Prairie Island Nuclear Generating Plant, Units 1 and 2 ...Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendments 230 and 218 RE: Issuance of Amendments RE:

November 12, 2019 Mr. Scott Sharp Site Vice President Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota 1717 Wakonade Drive East Welch, MN 55089 SUBJECT: PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND

2 - ISSUANCE OF AMENDMENTS RE: ADOPTION OF 10 CFR 50.69 – “RISK-INFORMED CATERIZATION AND TREATMENT OF STRUCTURE, SYSTEMS AND COMPONENTS OF NUCLEAR POWER REACTORS” (EPID L-2018-LLA-0196)

Dear Mr. Sharp: The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 230 to Renewed Facility Operating License No. DPR-42 and Amendment No. 218 to Renewed Facility Operating License No. DPR-60 for the Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2, respectively. The amendments consist of changes to the licenses in response to your application dated July 20, 2018, as supplemented by letters dated April 29, 2019, and August 5, 2019. The amendment adds a condition to the PINGP, Units 1 and 2, renewed facility operating license to allow 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 our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Sincerely,

/RA/

Robert F. Kuntz, Senior Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Docket Nos. 50-282 and 50-306 Enclosures: 1. Amendment No. 230 to DPR-42 2. Amendment No. 218 to DPR-60 3. Safety Evaluation cc: via Listserv

Enclosure 1

NORTHERN STATES POWER COMPANY

DOCKET NO. 50-282

PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 1

AMENDMENT TO RENEWED FACILITY OPERATING LICENSE

Amendment No. 230 License No. DPR-42

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

A. The application for amendment by Northern States Power Company, a Minnesota Corporation (NSPM, the licensee), dated July 20, 2018, as supplemented by letters dated April 29, 2019, and August 5, 2019, 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 rules and regulations of the Commission;

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations;

D. The issuance of this amendment will not be inimical to the common defense and

security or to the health and safety of the public; and

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Renewed Facility Operating

License as indicated in the attachment to this license amendment.

- 2 -

3. This license amendment is effective as of the date of its issuance and shall be

implemented within 90 days.

FOR THE NUCLEAR REGULATORY COMMISSION /RA/ Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment: Changes to the Renewed Facility Operating License Date of Issuance: November 12, 2019

Enclosure 2


DOCKET NO. 50-306

PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 2

AMENDMENT TO RENEWED FACILITY OPERATING LICENSE

Amendment No. 218 License No. DPR-60

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

A. The application for amendment by Northern States Power Company, a Minnesota Corporation (NSPM, the licensee), dated July 20, 2018, as supplemented by letters dated April 29, 2019, and August 5, 2019, 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 rules and regulations of the Commission;

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations;

D. The issuance of this amendment will not be inimical to the common defense and

security or to the health and safety of the public; and

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Renewed Facility Operating

License as indicated in the attachment to this license amendment.

- 2 -

3. This license amendment is effective as of the date of its issuance and shall be

implemented within 90 days.

FOR THE NUCLEAR REGULATORY COMMISSION /RA/ Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment: Changes to the Renewed Facility Operating License Date of Issuance: November 12, 2019

ATTACHMENT TO LICENSE AMENDMENT NOS. 230 AND 218

RENEWED FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60

DOCKET NOS. 50-282 AND 50-306 Replace the following pages of the Renewed Facility Operating License Nos. DPR-42 and DPR-60 with the attached revised pages. The changed areas are identified by a marginal line.

DPR-42

REMOVE INSERT Page -8- Page -8- ----------- Page -9-

DPR-60

REMOVE INSERT Page -8- Page -8- ----------- Page -9-

- 8 -

Renewed Operating License No. DPR-42 Amendment No. 230

b. The first performance of the periodic assessment of CRE habitability, Specification 5.5.16.c (ii), shall be 3 years, plus the 9 month allowance of SR 3.0.2, as measured from December 3, 2004, the date of the most recent successful tracer gas test, as stated in the December 18, 2006 letter in response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(8) License Renewal License Conditions

(a) The licensee may make changes to the programs and activities described

in the USAR supplement, submitted pursuant to 10 CFR 54.21(d), as revised during the license renewal application review process, provided 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.

(b) Appendix A of “Safety Evaluation Report Related to the License Renewal

of Prairie Island Nuclear Generating Plant, Units 1 and 2,” dated October 16, 2009, and supplemented on April 15, 2011, and the licensee’s USAR supplement submitted pursuant to 10 CFR 54.21(d) describe certain future programs and activities to be completed before the period of extended operation. The licensee shall complete these activities no later than August 9, 2013, and shall notify the NRC in writing when implementation of these activities is complete.

(c) All capsules in the reactor vessel that are removed and tested must meet

the test procedures and reporting requirements of American Society for Testing and Materials (ASTM) E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion. Any changes to storage requirements must be approved by the NRC.

(9) Adoption of 10 CFR 50.69, "Risk-informed categorization and treatment of

structures, systems, and components for nuclear power plants"

NSPM is approved to implement 10 CFR 50.69 using the approaches 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, with 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 SSCs and their associated supports; and the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards (e.g., external flooding and high winds) updated using the external hazard screening significance criteria identified in

- 9 -


ASME/ANS PRA Standard RA-Sa-2009, as endorsed in RG 1.200, Revision 2; as specified in PINGP License Amendment No. 230 dated November 12, 2019.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization approach specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach). NSPM will complete the implementation items listed in Attachment 1 of NSPM’s letter to the NRC dated August 5, 2019, prior to implementation of 10 CFR 50.69. NSPM shall ensure that the fire PRA model used for the 10 CFR 50.69 SSC categorization reflects the as-built, as-operated plant using the same fire PRA model used to support NFPA 805 implementation for both PINGP units prior to implementation of 10 CFR 50.69.

D. This renewed operating license is effective as of the date of issuance and shall expire at

midnight August 9, 2033.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Eric J. Leeds, Director Office of Nuclear Reactor Regulation

Attachments: 1. Appendix A - Technical Specifications 2. Appendix B - Additional Conditions Date of Issuance: June 27, 2011

- 8 -


(8) License Renewal License Conditions

(a) The licensee may make changes to the programs and activities described in the USAR supplement, submitted pursuant to 10 CFR 54.21(d), as revised during the license renewal application review process, provided 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.

(b) Appendix A of “Safety Evaluation Report Related to the License Renewal of

Prairie Island Nuclear Generating Plant, Units 1 and 2,” dated October 16, 2009, and supplemented on April 15, 2011, and the licensee’s USAR supplement submitted pursuant to 10 CFR 54.21(d) describe certain future programs and activities to be completed before the period of extended operation. The licensee shall complete these activities no later than October 29, 2014, and shall notify the NRC in writing when implementation of these activities is complete.

(c) All capsules in the reactor vessel that are removed and tested must meet

the test procedures and reporting requirements of American Society for Testing and Materials (ASTM) E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion. Any changes to storage requirements must be approved by the NRC.

(9) Adoption of 10 CFR 50.69, "Risk-informed categorization and treatment of

structures, systems, and components for nuclear power plants"

NSPM is approved to implement 10 CFR 50.69 using the approaches 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, with 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 SSCs and their associated supports; and the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards (e.g., external flooding and high winds) updated using the external hazard screening significance criteria identified in ASME/ANS PRA Standard RA-Sa-2009, as endorsed in RG 1.200, Revision 2; as specified in PINGP License Amendment No. 218 dated November 12, 2019. Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization approach specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

- 9 -


NSPM will complete the implementation items listed in Attachment 1 of NSPM’s letter to the NRC dated August 5, 2019, prior to implementation of 10 CFR 50.69. NSPM shall ensure that the fire PRA model used for the 10 CFR 50.69 SSC categorization reflects the as-built, as-operated plant using the same fire PRA model used to support NFPA 805 implementation for both PINGP units prior to implementation of 10 CFR 50.69.

D. This renewed operating license is effective as of the date of issuance and shall

expire at midnight October 29, 2034.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Eric J. Leeds, Director Office of Nuclear Reactor Regulation

Attachments: 1. Appendix A - Technical Specifications 2. Appendix B - Additional Conditions Date of Issuance: June 27, 2011

Enclosure 3

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

RELATED TO AMENDMENT NO. 230 TO RENEWED FACILITY

OPERATING LICENSE NO. DPR-42

AND AMENDMENT NO. 218 TO RENEWED FACILITY

OPERATING LICENSE NO. DPR-60


PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2

DOCKET NOS. 50-282 AND 50-306 1.0 INTRODUCTION By letter dated July 20, 2018 (Reference 1), as supplemented by letters dated April 29, 2019

(Reference 2), and August 5, 2019 (Reference 3), Northern States Power Company (NSPM, the licensee) submitted a license amendment request (LAR) for the Prairie Island Nuclear Generating Plant, Units 1 and 2 (PINGP). The licensee proposed to add a new license condition to the Renewed Facility Operating Licenses 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.” The provisions of 10 CFR 50.69 allow adjustment of the scope of structures, systems, and components (SSCs) subject to special treatment requirements (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation) based on a method of categorizing SSCs according to their safety significance. For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance requirements will not be changed or will be enhanced. The supplemental letters dated April 29, 2019, and August 5, 2019, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff’s original proposed no significant hazards consideration determination as published in the Federal Register (FR) September 11, 2018 (83 FR 45986). 2.0 REGULATORY EVALUATION 2.1 Risk-Informed Categorization and Treatment of SSCs A risk-informed (RI) approach to regulation enhances and extends the traditional deterministic regulation by considering risk in a comprehensive manner. Specifically, an RI approach allows

- 2 -

consideration of a broader set of potential challenges to safety, providing a logical means for prioritizing these challenges based on safety-significance, and allowing consideration of a broader set of resources to defend against these challenges. Probabilistic risk assessments (PRAs) address credible initiating events by assessing the event frequency. Mitigating system reliability is then assessed including the potential for common cause failures. To take advantage of the safety enhancements available through the use of PRA, the U.S. Nuclear Regulatory Commission (NRC or Commission) promulgated a new regulation, 10 CFR 50.69, in the Federal Register on November 22, 2004 (69 FR 68008), which became effective on December 22, 2004. 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, alternative treatment requirements may be implemented in accordance with the regulation. Section 50.69 of 10 CFR contains requirements regarding how a licensee categorizes SSCs using an RI process, adjusts treatment requirements consistent with the relative safety significance of the SSC, and manages the process over the lifetime of the plant. An RI categorization process is employed to determine the safety-significance of SSCs and place the SSCs into one of four RI safety class (RISC) categories. The determination of safety-significance is performed by an integrated decision-making process, which uses both risk and traditional engineering insights. The safety functions include the design-basis functions, as well as functions credited for severe accidents (including external events). Special or alternative treatment for the SSCs is applied as necessary to maintain functionality and reliability, and is a function of the SSC categorization results and associated bases. Finally, periodic assessment activities are conducted to make adjustments to the categorization and/or treatment processes as needed so that SSCs continue to meet all applicable functional requirements. Section 50.69 of 10 CFR 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. In 2004, when promulgating the 10 CFR 50.69 rule, the Commission stated:

It is important to note that this rulemaking effort, while intended to ensure that the scope of special treatment requirements imposed on SSCs is risk-informed, is not intended to allow for the elimination of SSC functional requirements or to allow equipment that is required by the deterministic design basis to be removed from the facility (i.e., changes to the design of the facility must continue to meet the current requirements governing design change; most notably § 50.59). Instead, this rulemaking should enable licensees and the staff to focus their resources on SSCs that make a significant contribution to plant safety by restructuring the regulations to allow an alternative risk-informed approach to special treatment. Conversely, for SSCs that do not significantly contribute to plant safety on an individual basis, this approach should allow an acceptable, though reduced, level of confidence (i.e., “reasonable confidence”) that these SSCs will satisfy functional requirements. However, continued maintenance of the health and safety of the public will depend on effective implementation of § 50.69 by the licensee or applicant applying the rule at its nuclear power plant.

- 3 -

Final Rule, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, 69 Fed. Reg. 68008, 68011 (Nov. 22, 2004). For SSCs that are categorized as high safety-significant (HSS)* existing treatment requirements are maintained for safety-related equipment or are potentially enhanced for nonsafety-related SSCs. Conversely, for SSCs categorized as low safety-significant (LSS) that do not significantly contribute to plant safety on an individual basis, the regulation allows an alternative RI approach to treatment that provides a reasonable, although reduced, level of confidence that these SSCs will satisfy functional requirements. 2.2 Licensee’s Proposed Changes The letter dated August 5, 2019, proposed to amend the PINGP Renewed Facility Operating Licenses by adding the following license condition that would allow for the implementation of 10 CFR 50.69:

NSPM is approved to implement 10 CFR 50.69 using the approaches 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, with 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 SSCs and their associated supports; and the results of non-PRA evaluations that are based on the IPEEE [Individual Plant Examination of External Events] Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards (e.g., external flooding and high winds) updated using the external hazard screening significance criteria identified in ASME/ANS PRA Standard RA-Sa-2009, as endorsed in RG 1.200, Revision 2; as specified in PINGP License Amendment No. [XXX] dated [DATE]. Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization approach specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach). NSPM will complete the implementation items listed in Attachment 1 of NSPM’s letter to the NRC dated August 5, 2019, prior to implementation of 10 CFR 50.69. NSPM shall ensure that the fire PRA model used for the 10 CFR 50.69 SSC categorization reflects the as-built, as-operated plant using the same fire PRA model used to support NFPA [National Fire Protection Association] 805 implementation for both PINGP units prior to implementation of 10 CFR 50.69.

2.3 Regulatory Review The NRC staff reviewed the LAR, as supplemented, to determine whether: (1) there is reasonable assurance that the health and safety of the public will not be endangered by * The Nuclear Energy Institute (NEI) 00 04 uses the term “high safety significant (HSS)” 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.

- 4 -

operation in the proposed manner, (2) activities proposed will be conducted in compliance with the Commission’s regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or the health and safety of the public. The NRC staff considered the following regulatory requirements and guidance during its review of the proposed changes. Regulatory Requirements Section 50.69 of 10 CFR provides an alternative approach for establishing requirements for treatment of SSCs for nuclear power reactors using a RI approach to categorize SSCs according to their safety significance. Paragraph 50.69(b)(2) of 10 CFR states that a licensee voluntarily choosing to implement this section shall submit an application for license amendment under 10 CFR 50.90 that contains the following information: (i) a description of the process for categorization of RISC–1, RISC–2, RISC–3 and RISC–4 SSCs; (ii) a description of the measures taken to assure that the quality and level of detail of the systematic processes that evaluate the plant for internal and external events during normal operation, low power, and shutdown (including the plant-specific PRA, margins-type approaches, or other systematic evaluation techniques used to evaluate severe accident vulnerabilities) are adequate for the categorization of SSCs; (iii) results of the PRA review process conducted to meet 10 CFR 50.69(c)(1)(i); and (iv) a description of, and basis for acceptability of, the evaluations to be conducted to satisfy 10 CFR 50.69(c)(1)(iv). Paragraph 50.69(b)(3) of 10 CFR states that the Commission will approve a licensee’s 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)(1), 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) the portion of 10 CFR 50.46a(b) that imposes requirements to conform to Appendix B to 10 CFR part 50, (iii) 10 CFR 50.49, (iv) 10 CFR 50.55(e), (v) certain 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 to 10 CFR Part 50, (x) certain containment leakage testing requirements, and (xi) certain requirements of Appendix A to 10 CFR Part 100. Paragraph 50.69(c) of 10 CFR requires licensees to use an integrated decisionmaking 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 functions

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

RISC-3: Safety-related SSCs that perform LSS functions

RISC-4: Nonsafety-related SSCs that perform LSS 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 RISC-1 SSCs, 10 CFR 50.69 maintains current regulatory requirements (i.e., it does not remove any requirements from these

- 5 -

SSCs) for special treatment. For RISC-2 SSCs, 10 CFR 50.69 provides that existing treatment should support the key assumptions in the categorization process that relate to their assumed performance or treatment enhancements may be needed. 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(c)(1) of 10 CFR states that:

SSCs must be categorized as RISC-1, RISC-2, RISC-3, or RISC-4 SSCs, using a categorization process that determines if an SSC performs one or more safety-significant functions and identifies those functions. The process must:

(i) Consider results and insights from the plant-specific PRA. This PRA

must at a minimum model severe accident scenarios resulting from internal initiating events occurring at full power operation. The PRA must be of sufficient quality and level of detail to support the categorization process, and must be subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC.

(ii) Determine SSC functional importance using an integrated, systematic

process for addressing initiating events (internal and external), SSCs, and plant operating modes, including those not modeled in the plant-specific PRA. The functions to be identified and considered include design bases functions and functions credited for mitigation and prevention of severe accidents. All aspects of the integrated, systematic process used to characterize SSC importance must reasonably reflect the current plant configuration and operating practices, and applicable plant and industry operational experience.

(iii) Maintain defense-in-depth [DID]. (iv) Include evaluations that 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 permitted by implementation of § § 50.69(b)(1) and (d)(2) are small.

(v) Be performed for entire systems and structures, not for selected

components within a system or structure.

Paragraph 50.69(c)(2) of 10 CFR states: “The SSCs must be categorized by an Integrated Decision-Making Panel (IDP) staffed with expert, plant-knowledgeable members whose expertise includes, at a minimum, PRA, safety analysis, plant operation, design engineering, and system engineering.” Paragraph 50.69(d) of 10 CFR provides requirements for ensuring categorized SSCs retain assumed functional performance. The licensee or applicant shall ensure that RISC-1 and RISC-2 SSCs, perform their functions consistent with the categorization process assumptions by evaluating treatment being applied to these SSCs to ensure that it supports the key assumptions in the categorization process that relate to their assumed performance. For

- 6 -

RISC-3 SSCs periodic inspection and testing shall be performed to ensure SSCs remain capable of performing their safety-related functions under design basis conditions. Conditions that would prevent a RISC-3 SSC from performing its safety-related functions under design basis conditions must be corrected in a timely manner. Paragraph 50.69(e) of 10 CFR describes feedback and process adjustment requirements for categorized SSCs. The licensee shall review changes to the plant, operational practices, applicable plant and industry operational experience, and, as appropriate, update the PRA and SSC categorization and treatment processes. The licensee shall perform this review in a timely manner but no longer than once every two refueling outages. Paragraph 50.69(f) of 10 CFR describes program documentation, change control and records requirements.

Paragraph 50.69(g) of 10 CFR describes reporting requirements associated with the categorization process. The licensee shall submit a licensee event report under 10 CFR 50.73(b) for any event or condition that prevented, or would have prevented, a RISC-1 or RISC-2 SSC from performing a safety significant function.

Guidance Nuclear Energy Institute (NEI) 00-04, Revision 0, “10 CFR 50.69 SSC Categorization Guideline” (NEI 00-04,Reference 4), describes a process for determining the safety-significance of SSCs and categorizing them into the four RISC categories defined in 10 CFR 50.69. Regulatory Guide (RG) 1.201 (For Trial Use), Revision 1, “Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance” (RG 1.201, Reference 5) endorses the categorization approach described in NEI 00-04, with clarifications, limitations, and conditions. The categorization process in NEI 00-04 is an integrated decision-making process that incorporates risk and traditional engineering insights. The guidance in NEI 00-04 provides options for licensees implementing different approaches depending on the scope of their PRA models. It also allows the use of non-PRA approaches when PRAs have not been performed. The guidance in NEI 00-04 identifies non-PRA approaches such as fire-induced vulnerability evaluation to address fire risk, seismic margin analysis (SMA) to address seismic risk, and guidance in Nuclear Management and Resource Council (NUMARC) 91-06, “Guidelines for Industry Actions to Assess Shutdown Management”

(Reference 6), to address shutdown operations. As stated in RG 1.201, such non-PRA-type evaluations will result in more conservative categorization, in that special treatment requirements will not be allowed to be relaxed for SSCs that are relied upon in such evaluations. Sections 2 through 10 of NEI 00-04 describe a method for meeting the requirements of 10 CFR 50.69(b)(2) and 10 CFR 50.69(c), as follows:

Section 2 provides expectations to address 10 CFR 50.69(b)(2)(i) Sections 3.2, 3.3 provide expectations and general guidance for assessment of PRA

scope and technical capability to address 10 CFR 50.69(b)(2)(ii) and (iii) Sections 8 provides specific guidance corresponding to 10 CFR 50.69(b)(iv) Sections 3.2 and 5.1 provide specific guidance corresponding to 10 CFR 50.69(c)(1)(i). Sections 3, 4, 5, and 7 provide specific guidance corresponding to

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

- 7 -

Section 8 provides specific guidance corresponding to 10 CFR 50.69(c)(1)(iv). Section 2 provides specific guidance corresponding to 10 CFR 50.69(c)(1)(v). Sections 9 and 10 provide specific guidance corresponding to 10 CFR 50.69(c)(2).

Additionally, Section 11 of NEI 00-04 provides guidance on program documentation and change control related to the requirements of 10 CFR 50.69(f), and Section 12 of NEI 00-04 provides guidance on 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 plant configuration and operating practices, and applicable plant and industry operational experience, as required by 10 CFR 50.69(c)(1)(ii). RG 1.201 states that licensees or applicants may use either PRAs or alternative approaches for hazards other than internal initiating events (e.g., the approach if a seismic margins analysis is relied upon is different and more limiting than the approach if a seismic PRA is used). The RG stipulates that such non-PRA-type evaluations will result in more conservative categorization, in that special treatment requirements will not be allowed to be relaxed for SSCs that are relied upon in such evaluations. The degree of relaxation that the NRC will accept under 10 CFR 50.69 (i.e., SSCs subject to relaxation of special treatment requirements) will be commensurate with the assurance provided by the evaluation. The licensee is expected to document the technical acceptability of the internal initiating events PRA as well as the bases for why any other PRA or non-PRA approaches employed to assign safety-significance to SSCs are technically acceptable. As stated in RG 1.201, as part of the NRC’s review and approval of a licensee’s or applicant’s application requesting to implement 10 CFR 50.69, the NRC staff intends to impose a license condition that will explicitly address the scope of the PRA and non-PRA approaches used in the licensee’s categorization approach. If a licensee or applicant wishes to change its categorization approach and the change is outside the bounds of the NRC’s license condition (e.g., switch from a seismic margins analysis to a seismic PRA), the licensee or applicant will need to seek NRC approval, via a license amendment, of the implementation of the new approach in their categorization process. In addition, RG 1.201 states 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). RG 1.200, Revision 2, “An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities” (RG 1.200, Reference 7), describes an acceptable approach for determining whether the technical acceptability of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision making for light-water reactors. The RG endorses, with clarifications, the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA Standard ASME/ANS RA-Sa-2009 (ASME/ANS 2009 PRA Standard or PRA Standard, Reference 8). This RG provides guidance for determining the technical acceptability of a PRA by comparing the PRA to the relevant parts of the ASME/ANS 2009 Standard using a peer review process. In accordance with the guidance, peer reviews should be used for PRA upgrades. A PRA upgrade is defined in the PRA Standard as “the incorporation into a PRA model of a new methodology or significant changes in scope or capability that impact the significant accident sequences or the significant accident progression sequences.” 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 9) provides guidance on the use of PRA findings and risk insights in support of changes to a plant’s licensing basis. This RG provides risk acceptance guidelines for evaluating the results of such evaluations.

- 8 -

3.0 TECHNICAL EVALUATION 3.1 Staff’s Method of Review The NRC staff evaluated the LAR, as supplemented, to determine if the proposed changes are consistent with the regulations and guidance discussed in Section 2 of this safety evaluation (SE). The LAR stated that the proposed change implements the provisions of 10 CFR 50.69 consistent with NEI 00-04, as endorsed by RG 1.201. Therefore, the NRC staff’s review, as documented in this SE, uses the framework of NEI 00-04, as endorsed by RG 1.201. Section 3.2 of this SE evaluates the licensee’s overview of the categorization process (Section 2 of NEI 00-04). Sections 3.3 to 3.9 of this SE evaluate each of the following primary categorization process steps described in Section 2 of NEI 00-04:

assembly of plant-specific inputs (Section 3 of NEI 00-04) system engineering assessment (Section 4 of NEI 00-04) component safety significance assessment (Section 5 of NEI 00-04) DID assessment (Section 6 of NEI 00-04) preliminary engineering categorization of functions (Section 7 of NEI 00-04) risk sensitivity study (Section 8 of NEI 00-04) IDP review and approval (Section 9 of NEI 00-04) SSC categorization (Section 10 of NEI 00-04)

Section 3.10 of this SE provides the NRC evaluation of the licensee’s proposed program documentation, change control, and periodic review (NEI 00-04, Sections 11 and 12). 3.2 Overview of the Categorization Process (NEI 00-04, Section 2) Paragraph 50.69(b)(2)(i) of 10 CFR states that a licensee voluntarily choosing to implement 10 CFR 50.69 shall submit an application for license amendment under 10 CFR 50.90 that contains a description of the process for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs. Paragraph 50.69(c)(1)(v) of 10 CFR states that the process for categorization must be performed for entire systems and structures, not for selected components within a system or structure. Section 2 of NEI 00-04 summarizes each of the primary categorization process steps and provides an overview of the categorization process in Figure 2-1. The guidance in RG 1.201 provides that the categorization process described in NEI 00-04, with any noted exceptions or clarifications, is acceptable for implementation of 10 CFR 50.69. The LAR provided a summary of the categorization process including a description of how:

system boundaries and functions are defined; components are assigned to system functions; safety-significance is assessed for active and passive components; DID and qualitative considerations are assessed;

- 9 -

cumulative sensitivity studies are used to confirm that the population of LSS components results in acceptably small increases to CDF and LERF; and

the IDP panel reviews preliminary categorization results and assigns final categorizations.

Table 3-1 of Section 3.1 of the LAR provided details on how some steps of the categorization process are performed at the component level (e.g., all PRA and non-PRA-modeled hazards, containment DID, passive categorization), the function level (e.g., qualitative criteria), or the function and component level (e.g., shutdown, core damage DID). LAR Section 3.1.1 stated that consistent with NEI 00-04, the categorization of a component or function is “preliminary” until it has been confirmed by the IDP review (see also Section 3.9 of this SE). The LAR section includes a discussion and associated Table (Table 3-1) summarizing what mechanisms are available to assign functions and components to preliminarily HSS and how this designation can, or cannot, be changed to LSS by the IDP as appropriate. As illustrated in LAR Table 3-1, all components that are assigned HSS based on risk significance as determined by the internal events importance measures, non-PRA risk models, DID considerations, passive categorization, or cumulative impact of the qualitative considerations must be assigned HSS. Components assigned preliminary HSS based on other PRA modelled risk hazards (e.g., fire and seismic PRAs) may be assigned LSS by the IDP. The licensee stated that a function is categorized as HSS if any component supporting the function is assigned HSS based on risk-significance. The qualitative considerations generally categorize functions directly and therefore, if the IDP determines that any one of the seven considerations is false for a system function, then the function will be assigned HSS. Once a system function is identified as HSS, then all the components supporting that function are preliminary HSS and will be presented to the IDP for review. Some components supporting HSS functions may not be necessary to achieve success of the safety significant mitigating function and these components may not be required to be HSS. Therefore, any component supporting an HSS function but not requiring an HSS assignment based on risk significance may be assigned LSS by the IDP. The NRC staff reviewed the overview of categorization steps and concludes that it meets the requirements of 10 CFR 50.69(b)(2)(i) and 50.69(c)(1)(v) because it is consistent with NEI 00-04, Section 2, guidance, as endorsed by RG 1.201. 3.3 Assembly of Plant-Specific Information (NEI 00-04, Section 3) Section 3 of NEI 00-04 states that the assembly of plant-specific inputs involves the collection and assessment of the key inputs to the RI categorization process. This includes design and licensing information, PRA analyses, and other relevant plant data sources. In addition, this step includes the critical evaluation of plant-specific risk information to ensure that the information is adequate to support this application. The guidance in Section 3 of NEI 00-04 summarizes the use of risk information and the general quality measures that should be applied to the risk analyses supporting the 10 CFR 50.69 categorization. These quality measures include characterization of technical acceptability of both the internal events at-power PRA and other risk analyses necessary to implement 10 CFR 50.69. The LAR stated that the risk categorization process uses PRAs to assess risks from internal events (including flooding), and from fire. For the other applicable hazard groups, the licensee’s process uses non-PRA methods for the risk categorization. The LAR stated that the risk categorization process uses the PINGP Seismic Safe Shutdown Equipment List (SSEL) from

- 10 -

the SMA to assess seismic risk, the individual plant examination of external events (IPEEE) screening process to assess other external hazards (high winds and external floods), and the qualitative DID shutdown model to assess shutdown risk. The use of risk information and quality of the PRA is reviewed in Section 3.5 of this SE. The NRC staff reviewed the description of assembly of plant-specific information to support the categorization process and concludes it is consistent with NEI 00-04, Section 3, guidance as endorsed by RG 1.201 and, therefore, meets the requirements of 10 CFR 50.69(b)(2)(ii) and (iii) and applicable portions of 10 CFR 50.69(c)(1)(i) and (ii). 3.4 System Engineering Assessment (NEI 00-04, Section 4) Paragraph 50.69(c)(1)(ii) of 10 CFR requires licensees to determine SSC functional importance using an integrated, systematic process for addressing initiating events (internal and external), SSCs, and plant operating modes, including those not modeled in the plant-specific PRA. The functions to be identified and considered include design-basis functions and functions credited for mitigation and prevention of severe accidents. Section 4 of NEI 00-04 provides guidance for developing a systematic engineering assessment involving the identification and development of base information necessary to perform the RI categorization. The assessment includes the following elements: system selection and system boundary definition, identification of system functions, and a mapping of components to functions. Section 4 of NEI 00-04 states that system selection and boundary definition include defining system boundaries, where the system interfaces with other systems. The guidance in NEI 00-04 states that the next step is the identification of system functions, including design basis and beyond design basis functions identified in the PRA, and that system functions should be consistent with the functions defined in design-basis documentation and maintenance rule functions. The guidance in NEI 00-04 states that the coarse mapping of components to functions involves the initial breakdown of system components into system functions they support. The licensee should then identify and document system components and equipment associated with each function. The guidance in NEI 00-04 also includes consideration of interfacing functions. Section 7.1 of the NEI 00-04 guidance states in part, “[d]ue to the overlap of functions and components, a significant number of components support multiple functions. In this case, the SSC or part thereof should be assigned the highest risk significance for any function that the SSC or part thereof supports.” In addition, Section 4 of NEI 00-04 states in part, “[t]here may be circumstances where the categorization of a candidate low safety-significant SSC within the scope of the system being considered cannot be completed because it also supports an interfacing system. In this case, the SSC will remain uncategorized until the interfacing system is categorized.” Section 2.2 of the LAR states that the 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 LAR summarizes the different hazards and plant states for which functional and risk significant information will be collected. Section 3.1.1 of the LAR also states 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). The NRC staff reviewed the description of engineering assessment to support the categorization process and concludes it is consistent with NEI 00-04, Section 4, guidance as endorsed by RG 1.201 and, therefore, meets the requirements of 10 CFR 50.69(c)(1)(ii).

- 11 -

3.5 Component Safety-Significance Assessment (NEI 00-04, Section 5) Section 5 of NEI 00-04 provides guidance for performing component safety-significance assessments. Figure 5-1 of Section 5 provides a high-level summary of the process for use of risk analyses for SSC categorization. Using the system engineering functional characterization results as an input, the NEI 00-04 process includes assessing component risk significance separately for five hazard groups:

• internal events (including internal flooding) • fire events • seismic events • other external events (e.g., tornadoes, external floods, high winds) • shutdown events

Figures 5-2 and 5-7 of NEI 00-04, Section 5, provide detail on the process for assessment of safety significance for components for each of the hazard groups. The guidance in NEI 00-04, Section 5, includes criteria (e.g., risk importance measures for PRA approaches) and sensitivity analyses, as applicable, to determine if a candidate component is risk significant from the individual hazard perspective. Section 5.6 of NEI 00-04 describes the process for assessing overall risk significance using an integral assessment process. Section 5 of NEI 00-04 provides one approach for addressing 10 CFR 50.69(c)(1)(i) and (ii). LAR, Section 3.1, summarizes the use of quantitative or qualitative risk information in the categorization process from a PRA or other risk assessment methods. The LAR process uses approaches for the risk characterization, as follows:

internal events, internal flooding, and internal fires uses PRAs; SMA for seismic hazards; IPEEE screening to assess the risk from other external hazards (e.g., high winds,

external floods); and shutdown safety assessment to assess shutdown risk

The methods described in the LAR to assess internal and external hazards are consistent with the methods included in the NEI 00-04 guidance, as endorsed by RG 1.201 and, therefore, acceptable to the NRC staff. The guidance considers the results and insights from the plant-specific PRA peer reviews as required by 10 CFR 50.69(c)(1)(i), and non-PRA risk characterization as required by 10 CFR 50.69(c)(1)(ii). The application of these methods is reviewed in the following SE subsections: PRA in Subsections 3.5.1 and 3.5.2, and the non-PRA methods in Subsection 3.5.3. 3.5.1 Capability and Quality of the PRA to Support the Categorization Process

Paragraph 50.69(c)(1)(i) of 10 CFR requires, in part, that the PRA must be of sufficient quality and level of detail to support the categorization process and must be subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC.

- 12 -

The PINGP PRA is comprised of an internal events PRA that calculates CDF and LERF from internal events, including internal flooding, at full power and a fire PRA. Paragraph 50.69(b)(2)(iii) of 10 CFR requires that the results of the PRA review process conducted to meet 10 CFR 50.59(c)(1)(i) be submitted as part of the application. The LAR provided this information and the NRC staff’s review of this information is presented below. The LAR did not specify if flexible coping strategies (FLEX) equipment and strategies were included in the PRAs to support categorization. In a letter dated May 30, 2017, the NRC staff provided its assessment of the challenges to incorporating diverse and FLEX equipment into a PRA model in support of RI decisionmaking in accordance with the guidance of RG 1.200. The response to NRC request for information (RAI) 10 provided by letter dated April 29, 2019, stated that the PINGP PRAs do not currently credit any FLEX equipment or strategies; however, that once issues identified in the letter dated May 30, 2017, are resolved, FLEX equipment or strategies may be used during the categorization process in the future in accordance with the NRC-accepted resolution. Therefore, the NRC staff concludes that no further review of FLEX modelling strategies was necessary to complete the review of this RI application. Internal Events and Internal Flooding PRAs The NRC staff’s review of the internal events and internal flooding PRAs was based on:

the results of PRA peer reviews and facts and observations (F&O) closure reviews described in LAR, Sections 3.2.1, 3.3, Attachments 2 and 3; provided in letters dated April 29, 2019, and August 5, 2019, in response to NRC RAIs,

previously docketed information on PINGP PRA acceptability submitted in support of the “Issuance of Amendments Re: Transition to NFPA 805 ‘Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants’ which was issued by the NRC staff evaluation (SE) dated August 8, 2017 (Reference 10); and

previously docked information to support the on-going review of a LAR to modify the list of required NFPA 805 modifications at PINGP, dated May 18, 2018 (Reference 11), as supplemented.

The LAR stated that the last full-scope peer review of the internal events (excluding internal flooding) PRA was performed in November 2010, against the ASME/ANS 2009 PRA Standard, as endorsed by RG 1.200, Revision 2. Additional focused-scope peer reviews against the ASME/ANS 2009 PRA Standard, as endorsed by RG 1.200, Revision 2, were performed in September 2012 and May 2014, to evaluate the internal flooding portion of the PRA and the incorporation of the Flowserve N9000 reactor coolant pump (RCP) model (i.e., WCAP-16175-P-A, Reference 12), respectively. The LAR stated that an F&O closure review was performed by an independent assessment (IA) team on all internal events, internal flooding and fire finding-level F&Os in October 2017. The LAR clarified that the F&O closure review was performed as detailed in Appendix X to the guidance in NEI 05-04 (Reference 13), NEI 07-12 (Reference 14), and NEI 12-13 (Reference 15), concerning the process “Close-Out of Facts and Observations” as accepted by the NRC staff in letter dated May 3, 2017 (Reference 16). Attachment 3 of the LAR provides a description of remaining open F&Os with a resolution and disposition for potential impacts on the categorization process. The NRC staff reviewed these descriptions, resolutions, and application-specific dispositions of the F&Os on the 10 CFR 50.69

- 13 -

categorization. The NRC staff requested additional information to clarify the licensee’s disposition for one of these findings as described below. The disposition to F&O SY-A17-01 stated in the LAR that, “the N-9000 RCP seal model must obtain NRC review and approval.” The NRC staff noted that at the time it reviewed a LAR to adopt NFPA 805 at PINGP, the N-9000 RCP model was not approved and required an implementation item to ensure a final, NRC-approved PRA model was incorporated (Section 3.4.2.2 of the NRC staff’s NFPA 805 SE dated August 8, 2017). Therefore, RAI 8 in an email dated February 26, 2019 (Reference 17), the NRC staff requested clarification that the N-9000 RCP seal model was implemented in accordance with NRC approved WCAP-16175-P-A guidance. The response provided by letter dated April 29, 2019, confirmed that an NRC approved N-9000 RCP seal model was incorporated into the PRA model in accordance with all NRC conditions, limitations, and modifications detailed in the NRC SE to WCAP-16175-P-A. In the February 26, 2019, RAI 8 email, the NRC staff also noted that abeyance seals are sometimes included as a backup to the N-9000 RCP seal packages, therefore, the NRC requested the licensee clarify if an abeyance seal was credited in the PINGP PRA and if so to provide details of the methodology for NRC review. In a letter dated April 29, 2019, the licensee stated it will not credit an abeyance seal in its PRA models used for categorization until an NRC accepted abeyance seal model is available. The proposed license condition to adopt 10 CFR 50.69 at PINGP referenced an implementation item in Attachment 1 of the August 5, 2019, letter, that the abeyance RCP seal will only be credited in categorization after the NRC has approved the modeling approach for the abeyance seal. Therefore, the NRC staff finds that credit for the abeyance seal is appropriately considered. The NRC staff reviewed the information regarding the PINGP internal events PRA and concludes it is sufficiently technically acceptable (i.e., the PRA has an appropriate scope, level of detail, conformance with PRA elements, and plant representation) to support the categorization process in accordance with RG 1.200, RG 1.174, and NEI 00-04, as endorsed by RG 1.201 and, therefore, the PRA meets the requirements of 10 CFR 50.69(c)(1)(i). Fire PRA The NRC staff’s review of the PINGP fire PRA was based on:

the results of PRA peer reviews and F&O closure reviews described in LAR, Sections 3.2.1, 3.3, Attachments 2 and 3; information in letters dated April 29, 2019, and August 5, 2019, in response to NRC RAIs;

previously docketed information on PINGP fire PRA acceptability summarized in support of the “Issuance of Amendments Re: Transition to NFPA 805 ‘Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants’ in the SE dated August 8, 2017; and

previously docked information to support the on-going review of a request to modify the list of required NFPA 805 modifications at PINGP, dated May 18, 2018, as supplemented.

The LAR stated that the last full-scope peer review of the fire PRA was performed in May 2012, against the ASME/ANS 2009 PRA Standard, as endorsed by RG 1.200, Revision 2. Additional focused-scope peer reviews against the ASME/ANS 2009 PRA Standard, as endorsed by RG 1.200, Revision 2, were performed in November 2013, and December 2017, to evaluate

- 14 -

upgrades to the hot gas layer temperature correlations, the apportioning of the main control board fire frequency, and the time to cable damage from thermal sources (in accordance with NUREG/CR-6850, Reference 18), respectively. As stated above, an F&O closure review was performed by an IA team on all internal events, internal flooding and fire finding-level F&Os in October 2017. Attachment 3 of the LAR provides a description of remaining open F&Os with a resolution and disposition for potential impacts on the categorization process. The NRC staff reviewed these descriptions, resolutions, and application-specific dispositions of the F&Os on the 10 CFR 50.69 categorization. The NRC staff requested additional information to clarify the disposition for two of these findings as described below. Open F&O ES-C1-01 stated that the identification of instrumentation supporting operator actions in fire scenarios and modelling of instrument failures that could cause undesired operator action was incomplete. The IA closure team disposition to F&O ES-C1-01 “determined that the finding should remain open pending the inclusion of additional documentation for screening additional instrumentation for HFEs [human failure events].” Therefore, the IA closure team did not close the F&O. The response to RAI 2.i provided in the letter dated April 29, 2019, documented additional criteria for screening and stated that the criteria were developed by NSPM to meet the intent of NUREG/CR-6850, “EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, Volume 2: Detailed Methodology”, Section 2.5.5.2. Criteria for identifying instrument failures that could cause undesired actions include the ASME 2009 PRA standard surveillance requirements ES-C1-01 insofar as only single instrumentation failures that cause undesired actions must be modelled to meet CC II, and the NUREG/CR-6850 guidelines that “it would take considerable redundant and particularly diverse indications to spuriously operate or [to] fail simultaneously to induce the undesired/unsafe action.” The criteria provided in the response to RAI 2.i screen out instrumentation when multiple instrumentation failures are required to trigger the undesired actions, and, therefore the NRC staff finds the criteria are consistent with the guidelines and acceptable. Open F&O FSS-D7-01 pertains to calculations involving the non-suppression probability for the deluge system. The IA closure team determined the finding should remain open pending update of the unreliability used for the pre-action suppression system. The LAR resolution stated that in the new fire PRA, “[t]he non-suppression probability for the deluge systems [for structural steel fire scenarios] was revised to correct the identified errors.” The response to RAI 2.ii provided in the letter dated April 29, 2019, clarified that for these scenarios the detection and suppression unreliabilities are taken from NUREG/CR-6850. Non-suppression probabilities were extensively reviewed during the earlier NFPA 805 review (RAI 01.g response in letter dated June 19, 2015 (Reference 19), and Table 1 in RAI 3 in letter dated May 24, 2016 (Reference 20)). In the NFPA 805 SE dated August 8, 2017, the NRC staff concluded that, “the licensee demonstrated that the FPRA's treatment of the total unavailability of credited detection and suppression systems is consistent with accepted guidance in NUREG/CR-6850.” The NRC staff finds this open F&O is adequately dispositioned because generic values from NUREG/CR-6850 are acceptable to meet capability category (CC) II per the ASME PRA standard (and the default acceptable CC for RI applications) unless plant-specific outliers are identified, and the results of the NFPA 805 review indicate that, if outliers were identified, they would have been incorporated into the fire PRA.

- 15 -

The LAR stated:

Fire Risks: Fire PRA model Revision 5.3-APP1. The fire PRA model described within this LAR is the same as the one described within the NSPM submittal of the LAR to modify the list of required NFPA 805 modifications.

The “NSPM submittal” referenced in the quote above is the May 18, 2018, LAR to revise the license condition associated with implementation of NFPA 805 as supplemented on July 10, 2018. The NFPA 805 LAR requested to delete several modifications required for full implementation of NFPA 805. The NRC staff approved the NFPA 805 LAR by letter dated July 30, 2019 (ADAMS Accession No. ML19140A447). RAI 1 provided by e-mail dated February 26, 2019, noted that plant modifications required to implement NFPA 805 and the associated PRA models may be changing based on the on-going review. The NRC staff requested clarification about what version of the fire PRA will be used for categorization and assurance that the fire PRA used for categorization will appropriately model the as-built and as-operated plant. The response to RAI 1 proposed adding the following license condition for implementing 50.69:

NSPM shall ensure that the fire PRA model used for the 10 CFR 50.69 SSC categorization reflects the as-built, as-operated plant using the same fire PRA model used to support NFPA 805 implementation for both PINGP units prior to implementation of 10 CFR 50.69.

Based on the license condition that the final NFPA 805 fire PRA will be used for categorization, the NRC staff finds that approval of the request to implement 50.69 can proceed independent of the request to implementing NFPA 805. The NRC staff reviewed the information regarding the PINGP fire PRA and concludes it is sufficiently technically acceptable (i.e., the PRA has an appropriate scope, level of detail, conformance with PRA elements, and plant representation) to support the categorization process in accordance with RG 1.200, RG 1.174, and NEI 00-04, as endorsed by RG 1.201 and, therefore, the PRA meets the requirements of 10 CFR 50.69(c)(1)(i). 3.5.2 Importance Measures and Sensitivity Studies Paragraph 50.69(c)(1)(i) of 10 CFR requires that the results and insights from the PRA be used during categorization. These requirements are met, in part, by using importance measures and sensitivity studies, as described in the methodology in NEI 00-04, Section 5.0. Fussell-Vesely and Risk Achievement Worth importance measures are obtained for each component and each PRA modeled hazard (i.e., separately for the internal events PRA and for the fire PRA) and the values are compared to specified criteria. Components that have internal event importance measure values exceeding the criteria are assigned HSS. Components that have fire event importance measures exceeding the criteria are assigned preliminary HSS. Integrated importance measures over all PRA modeled hazards are calculated per Section 5.6 of NEI 00-04, and components for which these measures exceed the criteria are assigned preliminary HSS. The guidance in NEI 00-04 specifies sensitivity studies to be conducted for each PRA model. The sensitivity studies are performed to ensure that assumptions associated with these specific uncertain parameters (i.e., human error, common cause failure, and maintenance probabilities)

- 16 -

are not masking the importance of a component. The NEI 00-04 guidance states that any additional “applicable sensitivity studies” from characterization of PRA adequacy should be considered. Section 3.2.7 of the LAR summarizes how the licensee searched for additional issues in the internal events (including internal flooding) and fire PRAs that should be evaluated with a sensitivity study. Section 3.7 states that the licensee used the NRC guidance in NUREG-1855, “Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making,” (Reference 21) supplemented with the Electric Power Research Institute (EPRI) topical report (TR)-1016737, “Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments,” (Reference 22)to identify sources of uncertainty in the internal events PRA. Attachment 6 of the LAR provided a list of key assumptions and sources of uncertainty resulting from its review of the internal events (including internal flooding) and fire PRAs and provided dispositions for each entry. Section 3.2.7 of the LAR states that Attachment 6 included, “only those assumptions or sources of uncertainty that could significantly impact the risk calculations were considered key for this application.” In RAI 6, dated February 26, 2019, the NRC noted that a revision to NUREG-1855 was available (NUREG 1855, Revision 1, Reference 23), which references updated EPRI guidance, TR-1026511 (Reference 24), in addition to the sources of uncertainty identified in EPRI TR-1016737. The RAI requested additional information regarding the process used to identify the key assumptions and sources of uncertainty including whether the process was consistent with NUREG-1855, Revision 1. In addition, in RAI 7, dated February 26, 2019, the NRC staff noted that many of the entries in Attachment 6 are dispositioned by sensitivity studies that will be performed “as needed” or treated on a “case-by-case basis.” The NRC staff requested further explanation of these dispositions in RAI 7. The response to RAI 6 provided in a letter dated April 29, 2019, as supplemented by letter dated August 5, 2019, stated that PRA uncertainty evaluation was based on all identified plant specific sources of uncertainty that were compiled and characterized in its PRA uncertainty notebooks. The uncertainty notebooks were developed by reviewing all modeling assumptions from the PINGP PRA using an approach that is consistent with the ASME/ANS RA-Sa-2009. The LAR stated that generic sources of uncertainty discussed in EPRI TR-1016737 were included in the original uncertainty evaluation. The RAI response stated the plant-specific fire PRA sources of uncertainty based on the most recent fire model evaluation have been updated with the generic sources of fire PRA uncertainty discussed in EPRI TR-1026511. The response to RAI 7 clarified that this final uncertainty evaluation meets the intent of NUREG-1855, Revision 1. The response to RAI 7 stated that additional evaluation was performed for the uncertainties and dispositions originally proposed in Attachment 6 of the LAR. The August 5, 2019, RAI response described additional evaluations on key sources of uncertainty and provided an updated Attachment 6 that replaced Attachment 6 in the LAR. The response stated that several uncertainties related to unmodelled SSCs in the residual heat removal and main feedwater systems could be excluded using screening criteria in the ASME PRA Standard. The PRA standard allows screening when modelled SSC failures are much more likely than related, but unmodelled, SSCs failure. Sensitivity studies were performed for the uncertainties associated with failure to trip circulating water pumps and with modelling of interfacing system loss-of-coolant accidents which concluded that no events would transition from LSS to HSS for bounding changes in the model parameters. The thermal induced steam generator tube rupture model was identified as consistent with accepted industry methods and therefore is not a key source of uncertainty. A single remaining model uncertainty associated with human reliability

- 17 -

analyses was included in the updated Attachment 6 where it was noted that the standard sensitivity studies on human reliability values would account for this uncertainty. The additional evaluations described in the response to RAI 7 to remove issues from Attachment 6 of the LAR are consistent with general guidelines in NUREG-1855 that consensus methods, assumptions that do not impact the decisions, and issues that can be screened out based on minimal impact are not key assumptions or key sources of uncertainty. The NRC staff concludes that the evaluation of key assumptions and sources of uncertainty was done consistent with NUREG-1855 and the two supporting EPRI documents based on the licensee’s statements that the guidance was followed, and the evaluations described in the submittal as supplemented. The NRC staff reviewed the information regarding use of importance measures and the approach to identify, evaluate, and disposition key assumptions and sources of uncertainty and concludes that, the use and approach meets the requirements of 10 CFR 50.69(c)(1)(i) and (ii) because the use and approach are consistent with NEI 00-04, Section 5, and NUREG-1855, Revision 1. 3.5.3 Non-PRA Methods According to 10 CFR 50.69(c)(1)(ii), SSC functional importance must use an integrated, systematic process for addressing initiating events, SSCs, and plant operating modes, including those not modeled in the plant-specific PRA. The functions to be identified and considered include design-bases functions and functions credited for mitigation and prevention of severe accidents. As described in the LAR, the categorization process uses the following non-PRA methods:

SMA results to address seismic risk; screening during the IPEEE to address risk from other external hazards; shutdown safety plan as described in NUMARC 91-06 to address shutdown risk.

These approaches are discussed below. Seismic Risk Section 5.3 of NEI 00-04 provides guidance regarding seismic assessment using a seismic PRA or the results of a SMA. The LAR proposed to use the results of its SMA performed for the IPEEE in response to Generic Letter 88-20, Supplement 426 to assess seismic risk for the 10 CFR 50.69 categorization process. The SMA is a screening approach that does not quantify CDF or LERF. The SMA includes the development of the SSEL, which contains the components that would be needed during and after a seismic event. The SSEL identifies one preferred and one alternate path capable of achieving and maintaining safe shutdown conditions for at least 72 hours following an earthquake. The LAR stated that the IPEEE SSEL had been updated to reflect the current as-built and as-operated plant and that future plant changes would be evaluated to determine their impact on the SMA and risk categorization process. Consistent with NEI 00-04, the LAR stated that all items on its SSEL list will be assigned as HSS based on seismic risk. The method proposed meets 10 CFR 50.69(c)(1)(ii) because all

- 18 -

items developed during its systematic IPEEE process and included in the resulting SSEL will be categorized as HSS based on seismic risk as described in the NRC-endorsed NEI 00-04. Therefore, the NRC staff finds the proposed method acceptable. Other External Hazards As indicated in the LAR, external hazards were initially evaluated during the IPEEE. This hazard category includes all non-seismic external hazards such as high winds, external floods, transportation and nearby facility accidents, and other hazards. The IPEEE analysis used a progressive screening approach and concluded that all these other hazards are negligible contributors to overall plant risk. In addition, the LAR stated that these other external hazards had been reevaluated using the criteria in the ASME/ANS 2009 Standard as summarized in Attachments 4 and 5 of the LAR. Section 3.2.4 of the LAR states that an evaluation is performed to determine if there are SSCs being categorized that participate in screened scenarios and whose failure would result in an unscreened scenario. Section 3.2.4 of the LAR concludes that all remaining hazards (i.e., excluding seismic) were screened from applicability and considered insignificant for every SSC and, therefore, will not be considered during the categorization process. The conclusion of LAR, Section 3.2.4, stated that the external hazard assessment has been completed (based on the IPEEE) and that all external hazards will never need evaluation during categorization. The response to RAI 3 and RAI 4 clarified that some credible external hazards could not be screened out as always insignificant contributors if SSCs credited for protection or mitigation of the hazard failed. The response provided the following list of hazards that will be reviewed during the categorization to determine if there are components being categorized that participate in screened scenarios and whose failure would result in an unscreened scenario.

• external flooding and local intense precipitation • extreme wind or tornado • lightning • low river, drought, and river diversion • low winter temperature, snow, and ice cover • transportation and pipeline accidents • toxic gases • forest or range fire • sand or dust storm • biological events

The responses to RAI 3 and RAI 4 clarified that, per NEI 00-04, the external hazard assessment is required for each SSC when the SSC is categorized. Therefore, each SSC being categorized will be assessed in accordance with NEI 00-04 Figure 5-6 for the identified external hazards. If the failure of the SSC results in the screening criterion from Attachment 5 (of the LAR) not being met, then the scenario would become unscreened and the SSC would become candidate HSS. The NRC staff finds this clarification acceptable because it is consistent with the guidance of Section 5.4 of NEI 00-04.

- 19 -

Shutdown Risk Consistent with the NEI 00-04 guidance endorsed by the NRC, the LAR proposes to use the shutdown safety assessment process based on NUMARC 91-06. 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. The guidance in NUMARC 91-06 specifies that a DID approach should be used with respect to each defined shutdown key safety function. The response to RAI 5 provided in the letter dated April 29, 2019, and consistent with the guidance in NEI 00-04, Section 5.5, clarified that “SSCs that meet one of-two criteria (i.e., considered part of a ‘primary shutdown safety system’ or whose failure would initiate an event during shutdown conditions) described in Section 5.5 of NEI 00-04 will be considered preliminary HSS.” The proposed shutdown safety assessment method is consistent with the NRC endorsed NEI 00-04 and, therefore, the NRC staff finds the proposed method acceptable. 3.5.4 Component Safety-Significance Assessment for Passive Components Passive components are not modeled in the PRA and, 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 function of the body of a motor-operated valve. The LAR proposed using a categorization method for passive components not cited in NEI 00-04 for passive component categorization but approved by the NRC for ANO-2 (Reference 26). The ANO-2 methodology is a RI safety classification and treatment program for repair/replacement activities for Class 2 and Class 3 pressure-retaining items and their associated supports, 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 27). 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 proposed repair/replacement methodology is acceptable and appropriate for passive component categorization of Class 2 and Class 3 SSCs. Because all Class 1 SSCs and supports will be considered HSS, and only Class 2 and Class 3 SSCs will be categorized using the ANO-2 passive categorization methodology consistent with previous NRC staff approval, the NRC staff finds the proposed approach for passive categorization acceptable for the 10 CFR 50.69 categorization process. 3.5.5 Summary The NRC staff reviewed the PRA and the non-PRA methods described in the LAR for use in the 10 CFR 50.69 categorization process to assess the safety-significance of active and passive components and finds these methods acceptable and consistent with RG 1.201 and the

- 20 -

NRC-endorsed guidance in NEI 00-04. The NRC staff approves the use of the following methods in the proposed 10 CFR 50.69 categorization process:

PRA to address internal events, including internal flooding risk fire PRA to address fire risk SMA results in the SSEL to address seismic risk screening using IPEEE models address the potential for credited barriers to be HSS

from other credible external hazards (e.g., high winds, external floods) shutdown safety management plan to assess shutdown risk ANO-2 passive categorization method to assess passive component risk for Class 2,

Class 3, and non-Code class SSCs and their associated supports Based on its review of the LAR, the staff recognized that additional information was necessary to support its conclusion that the proposed program meets the requirements in 10 CFR 50.69 and the guidance in RG 1.201 and NEI 00-04. The response to RAI 1 provided in the letter dated August 5, 2019, proposed the addition of the following license conditions for the implementation of 10 CFR 50.69.

NSPM will complete the implementation items listed in Attachment 1 of NSPM’s letter to the NRC dated August 5, 2019, prior to implementation of 10 CFR 50.69. NSPM shall ensure that the fire PRA model used for the 10 CFR 50.69 SSC categorization reflects the as-built, as-operated plant using the same fire PRA model used to support NFPA 805 implementation for both PINGP units prior to implementation of 10 CFR 50.69.

There is one implementation item provided in the August 5, 2019, RAI response that would be required per the condition above.

The NSPM PRA model used for categorization will only credit the abeyance RCP seal after the NRC has accepted the methodology to model the abeyance seal in PRAs. Prior to methodology acceptance, the PRA model used for categorization will not credit the abeyance seal.

The response to RAI 8 states that the abeyance RCP seal would be credited if an NRC accepted method is issued. Therefore, the proposed implementation item license condition states that the seal will only be credited when the NRC accepts a methodology and will not be credited before an accepted method. The license condition that the fire PRA model will be the same fire PRA model used to support NFPA 805 implementation, allows the staff to accept all the results and conditions of the previous and on-going NFPA associated reviews of the fire PRA without additional review or detailed discussions. The NRC has higher PRA technical adequacy guidelines that it applies to the NFPA 805 fire PRA and, therefore, can accept the same fire PRA for use to support 50.69 categorization.

- 21 -

3.6 Defense-in-Depth (NEI 00-04, Section 6) NEI 00-04, Section 6, provides guidance on assessment of DID. Section 3.1.1 of the LAR states that categorization of SSCs will be completed per the NEI 00-04 process as endorsed by RG 1.201, including the guidance on DID assessment. Figure 6-1 in NEI 00-04 provides guidance to assess design-basis DID based on the likelihood of the design-basis internal event initiating event and the number of redundant and diverse trains nominally available to mitigate the initiating event. The likelihood of the initiating events is binned, for different likelihoods, and HSS is assigned if fewer than the indicated number of mitigating trains are nominally available. Section 6 of NEI 00-04 also provides guidance to assess containment DID based on preserving containment isolation and long-term containment integrity and on preventing containment bypass and early hydrogen burns. The DID for beyond design-basis initiating events is addressed by the PRA categorization process. RG 1.201 endorses the guidance in NEI 00-04, Section 6, but notes that the containment isolation criteria in this section of NEI 00-04 are separate and distinct from those set forth in 10 CFR 50.69(b)(1)(x). The criteria in 10 CFR 50.69(b)(1)(x) are to be used in determining which containment penetrations and valves may be exempted from the Type B and Type C leakage testing requirements in both Options A and B of Appendix J to 10 CFR Part 50, but the 10 CFR 50.69(b)(1)(x) criteria are not used to determine the proper RISC category for containment isolation valves or penetrations. Based on the licensee’s statements that it will follow the endorsed guidance without identifying any deviations or exemptions, the NRC staff finds that the licensee’s categorization process is consistent with the NRC-endorsed NEI 00-04 guidance and fulfills the 10 CFR 50.69(c)(1)(iii) criterion that DID is maintained. 3.7 Preliminary Engineering Categorization of Functions (NEI 00-04, Section 7) All the information collected and evaluated in the different engineering evaluations is collected, organized, and provided to the IDP, as described in NEI 00-04, Section 7. The IDP will make the final decision about the safety-significance of SSCs based on guidelines in NEI 00-04, the information they receive and their expertise. LAR, Section 3.1.1, as summarized in Table 3.1 (repeated as Table 1 in this SE), stated that if any component is identified as HSS from either the integrated risk component safety-significance assessment (Section 5 of NEI 00-04), the DID assessment (Section 6 of NEI 00-04), or the Qualitative Criteria (Section 9 of NEI 00-04), the associated system function(s) would be identified as HSS. Once a system function is identified as HSS, then all the components that support that function are categorized as preliminary HSS. These guidelines, and the other associated guidelines, are consistent with the guidance in NEI 00-04, as endorsed by RG 1.201 and, therefore, acceptable. 3.8 Risk Sensitivity Study (NEI 00-04, Section 8) Paragraph 50.69(c)(1)(iv) of 10 CFR requires that any potential increases in CDF and LERF resulting from changes to treatment be small. The guidance in Section 8 of NEI 00-04, as endorsed by RG 1.201, includes an overall risk sensitivity study for all the LSS components to confirm that if the unreliability of the components were increased, the increase in risk would be small (i.e., meet the acceptance guidelines of RG 1.174). Section 3.1.1 and Section 3.2.7 of the

- 22 -

LAR clarify that in the sensitivity study, the unreliability of all LSS SSCs modeled in the PRA(s) will be increased by a factor of 3 which is within the range of 3 to 5 suggested by NEI 00-04. Separate sensitivity studies are to be performed for each system categorized, as well as a cumulative sensitivity study for all the SSCs categorized through the 10 CFR 50.69 process. This sensitivity study, together with the periodic review process discussed in Section 3.10 of this SE, assure that the potential cumulative risk increase from the categorization is small. 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 proposed risk sensitivity study will be performed consistent with the guidance in NEI 00-04, Section 8 and, therefore, will assure that the potential cumulative risk increase from the categorization is small, as required by 10 CFR 50.69(c)(1)(iv). 3.9 Integrated Decision-making Panel Review and Approval (NEI 00-04, Sections 9 and 10) Section 50.69(c)(2) of 10 CFR requires that the SSCs must be categorized by an IDP staffed with expert, plant-knowledgeable, members whose expertise includes, at a minimum, PRA, safety analysis, plant operations, design engineering, and system engineering. LAR, Section 3.1.1, 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 required expertise will be found in the IDP. The guidance in NEI 00-04, endorsed in RG 1.201, ensures 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). As provided by the NEI 00-04 guidance, and as indicated in LAR, Attachment 1, the process used by the IDP for the categorization of SSCs will be described and documented in a plant procedure. Section 3.1.1 of the LAR states that at least three members of the IDP will have a minimum of 5 years of experience at the plant, and there will be at least one member of the IDP who has a minimum of 3 years of experience in modeling and updating of the plant-specific PRA. In addition, Section 3.1.1 states that the IDP will be trained in the specific technical aspects and requirements related to the categorization process. Training will address, at a minimum, the purpose of the categorization; present treatment requirements for SSCs, including requirements for design-basis events; PRA fundamentals; details of the plant-specific PRA, including the modeling, scope, and assumptions; the interpretation of risk importance measures, and the role of sensitivity studies and the change-in-risk evaluations; and the DID philosophy and requirements to maintain this philosophy. Based on its review, the NRC staff finds that the proposed 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, as endorsed by RG 1.201. Therefore, all aspects of the integrated, systematic process used to characterize SSCs will reasonably reflect current plant configuration and operating practices and applicable plant and industry operational experience as required by 10 CFR 50.69(c)(1)(ii). Paragraph 50.69(c)(1)(iv) of 10 CFR requires reasonable confidence that sufficient safety margins are maintained for SSCs categorized as RISC-3. The licensee addresses safety

- 23 -

margins through an integrated engineering evaluation that would nominally be addressed by the IDP. Consistent with the discussion in the NEI 00-04 guidance endorsed by RG 1.201, the IDP need not explicitly consider safety margins. Sufficient safety margin will be maintained because the RISC-3 SSCs will remain capable of performing their safety-related functions as required by 10 CFR 50.69(d)(2), and because any potential increases in CDF and LERF that might stem from changes in RISC-3 SSC reliability due to reduced treatment permitted by 10 CFR 50.69 will be maintained small, as required by 10 CFR 50.69(c)(1)(iv). Therefore, the NRC staff finds that the program implemented by the licensee, consistent with the endorsed guidance in NEI 00-04, fulfills the 10 CFR 50.69(c)(1)(iv) criteria that sufficient safety margins are maintained. 3.10 Program Documentation, Change Control, and Periodic Review (NEI 00-04, Sections 11

and 12) Paragraph 50.69(c)(1)(ii) of 10 CFR requires, in part, that all aspects of the integrated, systematic process used to characterize SSC importance must reasonably reflect the current plant configuration and operating practices and applicable plant and industry operating experience. Section 11 of NEI 00-04, as endorsed in RG 1.201, provides guidance on program documentation and change control, and Section 12 provides guidance on periodic review. These sections are described in NEI 00-04 with respect to satisfying 10 CFR 50.69(e) and 10 CFR 50.69(f), respectively. Maintaining change control and periodic review will also maintain confidence that all aspects of the program reflect current plant operation. Section 50.69(e) of 10 CFR requires periodic updates to the PRA and SSC categorization. The NRC staff finds that changes over time to the PRA and SSC reliabilities are inevitable, and such changes are recognized by the 10 CFR 50.69(e) provision requiring periodic updates. As provided in RG 1.200, the NRC staff review of the PRA quality and level of detail reported in this SE is based primarily on determining how the licensee has resolved key assumptions and areas identified by peer reviewers as being of concern (i.e., F&Os). As discussed above in Section 3.5 of this SE, the NRC staff has concluded that any weaknesses or errors in the PRA related to SSC categorizations have been addressed. A single implementation item restricts the use of an, as yet, unaccepted method (i.e., credit for abeyance RCP seals) until that method has been accepted by the NRC staff. As described in LAR, Section 3.2.6, administrative controls are in place to ensure that the PRA models used to support the categorization reflect the as-built, as-operated plant over time. The process includes regularly scheduled PRA updates supported by interim, as needed, adjustments to the categorization or treatment processes to maintain the validity of the processes. The 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), for assessing the risk impact of unincorporated changes, and for controlling the model and associated computer files. The NRC staff finds that this description is consistent with the requirements for feedback and process adjustment of 10 CFR 50.69(e), and is, therefore, acceptable. Section 50.69(f) of 10 CFR requires program documentation, change control, and records. LAR, Section 3.2.6, stated that a process will be implemented that addresses the guidance in Section 11 of NEI 00-04 pertaining to program documentation and change control records.

- 24 -

Section 3.1.1 of the LAR stated that the RISC categorization process documentation will include the following 10 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

In addition, LAR, Attachment 1 (List of Categorization Prerequisites), stated that procedures will be established prior to the use of the categorization process that will contain the following elements: (1) IDP member qualification requirements, (2) qualitative assessment of system functions, (3) component safety-significance assessment, (4) assessment of DID and safety margin, (5) review by the IDP and final determination of safety-significance for system functions and components, (6) risk sensitivity studies to confirm that the risk acceptance guidelines of RG 1.174 are met, (7) periodic review to ensure continued categorization validity and acceptable performance for SSCs that have been categorized, and (8) documentation requirements identified in LAR Section 3.1.1. Procedures are formal plant documents, and changes will be tracked providing change control and records of the changes. The NRC staff finds that the change control and performance monitoring of categorized SSCs and PRA updates, as described in the LAR, as supplemented, will sufficiently capture and evaluate component failures to identify significant changes in the failure probabilities. In addition, the PRA update program and associated re-evaluation of component importance will appropriately consider the effects of changing failure probabilities and changing plant configuration on the component safety-significant categories. As discussed above, the NRC staff finds that the adherence to the process in NEI 00-04, Sections 11 and 12, as described in the LAR, will meet the requirements of 10 CFR 50.69(f) and 10 CFR 50.69(e), respectively. Therefore, the process used to characterize SSC importance will reasonably reflect the current plant configuration and operating practices, and applicable plant and industry operational experience required in 10 CFR 50.69(c)(1)(ii). 3.11 Technical Conclusion The NRC staff reviewed the proposed 10 CFR 50.69 categorization process provided in the LAR and concludes that it adequately implements 10 CFR 50.69 using models, methods, and approaches consistent with NEI 00-04 and RG 1.201, and, therefore, satisfies the requirements of 10 CFR 50.69(c). Based on its review, the NRC staff finds the proposed categorization process acceptable for categorizing the safety significance of SSCs. Specifically, the NRC staff concludes that the proposed categorization process:

considers results and insights from plant-specific internal events and fire PRAs that, with the restriction provided by the implementation item (summarized in Section 3.5.5 of this SE), will be of sufficient quality and level of detail to support the categorization process and that have been subjected to a peer review process against RG 1.200,

- 25 -

Revision 2, as reviewed in Section 3.5.1 of this SE and, therefore, meets the requirements in 10 CFR 50.69(c)(1)(i);

determines SSC functional importance using an integrated systematic process that reasonably reflects the current plant configuration, operating practices, and applicable plant and industry operational experience, as reviewed in Sections 3.3, 3.4, 3.5, 3.7, and 3.10 of this SE and, therefore, meets the requirements in 10 CFR 50.69(c)(1)(ii);

maintains DID, as reviewed in Section 3.6 of this SE and, therefore, meets the

requirements in 10 CFR 50.69(c)(1)(iii);

includes evaluations that provide reasonable confidence that for SSCs categorized as RISC-3, sufficient safety margins are maintained and that any potential increases in CDF and LERF resulting from changes in treatment are small, as reviewed in Section 3.8 and 3.9 of this SE and, therefore, meets the requirements in 10 CFR 50.69(c)(1)(iv);

is performed for entire systems and structures, rather than for selected components

within a system or structure, as reviewed in Section 3.3 of this SE and, therefore, the requirements in 10 CFR 50.69(c)(1)(v) will be met upon implementation; and

includes categorization by IDP, staffed with expert, plant-knowledgeable members whose expertise includes, at a minimum, PRA, safety analysis, plant operation, design engineering and system engineering, as reviewed in Section 3.9 of this SE and, therefore, meets the requirements in 10 CFR 50.69(c)(2).

3.12 10 CFR 50.69 Implementation License Condition Section 50.69(b)(2) of 10 CFR requires the licensee to submit an application that describes the categorization process. Section 50.69(b)(3) of 10 CFR states that the Commission will approve the license application if it determines that the categorization process satisfies the requirements of 10 CFR 50.69(c). As described in this SE, the NRC staff has concluded that the 10 CFR 50.69 categorization process described in the licensee’s application, as supplemented, includes a description of the categorization process that satisfies the requirements of 10 CFR 50.69(c). The response to RAI 1 provided in letter dated August 5, 2019, proposed the following license condition:

NSPM is approved to implement 10 CFR 50.69 using the approaches 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, with 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 SSCs and their associated supports; and the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards (e.g., external flooding and high winds)

- 26 -

updated using the external hazard screening significance criteria identified in ASME/ANS PRA Standard RA-Sa-2009, as endorsed in RG 1.200, Revision 2; as specified in PINGP License Amendment No. [XXX] dated [DATE]. Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization approach specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach). NSPM will complete the implementation items listed in Attachment 1 of NSPM’s letter to the NRC dated August 5, 2019, prior to implementation of 10 CFR 50.69.

NSPM shall ensure that the fire PRA model used for the 10 CFR 50.69 SSC categorization reflects the as-built, as-operated plant using the same fire PRA model used to support NFPA 805 implementation for both PINGP units prior to implementation of 10 CFR 50.69.

Based on its evaluation in this SE, the NRC staff finds that the proposed license condition and the referenced implementation item are acceptable because they adequately implement 10 CFR 50.69 using models, methods, and approaches consistent with the applicable guidance that has previously been endorsed as acceptable by the NRC. The NRC staff, through an onsite audit or during future inspections, may choose to examine the closure of the implementation items with the expectation that any variations discovered during this review, or concerns regarding adequate compliance with the implementation item, would be tracked and dispositioned appropriately under the licensee’s corrective action program, and could be subject to appropriate NRC enforcement action. 4.0 STATE CONSULTATION In accordance with the Commission's regulations, the Minnesota State official was notified of the proposed issuance of the amendments on September 26, 2019. The State official had no comments. 5.0 ENVIRONMENTAL CONSIDERATION The amendments change the requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration and there has been no public comment on such finding published on September 11, 2018 (83 FR 45986). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

- 27 -

6.0 CONCLUSION The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public. 7.0 REFERENCES 1. Letter (L-PI-18-012) from NSPM to the NRC, “Application to Adopt 10 CFR 50.69, ‘Risk Informed Categorization and Treatment of Structures, Systems and Components for

Nuclear Power Reactors’”, dated July 20, 2018 (ADAMS Accession No. ML18204A393). 2. Letter (L-PI-19-014) from NSPM to the NRC, “Response to Request for Additional

Information: Application to Adopt 10 CFR 50.69, “Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors” (EPID L-2018-LLA-0196),” dated April 29, 2019 (ADAMS Accession No. ML19119A216).

3. Letter (L-PI-19-029) from NSPM to the NRC, “Supplement to Response to Request for

Additional Information: Application to Adopt 10 CFR 50.69, “Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors” (EPID L-2018-LLA-0196),” dated August 5, 2019 (ADAMS Accession No. ML19217A333).

4. Nuclear Energy Institute, “10 CFR 50.69 SSC Categorization Guideline,” NEI 00-04,

Revision 0, dated July 2005 (ADAMS Accession No. ML052900163). 5. U.S. Nuclear Regulatory Commission, “Guidelines for Categorizing Structures, Systems,

and Components in Nuclear Power Plants According to Their Safety Significance,” Regulatory Guide 1.201 (For Trial Use), Revision 1, dated May 2006 (ADAMS Accession No. ML061090627).

6. Nuclear Management and Resources Council, “Guidelines for Industry Actions to Assess

Shutdown Management,” NUMARC 91-06, dated December 1991 (ADAMS Accession No. ML14365A203).

7. U.S. Nuclear Regulatory Commission, “An Approach for Determining the Technical

Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities,” Regulatory Guide 1.200, Revision 2, dated March 2009 (ADAMS Accession No. ML090410014).

8. American Society of Mechanical Engineers/American Nuclear Society, “Standard for Level

1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications,” ASME/ANS RA Sa 2009, dated February 2009 (ADAMS Accession No. ML092870592).

9. U.S. Nuclear Regulatory Commission, “An Approach for Using Probabilistic Risk

Assessment in Risk Informed Decisions on Plant Specific Changes to the Licensing Basis,” Regulatory Guide 1.174, Revision 3, dated January 2018 (ADAMS Accession No. ML17317A256).

- 28 -

10. Kuntz, Robert F., U.S. Nuclear Regulatory Commission, letter to Northard, Scott D.,

Northern States Power Company, “Prairie Island Nuclear Generating Plant, Units 1 and 2 – Issuance of Amendment RE: Transition to NFPA 805 “Performance-based Standard for Fire Protection for Light Water Reactor Electric Generating Plants” (CAC Nos. ME9734 and ME9735), dated August 8, 2017 (ADAMS Accession No. ML17163A027).

11. NSPM letter to NRC, “License Amendment Request to Revise License Condition

Associated with Implementation of NFPA 805”, dated May 18, 2018 (ADAMS Accession No. ML18138A402).

12. Nuclear Regulatory Commission SE re: Westinghouse Electric Company

WCAP-16175-NP-A, “Model for Failure of RCP Seals Given Loss of Seal Cooling in CE NSSS Plants”, Revision 0, dated March 2007 (ADAMS Accession No. ML071130383).

13. NEI 05-04, “Process for Performing Follow on PRA Peer Reviews Using the ASME/ANS

PRA Standard,” Revision 2 dated November 2008 (ADAMS Accession No. ML083430462).

14. Nuclear Energy Institute, “Fire Probabilistic Risk Assessment (FPRA) Peer Review

Process Guidelines,” NEI 07-12, Revision 1, dated June 2010 (ADAMS Package Accession No. ML102230049).

15. Nuclear Energy Institute, “External Hazards PRA Peer Review Process Guidelines,”

NEI 12-13, dated August 2012 (ADAMS Package Accession No. ML122400044). 16. Giitter, Joseph, and Ross Lee, Mary Jane, U.S. Nuclear Regulatory Commission, letter to

Greg Krueger, Nuclear Energy Institute, “U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 07-12, and 12-13, Close Out of Facts and Observations (F&Os),” dated May 3, 2017 (ADAMS Accession No. ML17079A427).

17. Email from the NRC to NSPM, “Request for Additional Information RE: Prairie Island 50.69

Amendment Request”, dated February 26, 2019 (ADAMS Accession No. ML19057A165). 18. NUREG/CR-6850 Volume 2, “EPRI/NRC-RES Fire PRA Methodology for Nuclear Power

Facilities, Volume 2: Detailed Methodology”, dated September30, 2005 (ADAMS Accession No. ML15167A411).

19. Letter (L-PI-15-052) from NSPM to the NRC “License Amendment Request to Adopt

NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactors - Response to Request for Additional Information - 90-Day Responses (TAC Nos. ME9734 and ME9735),” dated June 19, 2015 (ADAMS Accession No ML15174A139).

20. Letter (L-PI-16-005) from NSPM to the NRC “Prairie Island, Units 1 and 2 - License

Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactors - Response to Request for Additional Information (TAC Nos. ME9734 and ME9735),” dated May 24, 2016 (ADAMS Accession No. ML16152A046).

- 29 -

21. NUREG 1855, “Guidance on the Treatment of Uncertainties Associated with PRAs in Risk Informed Decisionmaking,” Volume 1, March 2009 (ADAMS Accession No. ML090970525).

22. Electric Power Research Institute, “Treatment of Parameter and Modeling Uncertainty for

Probabilistic Risk Assessments,” EPRI TR-1016737, dated December 2008. 23. NUREG 1855, “Guidance on the Treatment of Uncertainties Associated with PRAs in Risk

Informed Decisionmaking,” March 2017 (ADAMS Accession No. ML17062A466). 24. Electric Power Research Institute, “Practical Guidance on the Use of Probabilistic Risk

Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty,” EPRI TR-1026511, dated December 2012.

25. Markley, Michael, U.S. Nuclear Regulatory Commission, letter to Vice President,

Operation, Arkansas Nuclear One, Entergy Operations, Inc., “Arkansas Nuclear One, Unit 2 - Approval of Request for Alternative ANO-2 R&R-004, Revision 1, Request to Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 & 3 Moderate and High Energy Systems,” dated April 22, 2009 (ADAMS Accession No. ML090930246).

26. American Society of Mechanical Engineers, “Risk-Informed Safety Classification for Use in

Risk-Informed Repair/Replacement Activities,” ASME Code Case, N-660, dated July 2002. Principal Contributors: Candace de Messieres, NRR Date of issuance: November 12, 2019

S. Sharp - 2 -

SUBJECT: PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: - ISSUANCE OF AMENDMENTS RE: ADOPTION OF 10 CFR 50.69 – “RISK-INFORMED CATERIZATION AND TREATMENT OF STRUCTURE, SYSTEMS AND COMPONENTS OF NUCLEAR POWER REACTORS” (EPID L-2018-LLA-0196) DATED NOVEMBER 12, 2019

DISTRIBUTION: PUBLIC RidsNrrDorlLpl3 Resource RidsNrrPMPrairieIsland Resource RidsNrrLASRohrer Resource RidsOgcRp Resource RidsACRS_MailCTR Resource RidsNrrDraAplb Resource RidsRgn3MailCenter Resource RidsNrrDorlDpr Resource ADAMS Accession No.: ML19276F684 *memo dated **via e-mail

OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DE/EEOB/BC(A) NRR/DSS/SNSB/BC(A)

NAME RKuntz SRohrer DWilliams** RBeaton**

DATE 10/8/19 10/8/19 10/10/19 10/18/19

OFFICE NRR/DMLR/MPHB/BC(A) NRR/DMLR/MVIB/BC NRR/DEX/EICA/BC(A) NRR/DRA/APLA/BC*

NAME ABuford** DAlley** WMorton** SRosenberg (MLevine for)

DATE 10/15/19 10/10/19 10/18/19 8/30/19

OFFICE OGC NLO with comment DORL/LPL3/BC DORL/LPL3/PM

NAME KGamin** NSalgado (SWall for) RKuntz

DATE 11/1/19 11/8/19 11/12/19

OFFICIAL RECORD COPY

Documents

Prairie Island Nuclear Generating Plant, Units 1 and 2 ...Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendments 230 and 218 RE: Issuance of Amendments RE: