) and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for catego rical 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.

7.0 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.

8.0 REFERENCES

1. Pigott, E.R., Duke Energy letter to the NRC, Application to Adopt 10 CFR 50.69, Risk-informed categorization and treatment of stru ctures, systems, and components for nuclear power reactors, February 17, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23048A022).

2. Pigott, E.R., Duke Energy to NRC, Supplement to Application to Adopt Risk-Informed Completion Times TSTF -505, Revision 2 and Application to Adopt 10 CFR 50.69, Risk-informed categorization and treatment of stru ctures, systems, and components for nuclear power reactors, November 2, 2023 (ML23306A032).

3. RG 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance, Revision 1, May 2006 (ML061090627).

4. RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Acti vities, Revision 2, March 2009 (ML090410014).

5. RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 2, May 2011 (ML100910006).

6. NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, Revision 1, March 2017 (ML17062A466).

7. NUREG- 0800, Chapter 19, Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance, June 2007, (ML071700658).

8. NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0, July 2005 (ML052910035).

9. EPRI Technical Update 3002017583, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk--Informed Categorization, February 2020, (ML21082A170).

10. U.S. Nuclear Regulatory Commission (Exhibit 8), Individual Plant Examination of External Events (IPEEEs) for Severe Accident Vulnerabilities - 10 CFR 50.54(f) (Generic Letter 88-20, Supplement 4), June 28, 1991 (ML003769582).

11. American Society of Mechanical Engineers and American Nuclear Society ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, February 2009, New York, NY (Copyright).

12. NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management, December 1991 (ML14365A203).

13. Markley, M. T., U.S. Nuclear Regulatory Commission, letter to Vice President, Operations, Entergy Operations, Inc., Arkansas Nuclear One, Unit 2 - Approval of Request for Alternative ANO2-R&R-004, Revision 1, Request to Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Safety Systems, April 22, 2009 (ML090930246).

14. NEI 05- 04, Process for Performing Internal Events Peer Reviews Using the ASME/ANS PRA Standard, Revision 3, November 2009.

15. Anderson, V.K., NEI, letter to Stacey Rosenberg, NRC, Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations, February 21, 2017 (ML17086A431).

16. NEI 17- 07, Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard, Revision 2, August 2019 (ML19231A182).

17. NEI 07-12, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, Revision 1, June 2010 (ML102230070).

18. American Society of Mechanical Engineers and American Nuclear Society, ASME/ANS RA-Sb-2013, Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, September 30, 2013, New York, NY (Copyright).

19. Mahoney, Michael, NRC letter to Ray, Thomas D., Duke Energy Carolinas, LLC McGuire Nuclear Station, Units 1 and 2 - Issuance of Amendment Nos. 314 and 293 to Technical Specification 3.8.1, AC Sources - Operating, dated June 28, 2019 (ML19126A030).

20. NRC memorandum, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments, dated May 6, 2022 (ML22014A084).

21. Pigott, Edward, Duke Energy, letter to NRC, "License Amendment Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b," February 16, 2023 (ML23047A465).

22. Murray, D., Exelon Generation Company, letter to NRC, Response to Request for Additional Information regarding LaSalle License Amendment Request to Renewed Facility Operating Licenses to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors (EPID L-2020-LLA-0017), October 16, 2020 (ML20290A791).

23. Murray, D., Exelon Generation Company, letter to NRC, Response to Request for Additional Information Regarding the License Amendment Request to Adopt 10 CFR 50.69 (EPID L-2020-LLA-0017), January 22, 2021 (ML21022A130).

24. Murray, D., Exelon Generation Company, letter to NRC, Application to Adopt 10 CFR 50.69, Risk-informed categorization and treatm ent of structures, systems, and components for nuclear power reactors dated January 31, 2020 (ML20031E699).

25. EPRI 3002012988, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization, July 2018.

26. Marshall, M.L., NRC letter to Exelon Generation Company, LLC, Calvert Cliffs Nuclear Power Plant, Units 1 and 2 - Issuance of Amendment Nos. 332 and 310 Re: Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors (EPID L-2018-LLA-0482)," February 28, 2020. (ML19330D909).

27. Simpson, R., Exelon Generation Company, letter to NRC, Response to Request for Additional Information Regarding License Amendment Requests to Adopt TSTF-505, Revision 2, and 10 CFR 50.69, November 24, 2020. (ML20329A433).

28. Vaidya, B.K., NRC letter to Rhoades, David P., Exelon Generation Company, LLC, LaSalle County Station, Unit Nos. 1 and 2 - Issuance of Amendment Nos. 249 and 235 Related to Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors (EPID L-2020-LLA-0017), May 27, 2021 (ML21082A422).

29. Leeds, E.J., NRC Letter, Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated March 12, 2012 (ML12053A340).

30. Capps, S.D., Duke Energy letter to NRC, Seismic Hazard and Screening Report (CEUS Sites), Response to NRC 10 CFR 50.54(f) Request for Information Pursuant to Title 10 of the Code of Federal Regulation 50.54(f) Regarding Recommendation 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, March 20, 2014 (ML14098A421).

31. Vega, F.G., NRC letter to Vice President, Duke Energy Carolinas, LLC, McGuire Nuclear Power Station, Units 1 and 2 - Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, section 50.54(f), Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-term Task Force Review of Insights from the Fukushima Dai-ichi Accident (TAC Nos. MF3689 and MF3690), July 20, 2015 (ML15182A067).

32. Capps, S.D., Duke Energy Letter to NRC, McGuire Nuclear Station (MNS) Seismic Mitigating Strategies Assessment (MSA) Report for the Reevaluated Seismic Hazard Information - NEI 12-06, Appendix H, Revision 2, H.4.4 Path 4: GMRS < 2xSSE, dated August 10, 2017 (ML17233A167).

33. Gullott, D., Exelon Generation Company letter to NRC, Response to Request for Additional Information regarding LaSalle License Amendment Request to Renewed Facility Operating Licenses to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors (EPID L-2020-LLA-0017), October 1, 2020 (ML20275A292).

34. ASME Boiler & Pressure Vessel Code Case N -660, Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities.

35. Franovich, M. NRC letter to E. Kapopoulos, Duke Energy, Catawba Nuclear Station, Units 1&2 and McGuire Nuclear Station, Units 1 & 2, Screening and Prioritization Results Regarding Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Force Review of Insights from the Fukushima Dai-Ichi Accident, December 22, 2016 (ML16344A313).

Principal Contributors: M. Patterson, NRR B. Gurjendra, NRR K. Tetter, NRR A. Brown, NRR B. Lee, NRR D. Widrevitz, NRR D. Nold, NRR E. Kleeh, NRR J. Ambrosini, NRR S. Cumblidge, NRR

Date: April 8, 2024

ML24052A306 OFFICE NRR/LPL2-1/PM NRR/LPL2-1/LA NRR/DRA/APLA/BC NRR/DRA/APLC/BC NAME JKlos KGoldstein BPascarelli SVasavada DATE 2/17/2024 04/05/2024 1/24/2024 2/7/2024 OFFICE NRR/DEX/EMIB/BC NRR/DEX/EEEB/BC NRR/DNRL/NPHP/BC NRR/DNRL/NVIB/BC NAME SBailey WMorton MMitchell ABuford DATE 2/7/2024 2/9/2024 2/7/2024 2/7/2024 OFFICE NRR/DSS/SCPB/(A)BC NRR/DSS/SNSB/BC OGC - NLO NRR/LPL2-1/BC NAME NKaripineni PSahd MFWoods MMarkley DATE 2/7/2024 2/7/2024 3/15/2024 4/8/2024 OFFICE NRR/LPL2-1/PM NAME JKlos DATE 4/8/2024