ML23291A464

Issuance of Amendment No. 251 Regarding the Adoption of Title 10 the Code of Federal Regulations Section 50.69, Risk-Informed Categorization and Treatment of SSC for Nuclear Power Plants

ML23291A464
Person / Time
Site:	Nine Mile Point
Issue date:	12/07/2023
From:	Richard Guzman NRC/NRR/DORL/LPL1
To:	Rhoades D Constellation Energy Generation, Constellation Nuclear
References
EPID L-2022-LLA-0186
Download: ML23291A464 (1)

Text

December 7, 2023 David P. Rhoades Senior Vice President Constellation Energy Generation, LLC President and Chief Nuclear Officer Constellation Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

NINE MILE POINT NUCLEAR STATION, UNIT 1 - ISSUANCE OF AMENDMENT NO. 251 REGARDING 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 PLANTS (EPID L-2022-LLA-0186)

Dear David Rhoades:

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued the enclosed Amendment No. 251 to Renewed Facility Operating License No. DPR-63 for the Nine Mile Point Nuclear Station, Unit 1 (NMP1). The amendment is in response to your application dated December 15, 2022, as supplemented by letter dated August 4, 2023.

The amendment revises the NMP1 Renewed Facility Operating License No. DPR-63 to add a new license condition to allow for the implementation 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.

A copy of the related safety evaluation is enclosed. Notice of Issuance will be included in the next Commissions monthly Federal Register notice.

Sincerely,

/RA/

Richard V. Guzman, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-220

Enclosures:

1. Amendment No. 251 to NPF-63

2. Safety Evaluation cc: Listserv

NINE MILE POINT NUCLEAR STATION, LLC CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-220 NINE MILE POINT NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 251 Renewed License No. NPF-63

1.

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

A.

The application for amendment by Constellation Energy Generation, LLC dated December 15, 2022, as supplemented by letter dated August 4, 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 set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the 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 Commissions regulations; 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 Commissions regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes as indicated in the attachment to this license amendment, and Renewed Facility Operating License No. NPF-63 is hereby amended to add paragraph 2.D.(27) to read as follows:

(27)

Implementation of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Constellation Energy Generation, LLC 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, and internal fire; high wind safe shutdown equipment list to evaluate high wind / tornado missile events; 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 and Class 3 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the Individual Plant Examination of External Events (IPEEE)

Screening Assessment for External Hazards updated using the external hazard screening significance process identified in American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in Constellation Energy Generation, LLC's submittal letter dated, December 15, 2022, and all its subsequent associated supplements as specified in License Amendment No. 251 dated December 7, 2023.

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 amendment is effective as of its date of issuance and shall be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Hipólito J. González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-63 Date of Issuance: December 7, 2023 Hipolito J.

Gonzalez Digitally signed by Hipolito J.

Gonzalez Date: 2023.12.07 13:09:08 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 251 NINE MILE POINT NUCLEAR STATION, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-63 DOCKET NO. 50-220 Replace the following pages of Renewed Facility Operating License with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Page Insert Page 4

4 13 13 14 Renewed License No. DPR-63 Revised by letter dated February 21, 2007 Amendment No. 195, 209, 214, 219, 220, 247, 250, 251 (1)

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 1850 megawatts (thermal).

(2)

Technical Specifications The Technical Specifications contained in Appendix A, which is attached hereto, as revised through Amendment No. 251, is hereby incorporated into this license. Constellation Energy Generation, LLC shall operate the facility in accordance with the Technical Specifications.

(3)

Deleted D.

This license is subject to the following additional conditions:

(1)

NMP LLC will complete construction of a new radwaste facility in conformance with the design defined and evaluated in the FES, to be operational no later than June 1976.

(2)

Deleted by License Amendment No. 51 (3)

Deleted by License Amendment No. 51 (4)

Security, Training and Qualification and Safeguards Contingency Plans Constellation Energy Generation, LLC shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans, including amendments made pursuant to the provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contain Safeguards Information protected under 10 CFR 73.21 is entitled Nine Mile Point Nuclear Station, LLC Physical Security, Safeguards Contingency, and Security Training and Qualification Plan, Revision 1, and was submitted by letter dated April 26, 2006.

Constellation Energy Generation, LLC shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The licensees CSP was approved by License Amendment No. 209 and modified by License Amendment No. 219. The licensee has obtained Commission authorization to use Section 161A preemption authority under 42 U.S.C 2201a for weapons at its facility.

(5)

Paragraph 2.D(5) of the license has been combined with paragraph 2.D(4) as amended above into a single paragraph.

Renewed License No. DPR-63 Amendment No. 251 (27)

Implementation of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Constellation Energy Generation, LLC 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, and internal fire; high wind safe shutdown equipment list to evaluate high wind / tornado missile events; 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 and Class 3 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the Individual Plant Examination of External Events (IPEEE)

Screening Assessment for External Hazards updated using the external hazard screening significance process identified in American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in Constellation Energy Generation, LLC's submittal letter dated, December 15, 2022, and all its subsequent associated supplements as specified in License Amendment No. 251 dated December 7, 2023.

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).

Renewed License No. DPR-63 E.

This license is effective as of the date of issuance and shall expire on August 22, 2029.

F.

The UFSAR supplement, as revised, submitted pursuant to 10 CFR 54.21(d), shall be included in the next scheduled update to the UFSAR required by 10 CFR 50.71(e)(4) following the issuance of this renewed operating license. Until that update is complete, the licensee may make changes to the programs and activities described in the supplement without prior Commission approval, provided that the licensee evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

G.

The UFSAR supplement, as revised, describes certain future activities to be completed prior to the period of extended operation. NMP LLC shall complete these activities in accordance with the schedule in Appendix A of NUREG-1900, Safety Evaluation Report Related to the License Renewal of Nine Mile Point Nuclear Station, Units 1 and 2, dated September 2006, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

H.

All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of the most recent NRC-approved version of the Boiling Water Reactor Vessels and Internals Project (BWRVIP)

Integrated Surveillance Program (ISP) appropriate for the configuration of the specimens in the capsule. All capsules placed in storage must be maintained for future insertion. Any changes to storage requirements must be approved by the NRC, as required by 10 CFR Part 50, Appendix H.

FOR THE NUCLEAR REGULATORY COMMISSION Original Signed by J. E. Dyer, Director Office of Nuclear Reactor Regulation

Enclosure:

Appendix A - Technical Specifications Date of Issuance: October 31, 2006

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 251 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-63 CONSTELLATION ENERGY GENERATION, LLC NINE MILE POINT NUCLEAR STATION, UNIT 1 DOCKET NO. 50-220

1.0 INTRODUCTION

By letter dated December 15, 2022 (Reference [1]), as supplemented by letter dated August 4, 2023 (Reference [2]), Constellation Generation Company, LLC (CEG, the licensee) submitted a license amendment request (LAR) for the Nine Mile Point Nuclear Station, Unit 1 (NMP1) to the U.S. Nuclear Regulatory Commission (NRC). The licensee proposed the following license condition to the Renewed Facility Operating License (RFOL) 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 reactors.

Constellation Energy Generation, LLC 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, and internal fire; high wind safe shutdown equipment list to evaluate high wind / tornado missile events; 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 and Class 3 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the Individual Plant Examination of External Events (IPEEE) Screening Assessment for External Hazards updated using the external hazard screening significance process identified in American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in Constellation Energy Generation, LLC's submittal letter dated, December 15, 2022, and all its subsequent associated supplements as specified in License Amendment No. 251 dated December 7, 2023.

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 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 significance1.

The supplemental letter dated August 4, 2023 (Reference [2]), 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 February 21, 2023 (88 FR 10558).

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 as 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 SSC 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 their resources on SSCs that make a significant contribution to plant safety. For SSCs that are categorized as HSS, existing treatment requirements are maintained or potentially 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 provides a reasonable level of confidence that these SSCs will satisfy functional requirements. Implementation of 10 CFR 50.69 allows licensees to improve focus on equipment categorized as HSS.

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 (Reference 4),

describes the SSC categorization process in its entirety as an overarching approach that includes multiple approaches and methods identified for a PRA hazard and non-PRA methods.

2.2 Regulatory Guide (RG)

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

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; and RG 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities (Reference [4])

RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (Reference [5])

NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs [Probabilistic Risk Assessments] in Risk-Informed Decision Making (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])

NRC-Endorsed Guidance The Nuclear Energy Institute (NEI) issued NEI 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline (Reference [8]), as endorsed by RG 1.201 for trial use with clarifications and 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 and elements of the SSC categorization process for meeting the requirements of 10 CFR 50.69:

Section 2, Overview of Categorization Process, provides specific guidance corresponding to 10 CFR 50.69(c)(1)(v).

Sections 3.2, Use of Risk Information, and 5.1, Internal Events Assessment, provide specific guidance corresponding to 10 CFR 50.69(c)(1)(i).

Sections 3, Assembly of Plant-Specific Inputs; 4, System Engineering Assessment; 5, Component Safety Significance Assessment; and 7, Preliminary Engineering Categorization of Functions, provide specific guidance corresponding to 10 CFR 50.69(c)(1)(ii).

Section 6, Defense-In-Depth Assessment, provides specific guidance corresponding to 10 CFR 50.69(c)(1)(iii).

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

Sections 9, IDP [Integrated Decision Making Panel] Review and Approval; and 10, SSC Categorization, provide specific guidance corresponding to 10 CFR 50.69(c)(2).

Additionally, Section 11, Program Documentation and Change Control, of NEI 00-04 provides guidance on program documentation and change control related to the requirements of 10 CFR 50.69(f). Section 12, Periodic Review, 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).

2.3 Licensees Proposed Changes The licensee proposed the following License Condition 2.(D).(27) to the NMP1 RFOL to allow the implementation of 10 CFR 50.69 (27)

Implementation of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Constellation Energy Generation, LLC 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, and internal fire; high wind safe shutdown equipment list to evaluate high wind / tornado missile events; 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 and Class 3 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the Individual Plant Examination of External Events (IPEEE)

Screening Assessment for External Hazards updated using the external hazard screening significance process identified in American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in Constellation Energy Generation, LLC's submittal letter dated, December 15, 2022, and all its subsequent associated supplements as specified in License Amendment No. 251 dated December 7, 2023.

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).

3.0 TECHNICAL EVALUATION

3.1 Method of NRC Staff Review An acceptable approach for making risk-informed decisions about proposed technical specification changes, including both permanent 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 3 (Reference [5]). These key principles are:

Principle 1:

The proposed licensing basis change meets the current regulations unless it is explicitly related to a requested exemption.

Principle 2:

The proposed licensing basis change is consistent with the defense-in-depth philosophy.

Principle 3:

The proposed licensing basis change maintains sufficient safety margins.

Principle 4:

When the proposed licensing basis change results in an increase in risk, the increases 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 licensing basis 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 3 and are pertinent to: (1) compliance with current regulations, (2) the evaluation of defense-in-depth, and (3) the evaluation of safety margins.

3.2.1 Key Principle 1: Licensing Bases 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 nonsafety-related SSCs according to the safety significance of the functions they perform into one of the following four RISC categories, which are defined in 10 CFR 50.69(a), as follows:

RISC-1:

Safety-related SSCs that perform safety significant functions2 RISC-2:

Nonsafety-related SSCs that perform safety significant functions RISC-3:

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

Nonsafety-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 HSS SSCs, 10 CFR 50.69 maintains current regulatory requirements for special treatment (i.e., it does not remove any requirements 2 NEI 00-04, Revision 0 (Reference 8), 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 used in 10 CFR 50.69.

from these SSCs). For LSS SSCs, licensees can implement alternative treatment requirements in accordance with 10 CFR 50.69(b)(1) and 10 CFR 50.69(d). For RISC-3 SSCs, licensees can replace special treatment with an alternative treatment. For RISC-4 SSCs, 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 categorization process against the categorization process described in NEI 00-04, Revision 0, as endorsed in RG 1.201, Revision 1, and the acceptability of the licensees PRA for use in the application of the 10 CFR 50.69 categorization process. The NRC staffs review, as documented in this safety evaluation (SE), used the framework provided in RG 1.174, and NEI 00-04, Revision 0, as endorsed in RG 1.201, Revision 1.

Section 2 of NEI 00-04, Revision 0, states, in part, that the categorization process includes eight primary steps:

1. Assembly of Plant-Specific Inputs (Section 3 of NEI 00-04, Revision 0)

2. System Engineering Assessment (Section 4 of NEI 00-04, Revision 0)

3. Component Safety Significance Assessment (Section 5 of NEI 00-04, Revision 0)

4. Defense-In-Depth Assessment (Section 6 of NEI 00-04, Revision 0)

5. Preliminary Engineering Categorization of Functions (Section 7 of NEI 00-04, Revision 0)

6. Risk Sensitivity Study (Section 8 of NEI 00-04, Revision 0)

7. IDP [Integrated Decision-Making Panel] Review and Approval (Section 9 of NEI 00-04, Revision 0)

8. SSC Categorization (Section 10 of NEI 00-04, Revision 0)

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, Revision 0, as endorsed in RG 1.201, Revision 1. In Sections 3.2.3 and 3.2.4 of the enclosure to the LAR, the licensee has proposed the use of the Electric Power Research

Institute (EPRI) 3002017583 Tier 1 alternate seismic approach and the high winds safe shutdown equipment list (HWSSEL), as alternative methods to assess the applicable hazard contributions. The NRC notes that use of these alternative methods is a deviation from the NEI 00-04 guidance as endorsed. A more detailed staff review of the alternative methods is provided in section 3.3 of this SE.

The licensee provided further discussion of specific elements within the 10 CFR 50.69 categorization process that are delineated in the NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 0.

The regulatory requirements in 10 CFR 50.69 and 10 CFR Part 50, Appendix B, and the monitoring outlined in NEI 00-04, Revision 0 and clarifications in RG 1.201, Revision 1, ensure that the SSC categorization process is sufficient to assure 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/elements prescribed in NEI 00-04, Revision 0 to assure that SSCs specified are appropriately categorized consistent with 10 CFR 50.69. The staff performed a more detailed review of specific steps/elements of the licensees SSC categorization process where necessary to confirm consistency with the NEI 00-04 guidance, as endorsed. In light of 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 3.

3.2.2 Key Principle 2: Licensing Basis Change is Consistent With the Defense-In-Depth Philosophy In RG 1.174, Revision 3, the NRC identified the following considerations used for evaluating how the licensing basis change is maintained for the defense-in-depth philosophy:

Preserve a reasonable balance among the layers of defense; Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures; Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty; Preserve adequate defense against potential common-cause failures Maintain multiple fission product barriers Preserve sufficient defense against human errors Continue to meet the intent of the plants design criteria.

RG 1.201, Revision 1, endorses the guidance in Section 6 of NEI 00-04, Revision 0, 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, Prescriptive Requirements, and B, Performance-Based Requirements, of Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, to 10 CFR Part 50. The criteria provided in paragraph 50.69(b)(1)(x) of 10 CFR are not to determine the proper RISC category for containment isolation valves or penetrations.

In Section 3.1.1 of the enclosure to the LAR, the licensee clarified that it would require an SSC to be categorized as preliminary HSS based on the defense-in-depth assessment performed in accordance with NEI 00-04, Revision 0. In light of 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 3. The NRC staff finds that the licensee's process is consistent with the NRC-endorsed guidance in NEI 00-04; therefore, key principle 2 of risk-informed decision-making is met and fulfills the 10 CFR 50.69(c)(1)(iii) criterion that requires defense-in-depth to be maintained.

3.2.3 Key Principle 3: Licensing Basis Change Maintains Sufficient Safety Margins The regulations in 10CFR50.69(c)(1)(iv) requires the evaluations to provide reasonable confidence that for SSCs categorized as RISC-3, sufficient safety margins are maintained and that any potential increases in core damage frequency (CDF) and large early release frequency (LERF) resulting from changes in treatment are small. 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 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 sufficient safety margin will continue to exist.

Consistent with the guidance provided in NEI 00-04 for review of safety margins, and in accordance with the implementation of the SSC categorization program, the only requirements that are relaxed for LSS SSCs (includes RISC-3) are those related to treatment. The SSCs design basis function as described in the plants licensing basis, including the updated final safety analysis report and technical specifications 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. On this basis, the staff concludes that the licensees SSC categorization program ensures that sufficient safety margins are maintained in accordance with the third key safety principle of RG 1.174, Revision 3, and therefore, 10 CFR 50.69(c)(1)(iv) is satisfied.

3.3 Risk-Informed Assessment 3.3.1 Key Principle 4: Change in Risk is Consistent with the Safety Goals The risk-informed considerations prescribed in NEI 00-04, Revision 0, endorsed by RG 1.201, Revision 1, addresses the fourth and fifth key principles of the NRC staffs standards for risk-informed decision-making, pertaining to the assessment for change in risk and monitoring the impact of the licensing basis change.

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

Assembly of Plant-Specific Inputs (NEI 00-04, Revision 0, Section 3)

The NRC staff acknowledges that elements of the categorization process (e.g., system selection, system boundary definition, identification of system functions, and mapping of components to functions) are not always performed in chronological order and may be performed in parallel. This is further discussed in Section 3.2 of this SE. The licensees risk

categorization process uses PRAs to assess risks from the internal events (IEPRA) (including internal flooding), and fire PRA (FPRA). For non-PRA methods that depart from the methodology prescribed in NEI 00-04, additional staff review is discussed in this section of the SE.

Paragraph 50.69(c)(1)(v) of 10 CFR requires that SSC categorization be performed for entire systems and structures, not for selected components within a system or structure. Based on its review, the NRC staff finds the process described in the LAR, as supplemented by letter dated August 4, 2023, for collecting and organizing information at the system level for defining boundaries, functions, and components is consistent with NEI 00-04 as endorsed by the staff in RG 1.201 and therefore meets the requirements set forth in 10 CFR 50.69(c)(1)(v).

System Engineering Assessment (NEI 00-04, Revision 0, Section 4)

In Section 2.2 of the enclosure to the LAR, the licensee stated [t]he safety functions [in the categorization process] include the design basis functions, as well as functions credited for severe accidents (including external events). Section 3.1.1 of the enclosure to the LAR summarizes the different hazards and plant states for which functional and risk significant information will be collected. In Section 3.1.1 of the enclosure to the LAR, the licensee confirmed that the SSC categorization process documentation will include, among other items, system functions, identified and categorized with the associated bases, and mapping of components to support function(s).

Based on its review, the NRC staff finds that the process described in the LAR is consistent with NEI 00-04, Revision 0, as endorsed by the NRC in RG 1.201, Revision 1 and meets the requirements set forth in paragraph 50.69(c)(1)(ii) and 50.69(c)(1)(iv).

Component Safety Significance Assessment (NEI 00-04, Section 5)

This step in the licensees categorization process assesses the safety significance of components using quantitative or qualitative risk information from a modeled PRA hazard, other hazards that can be screened, and non-PRA method(s). In the NEI 00-04 guidance, component risk significance is assessed separately for the following hazard groups:

internal events (including internal floods) internal fire events seismic events external hazards (e.g., high winds, external floods) shutdown events In Sections 3.2.1 and 3.2.2 of the enclosure to the LAR, the licensee described that the NMP1 categorization process uses PRA modeled hazards to assess risks for the [internal events]

(includes internal flood) and internal fires. For the other risk contributors, the licensee's process uses the following non-PRA methods to characterize the risk:

Seismic Hazard: Alternative seismic treatment using guidance from EPRI Report 3002017583 dated February 29, 2020 (Reference [9]), and qualitative insights about seismic risk at NMP1.

High Wind and Tornado Missile Hazard: Development and use of a high wind safe shutdown equipment list (HWSSEL) approach.

Other External Hazards: Screening analysis performed for IPEEE in according with GL 88-20 (Reference [10]) and updated using criteria from Part 6 of the SME/ANS RA-Sa-2009, Addendum A to RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, (the PRA Standard, Reference [15]), as endorsed by the NRC.

Shutdown Events: Safe Shutdown Risk Management program consistent with NUMARC 91-06 (Reference [8] and [17]).

Passive Components: ANO-2 passive categorization methodology (Reference [12]).

The approaches and methods proposed by the licensee to address internal events, seismic, high winds and other external events, defense-in-depth, and shutdown events are consistent with the approaches and methods included in the guidance in NEI 00-04, Revision 0. The non-PRA method for the categorization for passive components is consistent with the ANO-2 methodology for passive components approved for risk-informed safety classification and treatment for repair/replacement activities in class 2 and 3 moderate-and high-energy systems.

The use of the ANO-2 methodology in the SSC categorization process is provided in section 3.3.2 of this SE. To address seismic hazard in the SSC categorization process, the licensee proposed to use an alternative method not endorsed by the NRC in NEI 00-04. A detailed NRC staff review of the licensees proposed alternative seismic approach is provided in section 3.3.2 of this SE. To address the high wind hazard in the SSC categorization process, the licensee proposed to develop and use a HWSSEL. A detailed NRC staff review of the licensees proposed alternative high wind approach is provided in section 3.3.2 of this SE.

3.3.1.1 Scope of the PRA The NMP1 PRA is comprised of a full-power, Level 1, IEPRA, IFPRA, and FPRA, which evaluate the core damage frequency (CDF) and large early release frequency (LERF) risk metrics. The NMP1 NFPA 805 and TSTF-425 NRC Safety Evaluations (References [11] and

[12]) state that the February 2008 peer review for the IEPRA (includes internal floods) model was assessed against the PRA Standard ASME/ANS RA-Sc-2007 and RG 1.200, Revision 1, which only endorsed the ASME/ANS RA-Sb-2005 standard. Given that RG 1.174, Revision 3, provides that the 2009 PRA Standard (Reference [13]) endorsed by Revision 2 of RG 1.200 be used, the licensee provided gap assessments between the three standards (2005, 2007, and 2009) for NRC staff approval during the NFPA 805 and TSTF-425 application reviews. The NRC staff reviewed those assessments and determined that there was no significant impact for this application. The licensees peer review of the FPRA model was assessed against RG 1.200, Revision 2.

Furthermore, Section 3.3 of the enclosure to the LAR states that finding closure reviews and focused-scope peer reviews were conducted on both the IEPRA and FPRA models in November 2017 and December 2021. Open findings were reviewed and closed using the NRC-accepted process documented in the NEI letter to the NRC Final Revision of Appendix X to NEI 05-04/07-12/12-16, 'Close-out of Facts and Observations, dated February 21, 2017 (Reference [14]).

The NRC staff finds that the LAR provides sufficient information to support the staff review of the IEPRA (includes internal flooding) and FPRA for technical acceptability, and therefore, meets the requirements set forth in paragraph 50.69(b)(2)(iii) of 10 CFR.

Aspects considered by the staff to evaluate the scope of the PRA include: (1) peer review history and results, (2) the Appendix X, Independent Assessment process, (3) credit for FLEX in

the PRA, and (4) assessment of assumptions and approximations. The staff's review of these aspects of the PRA to assess for consistency with the applicable processes as endorsed by the NRC, where necessary, are provided below.

Internal Events PRA (includes internal floods) Peer Review History In Section 3.3 of the enclosure to the LAR, the licensee states that the internal events PRA model was subjected to a full-scope peer review in February 2008. Subsequently, in November 2017 and December 2021 the licensee conducted Independent Assessments for closure of the finding-level facts and observations (F&Os) and performed a focused-scope peer review in December 2021. These reviews concluded all the IEPRA (includes internal floods)

F&Os have been closed.

In Section 3.2 of the enclosure to the LAR, for the IEPRA (includes internal floods), states in part, there are no PRA upgrades that have not been peer reviewed. In review of the licensees reports, the NRC staff concluded that all F&Os were appropriately assessed by the Independent Assessment team to assure that no new methods or upgrades were inadvertently incorporated into the IEPRA without a peer review in accordance with the ASME/ANS RA-Sa-2009 PRA standard as endorsed by the NRC.

Therefore, the NRC staff concludes that the NMP1 IEPRA (including internal floods) was appropriately peer reviewed, consistent with RG 1.200, Revision 2, and the F&Os have been adequately closed.

Internal Fire PRA Peer Review History The licensee's FPRA was subject to a full-scope industry peer review in January 2012, consistent with RG 1.200, Revision 2. In February 2017 and January 2021, focused-scope peer reviews were performed. The finding-level F&Os from the 2012 full-scope were considered fully resolved by the Independent Assessment review team based on closure reviews conducted in February 2017 and January 2021 which closed out all of the findings. As a result of the 2021 focused-scope peer review of the FPRA, four F&Os were generated. Therefore, in accordance with RG 1.200, Revision 2, these F&Os associated with the FPRA were provided in of the LAR.

For each F&O, the licensee provided a disposition for this application in the LAR. The NRC staff reviewed the F&O dispositions and determined that they have no significant impact for this application.

The NRC staff has reviewed the FPRA peer review results and the licensee's resolution of the results and concludes that the NMP1 FPRA was appropriately peer reviewed, consistent with RG 1.200, Revision 2, and the F&Os have been adequately dispositions to assess the impact on the risk-informed application.

Appendix X, Independent Assessment Process for F&O Closure Section X.1.3 of Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13 provides guidance to perform an Independent Assessment for the closure of F&O identified from a full-scope or focused-scope peer review.

In review of the LAR the NRC staff concluded that all F&Os were appropriately assessed by the Independent Assessment team to assure that no new methods or upgrades were inadvertently incorporated into the IEPRA without a peer review in accordance with the ASME/ANS RA-Sa-2009 PRA standard as endorsed by the NRC. Therefore, the NRC staff finds that the NMP1 IEPRA (includes internal floods) and FPRA were appropriately peer reviewed consistent with RG 1.200, Revision 2 and meets the requirements set forth in 10 CFR 50.69(c)(1)(i).

Credit for FLEX Equipment The NRC memorandum dated May 6, 2022, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Risk Assessments (Reference [15]), provides the NRC staffs assessment of challenges to incorporating FLEX equipment and strategies into a PRA model in support of risk-informed decision-making in accordance with the guidance of RG 1.200, Revision 2.

The licensee provided in the August 4, 2023, supplement results of a sensitivity study that demonstrated that the NMP1 PRA implementation of FLEX equipment or FLEX strategies in the PRA does not impact the categorization results. Therefore, the NRC staff finds that the NMP1 credit of FLEX equipment does not impact the SSC categorization process.

Assessment of Assumptions and Approximations Identification of Key Assumptions and Sources of Uncertainty NMP1 confirmed that the detailed process of identifying, screening, and characterizing those sources of model uncertainty and related assumptions in the base PRA that are relevant to this application was performed consistent with NUREG-1855, Revision 1 (Reference [6]).

Substep E-1.4 of the guidance is a qualitative screening process that involves identifying and validating whether consensus3 models have been used in the PRA to evaluate identified model uncertainties. The licensee confirmed that for the NMP1 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 consistent with the guidance provided in NUREG-1855, Revision 1.

Treatment of the Key Assumptions and Sources of Uncertainty NUREG-1855, Revision 1, 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. The licensee confirmed that sensitivity studies will be performed consistent with Table 5-2 of the NEI 00-04 guidance. In accordance with NEI 00-04, the results of the sensitivity study 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 the licensee will perform a sensitivity study consistent with Table 5-2 of the NEI 00-04 guidance to address the identified key assumptions and sources of uncertainty in the context of the decision-making under consideration for the categorization of the SSC at the time of the risk analysis being performed.

3 Per NUREG-1855, Revision 1, 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 addition, the NRC staff recognizes that the licensee will perform routine PRA changes and updates to assure 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) of 10 CFR provide 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 NMP1 includes the IEPRA (includes internal floods) and FPRA. The NRC staff finds that the licensees use and treatment of importance measures is consistent with the guidance in NEI 00-04, Revision 0, as endorsed in RG 1.201, Revision 1. A more detailed staff review of the alternate methods for assessing the risk for seismic and other external hazards is provided in section 3.3.2 of this SE.

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 the IEPRA and FPRA PRAs 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 acceptability of the PRA by comparing the PRA to the relevant parts of the ASME/ANS 2009 Standard using a peer review process.

The licensee has subjected the IEPRA and FPRA PRAs to the peer review processes and submitted the results of the peer review. The NRC staff reviewed the peer review history (which included 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 IEPRA and FPRA PRAs 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, as applicable, and addressed appropriately for this application.

The NRC staff finds the licensee provided the required information, and the IEPRA (includes internal floods) and FPRA, are acceptable, and therefore, meet the requirements set forth in paragraphs 50.69(c)(1)(i) and (ii) of 10 CFR 50.69.

3.3.1.2 Evaluation of the Use of Non-PRA Methods in SSC Categorization The licensees categorization process uses the following non-PRA method(s), respectively:

EPRI Technical Update 3002017583, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," Tier 1 (Reference [9]).

HWSSEL performed for the NMP1 categorization program.

Screening analysis performed for the IPEEE for other hazards.

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

[16]).

Passive Components: ANO-2 passive categorization (Reference [17]).

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

Alternative Seismic Approach As part of its proposed process to categorize SSCs according to safety significance, the licensee proposed to use a non-PRA method to consider seismic hazards. The regulations in 10 CFR 50.69(c)(1)(ii) and 50.69(b)(2)(ii) permit the use of systematic evaluation techniques in the risk-informed categorization process. The licensee provided a description of its proposed alternative seismic approach for considering seismic risk in the categorization process and described how the proposed alternative seismic approach would be used in the categorization process in Section 3.2.3 of the enclosure to the LAR and its supplement dated August 4, 2023.

In part, the licensee based its plant-specific evaluation on the case studies performed in EPRI 3002017583 (Reference [9]), and stated that the case studies are applicable to NMP1 and are used in the alternative seismic approach; how the licensees proposed alternative seismic approach would be used in the categorization process; and the measures for assuring the quality and level of detail for the licensees proposed alternative seismic approach are adequate for the categorization of SSCs. Therefore, based on the above, the NRC staff finds that the requirements in 10 CFR 50.69(b)(2)(ii) for the proposed alternative seismic approach are met.

EPRI 3002017583 includes the results from case studies performed to determine the extent and type of unique HSS SSCs from seismic PRAs (SPRAs). The licensee states in Footnote 1, Item

1. 2 of Section 3.2.3 of the NMP1 LAR that the application incorporates a prior and similar Exelon response to RAI [Request for Additional Information] APLC-03 for the Clinton Power Station 10 CFR 50.69 LAR (Reference [18]) using the same alternative seismic approach. The NRC staffs review confirmed that the case studies in EPRI 3002017583 used by the licensee to support its proposed alternative seismic approach, as well as the information in its supplement, provided sufficient plant-specific evaluation of the applicability of the case studies to NMP1. The information presented in the LAR and its supplement provided a sufficient description of, and basis for acceptability of, the evaluations to be conducted to satisfy 10 CFR 50.69(c)(1)(iv) for the alternative seismic approach. Therefore, the NRC staff finds that the requirements in 10 CFR 50.69(b)(2)(iv) are met for the proposed alternative seismic approach.

Evaluation of the EPRI 3002017583 Case Studies In Section 3.2.3 of the enclosure to the LAR, the licensee stated that the plant-specific case studies from other licensees in the EPRI 3002017583 are incorporated by reference to support its proposed alternative seismic approach. The licensee stated that the case study Plants A, C, and D, pertaining to the technical acceptability of the PRAs used, as well as the technical adequacy of certain technical details of the conduct of the case studies are applicable to NMP1.

The NRC staff reviewed and evaluated the technical acceptability of the PRAs used in case studies for Plants A, C, and D, in EPRI 3002017583, and the licensees assertion of plant-specific applicability to the approach used in the amendment approved by the NRC for Calvert Cliffs on February 28, 2020 (Reference [19]).

The NRC staff finds that the acceptability of PRAs used in the Plants A, C, and D, case studies in EPRI 3002017583, the mapping approach used in those case studies, and the conclusions on the determination of unique HSS SSCs from the case studies in the Calvert Cliffs amendment are applicable to this licensees proposed plant-specific alternative seismic approach.

Evaluation of the Criteria for the Proposed Alternative Seismic Approach In the LAR the licensee states, in part, that the ground motion response spectrum (GMRS) peak acceleration for NMP1 is below the safe shutdown earthquake (SSE) between 1.0 Hz and 10 Hz, which demonstrates that NMP1 qualifies as a Tier 1 plant under the criteria in EPRI 3002017583.

The NRC staff notes that the licensees plant-specific evaluation is supported by its NRC 10 CFR 50.54(f) response dated March 31, 2014 (Reference [20]). The NRC staff reviewed the licensees submittal and supplements and plant-specific evaluation and concludes that the proposed criteria in EPRI 3002017583 to determine the applicability and use of the proposed seismic Tier 1 approach is acceptable.

Evaluation of Applicability of Criteria for 10 CFR 50.69 In Section 3.2.3 of the enclosure to the LAR, the licensee compared the NMP1 GMRS from the reevaluated seismic hazard developed and submitted by the licensee in response to Near-Term Task Force (NTTF) Recommendation 2.1 against the sites design basis SSE, as shown in Figure 1 of Attachment 4 of the LAR, letter, to demonstrate that the site meets the criteria for application of the proposed alternative seismic approach as a Tier 1 plant. The NRC staffs review confirmed the licensees statements and the comparison of the GMRS from the reevaluated seismic hazard against the SSE. Based on its review, the NRC staff finds that the licensees seismic hazard meets the criteria for the proposed alternative seismic approach.

In Section 3.2.3 of the enclosure to the LAR, the licensee stated that the small percentage contribution of seismic to total plant risk makes it unlikely that an integral importance assessment for a component, as defined in NEI 00-04, would result in an overall HSS determination. The NRC staff cited Section 2.2.2 of EPRI report which identifies the expectation that low contribution of seismic risk to the total plant risk reduces the likelihood of a unique seismic condition that would cause an SSC to be designated HSS.

The NRC staffs evaluation of seismic risk to total plant risk was based on information in the NMP1 TSTF-505 LAR (Reference [21]). The NRC staff noted that seismic CDF contribution to total plant CDF is low (i.e., less than 5 percent). The NRC staff reviewed SLERF estimate in the NMP1 TSTF-505 submittal and noted that SLERF is about 20 percent of total plant LERF.

Therefore, the NRC staff finds that overall seismic risk is relatively low compared to total plant risk due to its low seismic CDF and seismic LERF.

Further, as noted in Section 3.6.5 of EPRI 3002017583, containment defense-in-depth assessment addresses containment failures and containment bypass situations. Section 3.6.6 of EPRI 3002017583, used for the licensees proposed alternative seismic approach, recommends that if the licensee chooses to categorize civil structures housing HSS SSCs, the structures are considered as HSS. Therefore, based on its evaluation and review, the NRC staff concludes that the proposed alternative seismic approach, in conjunction with the other elements of the 10 CFR 50.69 categorization program, will appropriately determine the safety significance of any SSCs whose seismic-induced failures led directly to core damage and large early release and that the seismic risk contribution would not solely result in any additional SSC being categorized as HSS.

The NRC staff finds that the licensees basis for applying the proposed alternative seismic approach to its site is acceptable because: (1) the reevaluated hazard meets the criteria for use

of the proposed alternative seismic approach, (2) in conjunction with the other elements of the 10 CFR 50.69 categorization program, the approach will appropriately determine the safety significance of any SSCs whose seismic-induced failures would lead directly to core damage and large early release, and (3) the seismic risk contribution would not solely result in any additional SSC being categorized as HSS.

Evaluation of the Implementation of Conclusions from the Case Studies The licensee stated that the proposed categorization approach for seismic hazards will include qualitative consideration of the mitigation capabilities of SSCs during seismically-induced events and seismic failure modes, based on insights obtained from prior seismic evaluations performed for NMP1. The licensee explained that the qualitative characterization of seismic risk performed for the independent decision-making panel will include information from the various post-Fukushima seismic reviews including results of seismic walkdowns, seismic mitigation strategy assessment, and seismic high frequency evaluations. The objective of the alternative seismic approach is to identify plant-specific seismic insights derived from the components in the system being categorized.

The NRC staffs review of the licensees proposed alternative seismic approach determined that the approach used in the Calvert Cliffs amendment is applicable to this licensees proposed alternative seismic approach and that the plant-specific evaluation on the implementation of the alternative seismic approach is acceptable. The NRC staffs review of the proposed alternative seismic approach, in conjunction with the requirements in 10 CFR 50.69 and the corresponding statement of consideration, finds that the proposed alternative seismic approach includes the evaluations required by 10 CFR 50.69(c)(1)(ii), as well as 10 CFR 50.69(c)(1)(iv) because:

1. The proposed alternative seismic 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.

2. The proposed alternative seismic approach presents system-specific seismic insights to the IDP for consideration as part of the IDP review process as each system is categorized, thereby providing the IDP a means to consider potential impacts of seismic events in the categorization process.

3. The 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 SPRAs in EPRI 3002017583. The insights will use prior plant-specific seismic evaluations and, therefore, in conjunction with performance monitoring for the proposed alternative seismic approach, reasonably reflect the current plant configuration. Further, the recommendation for categorizing civil structures in the alternative seismic approach provides appropriate consideration of such failures from a seismic event.

4. The proposed alternative seismic approach presents the IDP with the basis for the proposed alternative seismic approach, including the low seismic hazard for the plant and the criteria for use of the proposed alternative seismic approach.

5. The proposed alternative seismic approach includes qualitative consideration and insights related to the impact of a seismic event on SSCs for each SSC that is categorized and does not limit the scope to SSCs from the case studies supporting this application.

Consideration of Changes to Seismic Hazard The possibility exists for the seismic hazard at the site to increase such that the criteria for use of the proposed alternative seismic approach are challenged. The licensee stated that the continued comparison of GMRS to SSE applies to the NMP1 site. The licensee also stated that the seismic hazard at the plant is subject to periodic reconsideration as new information became available through industry evaluations.

The NRC staff finds that the consideration of changes to the seismic hazard in the licensees plant-specific proposed alternative seismic approach is the same as that approved in the Calvert Cliffs amendment. Consequently, the NRC staff finds that the consideration of changes to the seismic hazard at NMP1 that exceed the criteria for use of the proposed alternative seismic approach is acceptable for the proposed approach because: (1) the criteria for use of the proposed alternative seismic approach is clear and traceable, (2) the proposed alternative seismic approach includes periodic reconsideration of the seismic hazard as new information becomes available, (3) the proposed alternative seismic approach satisfies the requirements in 10 CFR 50.69 discussed above, and (4) the licensee has included a proposed license condition in the LAR to require NRC approval for a change to the specified seismic categorization approach.

Monitoring of Inputs to and Outcome of Proposed Alternative Seismic Approach In Section 3.5 of the enclosure to the LAR, 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 ensure that the qualitative determinations for the seismic hazard continue to remain in compliance with the requirements of 10 CFR 50.69.

Based on its review, the NRC staff found that consideration of the feedback and adjustment process in the licensees proposed alternative seismic approach is acceptable. The NRC staff finds that:

1. The licensees programs provide reasonable assurance that the existing seismic capacity of LSS components would not be significantly impacted, and

2. The monitoring and configuration control program ensures that potential degradation of the seismic capacity would be detected and addressed before significantly impacting the plant risk profile.

Therefore, the NRC staff finds that the potential impact of the seismic hazard on the categorization is maintained acceptably low and the requirements in 10 CFR 50.69(c)(1)(iv) are met for the proposed alternative seismic approach.

High Winds and Tornado Missiles Approach The licensee determined that the wind pressure and missile hazard from high winds and tornadoes do not screen out for the NMP1 categorization process. In the LAR the licensee

stated that the process used for wind pressure/missile hazard safety significance is a HWSSEL of SSCs. The licensee stated that the HWSSEL will be developed from a list of SSCs needed to achieve and maintain safe shutdown assuming unavailability of offsite power. The licensee explained further that the SSCs that fulfill the wind pressure/missile hazard safe shutdown functions, as well as any high winds or missile barriers that are credited with protecting equipment that fulfills a HWSSEL function, will be identified as candidate HSS for the system being categorized regardless of their tornado damage susceptibility or frequency of challenge.

In an audit question, the staff requested the licensee to specify when the HWSSEL would be developed and to ensure that the HWSSEL would be developed prior to the start of the 10 CFR 50.69 program at NMP1. In the licensees supplement, dated August 4, 2023, the licensee stated that the HWSSEL will be developed prior to categorizing any system at NMP1.

The licensee also provided a revised Attachment 1 in the August 4, 2023, supplement, which added the HWSSEL to the list of prerequisites.

In its August 4, 2023, supplement, the licensee clarified that the high winds initiator should have been listed separately in the LAR Table 3.1, Categorization Evaluation Summary, which lists the different methods for categorization and the rules under which the Integrated Decision-Making Panel can or cannot change an SSCs HSS categorization to LSS. The licensee provided a revised Table 3-1 in the supplement, which specifically includes Extreme Wind or Tornado and stipulates that at the Function/Component level the IDP is not allowed to change HSS to LSS.

The NRC staff, in an audit question, requested justification that the HWSSEL method meets the expectations in the Statements of Consideration for 10 CFR 50.69 that the non-PRA methods used in the categorization are conservative and to identify the industry assessments referenced in the LAR that support this conclusion. In the August 4, 2023, supplement the license stated that the proposed approach is conservative since all SSCs and their associated functions that are needed to achieve and maintain safe shutdown of the reactor given high wind events that are assumed to result in unavailability of offsite power are assigned to HSS. The licensee also compared its HWSSEL with the NRC approved ANO High Wind Equipment List (Reference [19]).

The licensee, in its August 4, 2023, supplement provided details regarding the methodology to be used to develop the HWSSEL which includes the criteria used to screen out SSCs. These criteria include screening SSCs not otherwise credited for mitigating the effects of a high winds event including SSCs not powered by emergency on-site alternate current (AC) sources, SSCs not required to function during or after a loss of offsite AC power event, SSCs in systems that are assumed unavailable following a high wind event, and a few others. The remaining SSCs not screened out will be identified as the HSS SSCs on the HWSSEL.

In an audit question the NRC staff requested an explanation of how the failure probability of operator actions is incorporated in the analysis for determining SSCs in the HWSSEL. In its August 4, 2023, supplement, the licensee stated that there is no reliance on operator actions in the determination of whether an SSC should be assigned to the HWSSEL. In another audit question the NRC staff requested clarification of SSCs that would fall into the Candidate LSS terminator oval in Figure 3-2 in the LAR. In its response, the licensee provided that SSCs not powered by emergency on-site AC sources and SSCs not required to function after a loss of offsite power event would be included as Candidate LSS in this terminator oval.

Therefore, the NRC staff finds that the HWSSEL approach to categorizing SSCs for high winds and tornado missiles is conservative because (1) the approach ensures that SSCs that are credited to achieve and maintain the capability for safe shutdown are retained as safety-significant, (2) SSCs identified as HSS by non-PRA methods for external events may not be re-categorized by the IDP (as stated in NEI 00-04 and the LAR), (3) the high winds approach does not rely on operator actions in determination of SSC safety categorization, and (4) the SSCs that fulfill the wind pressure/missile hazard safe shutdown functions, as well as any high winds or missile barriers that are credited with protecting equipment that fulfills a HWSSEL function, will be identified as candidate HSS for the system being categorized regardless of their tornado damage susceptibility or frequency of challenge.

Methods for Assessing Other External Hazards This hazard category includes all non-seismic and non-high winds external hazards such as external floods, aircraft impact, and other external hazards. The licensee discussed its consideration of other external hazards and concluded that all external hazards, except for seismic and high winds, were screened from applicability to NMP1 per a plant-specific evaluation in accordance with GL 88-20 and the criteria in ASME/ANS RA-Sa-2009 PRA Standard.

In the August 4, 2023, LAR supplement, the licensee confirmed that, during categorization of SSCs, NMP1 will follow the process consistent with the NEI 00-04 guidance in the Figure 5-6 flow chart to determine whether an SSC is credited in the screening of an external hazard (excluding high winds and seismic hazards) and should be categorized accordingly. The NRC staff finds that the licensees SSC categorization process will evaluate the safety significance of SSCs for all other external hazards consistent with the guidance provided in NEI 00-04, as endorsed by the NRC, and is therefore acceptable.

Regarding the external flooding and snowfall hazards, the licensee provided additional information in its August 4, 2023, supplement, in relation to the screening of these hazards from this application. Regarding external flooding, the licensee provided a list of doors required to be closed for screening external flood scenarios and confirmed that these SSCs will be considered HSS in categorization. The licensee also clarified that temporary flood protection barriers are only needed for defense-in-depth and not for screening the external flooding hazard from this application. For snow loading on buildings, the licensee provided, in its supplement, details of the amount of snow which would be needed to challenge NMP1 structures and the largest snowfall total ever recorded on a single day to show a large margin available. The licensee also cited their severe weather procedure to show that snow would not be allowed to accumulate on roofs of critical buildings. Given the supplemental information, the NRC staff concludes that the external flooding and snowfall hazards were appropriately screened for this application.

In summary, the NRC staff finds that the use of the NMP1 IPEEE results described by the licensee in the LAR, and the licensee's assessment of the other external hazards (e.g., external flood, transportation accident) is consistent with Section 5 of NEI 00-04, Revision 0, as endorsed in RG 1.201, Revision 1. Therefore, the NRC staff concludes that the licensee's treatment of other external hazards is acceptable and meets 10 CFR 50.69(c)(1)(ii).

Shutdown Risk Consistent with the guidance in NEI 00-04, Revision 0, the licensee proposed using the shutdown safety assessment based on NUMARC 91-06 (Reference [16]). NUMARC 91-06

provides considerations for maintaining DID 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 DID 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 described by the licensee in the submittal is consistent with the guidance in NEI 00-04, Revision 0, as endorsed in the NRC in RG 1.201, Revision 1. The approach uses an integrated and systematic process to identify HSS components, consistent with the shutdown evaluation process. Therefore, 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. 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 of the enclosure to the LAR, the licensee proposed using a categorization method, for passive components not cited in NEI 00-04, Revision 0, or RG 1.201, Revision 1, for passive component categorization, but was approved by the NRC for ANO-2 (Reference [17]).

The ANO-2 methodology is a risk-informed safety classification and treatment program for repair/replacement activities for Class 2 and 3 pressure retaining items and their associated supports (exclusive of Class CC and MC items), using a modification of the ASME Code Case N-660, Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities,Section XI, Division 1 (Reference [22]). The ANO-2 methodology relies on the conditional core damage and large early release probabilities associated with pipe ruptures.

Safety significance is generally measured by the frequency and the consequence of, in this case, pipe ruptures. Treatment requirements (including 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. Therefore, the NRC staff finds that the use of the repair/replacement methodology is acceptable and appropriate for passive component categorization of Class 2 and Class 3 SSCs.

In Section 3.1.2 of the enclosure to the LAR, the licensee stated, [t]he passive categorization process is intended to apply the same risk-informed process accepted in the AN0 2-R&R-004 for the passive categorization of Class 2, 3, and non-class components. Consistent with ANO-2-R&R-004, Class 1 pressure retaining SSCs in the scope of the system being categorized will be assigned HSS and cannot be changed by the IDP. The NRC staff finds the licensee's proposed approach for passive categorization is acceptable for the 10 CFR 50.69 SSC categorization process.

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

Section 3.1.1 of the enclosure to the LAR 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,

Revision 0. Section 3.2.7 of the enclosure to 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, Revision 0, as endorsed by RG 1.201, Revision 1.

In Section 3.1.1 of the enclosure to the LAR, for the Overall Categorization Process, the licensee specifically noted that the implementation of all processes described in NEI 00-04 (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 discussed in Section 3.4 of this SE, assure 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, Revision 0, 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).

3.3.1.4 Integrated Decision-Making Appendix B of SRP Chapter 19, Section 19.2 provides guidance and the staff expectations for the licensees integrated decision-making process. The appendix states in part, [r]isk-informed applications are expected to require a process to integrate traditional engineering and probabilistic considerations to form the basis for acceptance. NEI 00-04 guidance 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 engineering 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 categorization. The staff review of the two steps to ensure the process is well-defined, systematic, repeatable, and scrutable are provided as follows:

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

All the information collected and evaluated in the licensees engineering evaluations is provided to the IDP as described in Section 7 of NEI 00-04, Revision 0. The IDP will make the final decision about the safety significance of SSCs based on guidelines in NEI 00-04, Revision 0, the information they receive, and their expertise.

In Section 3.1.1 of the enclosure to the LAR, the licensee stated, in part,... if any SSC is identified as HSS from either the integrated PRA component safety significance assessment (Section 5 of NEI 00-04) or the DID [defense-in-depth] assessment (Section 6), the associated system function(s) would be identified as HSS. The licensee also stated that, [o]nce a system function is identified as HSS, then all the components that support that function are preliminary HSS.

The NRC staff finds that the above description provided by the licensee for the preliminary categorization of functions is consistent with NEI 00-04, Revision 0, as endorsed in RG 1.201, Revision 1, and is therefore acceptable.

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

In Section 3.1.1 of the enclosure to the LAR, 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. Therefore, the IDP will comprise the required expertise.

The guidance in NEI 00-04, Revision 0, as endorsed in RG 1.201, Revision 1, 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). In Section 3.1.1 of the enclosure to the LAR, the licensee discusses that 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 modeling and updating of the plant-specific PRA. The licensee further states that the IDP will be trained in the specific technical aspects and requirements related to the categorization process. This 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 DID philosophy and requirements to maintain this philosophy. Based on its review, the NRC staff finds that the licensee's IDP areas of expertise meet the requirements in 10 CFR 50.69(c)(2) and the additional descriptions of the IDP characteristics, training, processes, and decision guidelines are consistent with NEI 00-04, Revision 0, as endorsed in RG 1.201, Revision 1.

As discussed in NEI 00-04, Revision 0, the only LSS SSC requirements that are relaxed for RISC-3 (LSS) SSCs are those related to treatment, not design or capability, and 10 CFR 50.69(d)(2)(i) requires the licensee ensures, 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 NMP1 categorization process, is consistent with the endorsed guidance in NEI 00-04, Revision 0, and, therefore, fulfills 10 CFR 50.69(c)(1)(iv).

Based on the above staff review for: (1) IEPRA, IFPRA, and FPRA acceptability, (2) PRA importance measures and integrated importance measure, (3) evaluation of the use of non-PRA methods, (4) risk sensitivity study, and (5) integrated decision-making, the staff has determined that the proposed change satisfies the fourth key principle for risk-informed decision-making prescribed in RG 1.174, Revision 3.

3.3.2 Key Principle 5: Monitor the Impact of the Proposed Change NEI 00-04, Revision 0 provides guidance that includes programmatic configuration control and a periodic review to ensure that the 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 overtime are continually incorporated.

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

Sections 11 and 12 of NEI 00-04, Revision 0, includes 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 NMP1 as-built, as-operated plant. A more detailed NRC staff review is provided as follows:

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. In Section 3.2.6 of the enclosure to the LAR, the licensee stated that it will implement a process that addresses the requirements in Section 11 of NEI 00-04, Revision 0, pertaining to program documentation and change control records. Section 3.1.1 of the enclosure to the LAR states that the RISC categorization process documentation will include the following ten elements:

• Program procedures used in the categorization

• System functions, identified and categorized with the associated bases

• Mapping of components to support function(s)

• PRA model results, including sensitivity studies

• Hazards analyses, as applicable

• Passive categorization results and bases

• Categorization results including all associated bases and RISC classifications

• Component critical attributes for HSS SSCs

• Results of periodic reviews and SSC performance evaluations

• IDP meeting minutes and qualification/training records for the IDP members The NRC staff also recognizes that for facilities licensed under 10 CFR Part 50, Appendix B Criterion VI, for Document Control, procedures are considered formal plant documents that require [m]easures shall be established to control the issuance of documents, such as instructions, procedures, and drawings, including changes thereto, which prescribe all activities affecting quality.

Based on its review of the LAR, the NRC staff finds the RISC categorization process for program documentation, change control and records, as described by the licensee in Section 3.1.1 of the enclosure to the LAR and in the list of categorization prerequisites provided in to the LAR as supplemented by the August 4, 2023, letter, is consistent with Section 11 of NEI 00-04, Revision 0, as endorsed by the NRC in RG 1.201, Revision 1, and is in conformance with the requirements of 10 CFR 50.69(f).

Periodic Review (NEI 00-04, Section 12)

Section 50.69(e), Feedback and Process Adjustment, of 10 CFR requires 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 of the enclosure to the LAR, the licensee described the process for maintaining and updating the NMP1 PRA models used for the 10 CFR 50.69 categorization process.

Consistent with NEI 00-04, the licensee confirmed that the NMP1 risk management process

ensures that the applicable PRA mode(s) used in this application continue to reflect the as-built and as-operated plant. The licensee's process includes provisions for: monitoring issues affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, industry operational experience); assessing the risk impact of unincorporated changes; and controlling the model and associated computer files. The process also includes reevaluating previously categorized systems to ensure the continued validity of the categorization.

Routine PRA updates are performed every two refueling cycles at a minimum. The NRC staff finds the risk management process described by the licensee in the LAR is consistent with Section 12 of NEI 00-04, Revision 0 guidance as endorsed by the NRC.

The NRC staff finds the risk management process described by the licensee in the LAR as updated is consistent with Sections 11 and 12 of NEI 00-04, Revision 0, as endorsed by RG 1.201, Revision 1, and consistent with the requirements in 10 CFR 50.69(e). 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, Revision 3.

4.0 CHANGES TO THE OPERATING LICENSE The licensee proposed the following amendment to the Renewed Facility Operating License for NMP1. The proposed License Condition 2.D.(27) would state:

(27)

Implementation of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Constellation Energy Generation, LLC 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, and internal fire; high wind safe shutdown equipment list to evaluate high wind / tornado missile events; 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 and Class 3 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the Individual Plant Examination of External Events (IPEEE) Screening Assessment for External Hazards updated using the external hazard screening significance process identified in American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in Constellation Energy Generation, LLC's submittal letter dated, December 15, 2022, and all its subsequent associated supplements as specified in License Amendment No. 251 dated December 7, 2023.

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 condition is acceptable, because: (1) 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; and (2) the evaluation in SE Section 3.3.1.2, finds the non-PRA methods for assessing risk for 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, 2004,4 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 technical specifications (including Improved technical specifications) and the associated technical requirements manual to be part of the 10 CFR 50.69 rule. Therefore, the licensee must address proposed changes to its technical specifications separately.

5.0 STATE CONSULTATION

In accordance with the Commissions regulations, the New York State official was notified of the proposed issuance of the amendment on October 11, 2023. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves 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 amendment involves no significant hazards consideration, as published in Federal Register on February 21, 2023 (88 FR 10558), and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

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

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 Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

4 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.

8.0 REFERENCES

[1] "Gudger, D.T., Constellation Energy Generation, "License Amendment Request to Adopt 10 CFR 50.69, 'Risk-informed categorization and treatement of structures, systems and components for nuclear power reactors"," (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22349A521), December 15, 2022.

[2] "Gudger, D.T., Constellation Energy Generation, LLC to US NRC, "Supplemental Information Letter for Nine Mile Point Nuclear Station, Unit 1 to Adopt TSTF-505 and 10 CFR 50.69"," (ML23216A110), August 4, 2023.

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

[4] Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed Activities," Revision 2, March 2009 (ML090410014).

[5] Regulatory Guide 1.174, Revision 3, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis", January 2018 (ML17317A256).

[6] NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision-making," 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, Issued June 2007, ML071700658.

[8] NEI 00-04, Revsion 0, "10 CFR 50.69 SSC Categorization Guideline", July 2005 (ML052910035).

[9] "Electric Power Research Institute (EPRI) Report 3002017583, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization,,"

(ML21082A170), February 2020.

[10] U.S. Nuclear Regulatory Commission, "Individual Plant Examination of External Events (IPEEEs) for Severe Accident Vulnerabilities - 10 CFR 50.54(f), Generic Letter 88-20, June 1991.

[11] "Vaidya, B., U.S. Nuclear Regulatroy Commission, letter to Christopher Costanzo, Exelon Generation Company, LLC, "Nine Mile Point Nuclear Station, Unit No. 1 - Issuance of Amendment Regarding Transition to a Risk-Informed, Performance-Based Fire Protection," (ML14126A003), June 30, 2014.

[12] "Mozafari, B.L., U.S. Nuclear Regulatory Commission, letter to Bryan Hanson, Exelon Nuclear, LLC, "Nine Mile Point Nuclear Station, Unit 1 - Issuance of Amendment RE:

Adoption of Technical Specification Task Force Traveler 425, Revision 3 (CAC No.

MF6061),," (ML16081A256), May 31, 2016.

[13] American Society of Mechnical Engineers (ASME) and American Nuclear Society (ANS)

PRA standard 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).

[14] Anderson, V.K., Nuclear Energy Institute, letter to Stacey Rosenbergy, U.S. Nuclear Regulatory Commission, "Final Revision of Appendix X to NEI 05-04/07-12-12-16, Close-Out of Facts and Observations,", February 21, 2017 (ML17086A431).

[15] "U.S. NRC memorandum, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Risk Assessments,," (ML22014A084), May 6, 2022.

[16] NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown Management",

December 1991 (ML14365A203).

[17] 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 AN02-R&R-004, Revision 1, Request to Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activites in Class 2 and 3, Moderate and High Energy Safety Systems", April 22, 2009 (ML090930246).

[18] "Simpson, Patrick R., Exelon Generation Company, LLC, letter to U.S. Nuclear Regulatory Commission, Clinton Power Station, Unit 1, Response to Request for Additional Information Regarding License Amendment Requests to Adopt TSTF-425 and 10 CFR 50.69"," (ML20329A433), November 24, 2020.

[19] "Marshal, M. L., U.S. Nuclear Regulatory Commission letter to Senior Vice President,

Exelon Generation Company, LLC, "Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Issuance of Amendment Nos. 332 and 310," (ML19330D909), February 28, 2020.

[20] "Constellation Letter to U.S. NRC "Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the NTTF Review of the Fukushima Dai-ichi Accident," Nine Mile Point Nuclear Station Units 1 and 2," (ML14099A196), March 31, 2014.

[21] "Gudger, David, T., Constellation Energy Generation, LLC, letter to U.S. NRC, Nine Mile Point Nuclear Station, Unit 1, "License Amendment Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505"," (ML22349A108),

December 15, 2022.

[22] American Society of Mechanical Engineers, "Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities, ASME Code Case, N-660", July 2002.

Principal Contributors: J. Patel, NRR J. Circle, NRR S. Rosenberg, NRR D. Wu, NRR H. Kodali, NRR M. Li, NRR R. Hernandez, NRR D. Nold, NRR G. Bedi, NRR S. Cumblidge, NRR M. Benson, NRR S. Haider, NRR Date: December 7, 2023

ML23291A464 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DRA/APLA/BC NRR/DRA/APLC/BC NAME RGuzman KEntz RPascarelli SVasavada DATE 10/27/2023 10/27/2023 10/02/2023 10/02/2023 OFFICE NRR/DEX/EEEB/BC NRR/DEX/EICB/BC(A)

NRR/DEX/EMIB/BC NRR/DSS/SCPB/BC NAME WMorton RStattel SBailey BWittick DATE 10/19/2023 10/05/2023 10/20/2023 10/11/2023 OFFICE NRR/DSS/SNSB/BC NRR/DNRL/NPHP/BC NRR/DNRL/NVIB/BC OGC (NLO)

NAME PSahd MMitchell ABuford AGhosh DATE 10/06/2023 10/10/2023 10/04/2023 11/09/2023 OFFICE NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME HGonzález RGuzman DATE 12/7/2023 12/7/2023