ML22334A073

Issuance of Amendment Nos. 363 and 357 Regarding Modification of the Approved Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

ML22334A073
Person / Time
Site:	Sequoyah
Issue date:	02/06/2023
From:	Perry Buckberg Plant Licensing Branch II
To:	Jim Barstow Tennessee Valley Authority
Wentzel M
References
EPID L-2022-LLA-0033
Download: ML22334A073 (20)

Text

February 6, 2023 Mr. James Barstow Vice President, Nuclear Regulatory Affairs and Support Services Tennessee Valley Authority 1101 Market Street, LP 4A-C Chattanooga, TN 37402-2801

SUBJECT:

SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 363 AND 357 REGARDING MODIFICATION OF THE APPROVED RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS, AND COMPONENTS FOR NUCLEAR POWER REACTORS (EPID L-2022-LLA-0033)

Dear Mr. Barstow:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 363 to Renewed Facility Operating License No. DPR-77, and Amendment No. 357 to Renewed Facility Operating License No. DPR-79, for the Sequoyah Nuclear Plant, Units 1 and 2, respectively. These amendments are in response to your application dated February 24, 2022, as supplemented by a letter dated September 16, 2022.

The amendments modify the Sequoyah Nuclear Plant, Units 1 and 2 license to implement a change to the approved voluntary implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.69, Risk-Informed categorization and treatment of structures, systems and components [SSCs] for nuclear power reactors.

Specifically, the amendments replace the previously approved approaches to evaluate seismic and fire risk with peer reviewed, plant-specific Sequoyah Nuclear Plant, Units 1 and 2 seismic probabilistic risk assessment and fire probabilistic risk assessment models for categorization of SSCs under the licensees previously approved 10 CFR 50.69 program.

Sincerely,

/RA/

Perry H. Buckberg, Senior Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-327 and 50-328

Enclosures:

1. Amendment No. 363 to DPR-77

2. Amendment No. 357 to DPR-79

3. Safety Evaluation cc: Listserv

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SEQUOYAH NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 363 Renewed License No. DPR-77

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated February 24, 2022, as supplemented by a letter dated September 16, 2022, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in Title 10 of the Code of Federal Regulations (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 Commissions regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

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

2.

Accordingly, by amendment number 363, Renewed Facility Operating License DPR-77 is hereby amended as set forth in the licensees application dated February 24, 2022, as supplemented by a letter dated September 16, 2022, and evaluated in the NRC staffs safety evaluation dated February 6, 2023.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: February 6, 2023 David J.

Wrona Digitally signed by David J. Wrona Date: 2023.02.06 12:52:33 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 363 SEQUOYAH NUCLEAR PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 50-327 Revise the Renewed Facility Operating License No. DPR-77 by removing the page identified below and inserting the attached page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page Insert Page 14a 14a

- 14a -

(d)

For SRs that existed prior to this amendment that have modified acceptance criteria, the first performance subject to the modified acceptance criteria is due at the end of the first Surveillance interval that began on the date the Surveillance was last performed prior to the implementation of this amendment.

(33) Adoption of 10 CFR 50.69, "Risk-Informed categorization and treatment of structures, systems and components for nuclear power plants" (1)

TVA is approved to implement 10 CFR 50.69 using the processes for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) model to evaluate risk associated with internal events, including internal flooding; a fire PRA; a seismic PRA; the NUMARC 96-01 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., a screening of external hazards updated using the criteria in the endorsed ASME/ANS RA-Sa-2009 PRA Standard for external hazard screening significance; as specified in Unit 1 License Amendment 346 (as revised by License Amendment 363).

(2)

Prior to implementation of the provisions of 1 0CFR 50.69, TVA shall complete the items below; a.

Items listed in Enclosure 1, Attachment 1, "SON 10 CFR 50.69 PRA Implementation Items," in TVA letter CNL-19-002, "Response to Request for Additional Information Regarding Application to Modify Sequoyah Nuclear Plant Units 1 and 2, Application to Adopt 10 CFR 50.69, "Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, (SON-TS-17-06)(EPID: L-2018-LLA-0066)," dated March 21, 2019.

(3)

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from the NUMARC 96-01 shutdown safety assessment process to assess shutdown risk to a shutdown PRA approach).

AmendmentM. 346, 363 Renewed License No. DPR 77

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 357 Renewed License No. DPR-79

1.

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated February 24, 2022, as supplemented by a letter dated September 16, 2022, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

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

2.

Accordingly, by amendment number 357, Renewed Facility Operating License DPR-79 is hereby amended as set forth in the licensees application dated February 24, 2022, as supplemented by a letter dated September 16, 2022, and evaluated in the NRC staffs safety evaluation dated February 6, 2023.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: February 6, 2023 David J.

Wrona Digitally signed by David J. Wrona Date: 2023.02.06 12:52:55 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 357 SEQUOYAH NUCLEAR PLANT, UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-79 DOCKET NO. 50-328 Revise the Renewed Facility Operating License No. DPR-79 by removing the pages identified below and inserting the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Page Insert Page 13a 13b 13a 13b

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Amendment No. 338, 340, 348, 357 Renewed License No. DPR 79 relocation of the requirements to the specified documents, as described in Table R, Relocated Specifications and Removed Detail Changes, attached to the NRC staff's Safety Evaluation, which is enclosed in this amendment.

2.

Schedule for New and Revised Surveillance Requirements (SRs) The schedule for performing SRs that are new or revised in License Amendment 327 shall be as follows:

(a)

For SRs that are new in this amendment, the first performance is due at the end of the first Surveillance interval, which begins on the date of implementation of this amendment.

(b)

For SRs that existed prior to this amendment, whose intervals of performance are being reduced, the first reduced Surveillance interval begins upon completion of the first Surveillance performed after implementation of this amendment.

(c)

For SRs that existed prior to this amendment, whose intervals of performance are being extended, the first extended Surveillance interval begins upon completion of the last Surveillance performed prior to implementation of this amendment.

(d)

For SRs that existed prior to this amendment that have modified acceptance criteria, the first performance subject to the modified acceptance criteria is due at the end of the first Surveillance interval that began on the date the Surveillance was last performed prior to the implementation of this amendment.

(26)

DELETED (27)

Adoption of 10 CFR 50.69, "Risk-Informed categorization and treatment of structures, systems and components for nuclear power plants" (1)

TVA is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) model to evaluate risk associated with internal events, including internal flooding; a fire PRA; a seismic PRA; the NUMARC 96-01 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

Amendment No. 342, 357 Renewed License No. DPR 79

- 13b -

IPEEE Screening Assessment for External Hazards, i.e., a screening of external hazards updated using the criteria in the endorsed ASME/ANS RA-Sa-2009 PRA Standard for external hazard screening significance; as specified in Unit 2 License Amendment 340 (as revised by License Amendment 357).

(28)

Prior to Cycle 24 startup from Unit 2 Refueling Outage 23, TVA shall ensure the Cycle 24 core design will not adversely affect the safety of the plant in accordance with TVA procedure, NFDP-111, Nuclear Design and Core Analysis.

(2)

Prior to implementation of the provisions of 1 0 CFR 50.69, TVA shall complete the items below; a.

Items listed in Attachment 1, "SQN 10 CFR 50.69 PRA Implementation Items," in TVA letter CNL-19-002, "Response to Request for Additional Information Regarding Application to Modify Sequoyah Nuclear Plant Units 1 and 2, Application to Adopt 10 CFR 50.69, "Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, (SQN-TS-17-06)

(EPID: L-2018-LLA-0066)," dated March 21, 2019.

(3)

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from the NUMARC 96-01 shutdown safety assessment process to assess shutdown risk to a shutdown PRA approach).

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 363 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-77 AND AMENDMENT NO. 357 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-79 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328

1.0 INTRODUCTION

By letter to the U.S. Nuclear Regulatory Commission (NRC, the Commission) dated February 24, 2022 [1], as supplemented by letter dated September 16, 2022 [2], Tennessee Valley Authority (TVA, the licensee) submitted a license amendment request (LAR) for the Sequoyah Nuclear Plant, Units 1 and 2 (SQN).

The proposed amendments would modify the SQN licensing basis to implement a change to the approved voluntary implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.69, Risk-Informed categorization and treatment of structures, systems and components [SSCs] for nuclear power reactors. Specifically, the proposed amendments would replace the previously approved approaches to evaluate seismic and fire risk with peer reviewed, plant-specific SQN seismic probabilistic risk assessment (SPRA) and fire probabilistic risk assessment (FPRA) models for categorization of SSCs under the licensees previously approved 10 CFR 50.69 program. In addition, the licensee proposed an administrative change to paragraph (3) of the approved 10 CFR 50.69 License Conditions to cite a different example of a change to the categorization process which would require NRC prior approval under 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit.

To support its review, the NRC staff requested additional information from the licensee [3]. The licensee responded to the requests for additional information (RAIs) in a supplemental letter dated September 16, 2022 [2]. The supplemental letter provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on April 19, 2022 (87 FR 23274).

2.0 REGULATORY EVALUATION

2.1 Applicable Regulations The provisions of 10 CFR 50.69 allow adjustment of the scope of SSCs subject to special treatment requirements. Special treatment refers to those requirements that provide increased assurance beyond normal industry practices that SSCs perform their design basis functions. For SSCs categorized as low safety significance (LSS), alternative treatment requirements may be implemented in accordance with the regulation. For SSCs determined to be of high safety significance (HSS), requirements may not be changed.

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

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

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

2.2 Regulatory Guidance Documents The NRC staff considered the following regulatory guidance during its review of the proposed changes:

Regulatory Guide (RG) 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance [4];

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities [5];

RG 1.174, Revision 2, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis [6];

NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking [7]; and NUREG-0800, Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Chapter 19, Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis:

General Guidance [8].

NRC-Endorsed Guidance The Nuclear Energy Institute (NEI) issued NEI 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline [9], as endorsed by RG 1.201, Revision 1 with clarifications, describes a process that the NRC staff considers acceptable for complying with 10 CFR 50.69. This process determines the safety significance of SSCs and categorizes them into one of four RISC categories defined in 10 CFR 50.69.

3.0 TECHNICAL EVALUATION

The NRC staff reviewed the proposed change to the previously approved SQN 10 CFR 50.69 program [10]. In its LAR, the licensee stated that this LAR is to replace, within the approved SQN 10 CFR 50.69 program, the use of the Seismic Safe Shutdown Equipment List (SSSEL) with the use of the SQN SPRA, and the use of the Fire Safe Shutdown Equipment List (FSSEL) with the use of the SQN FPRA, in accordance with NEI 00-04 [9] and RG 1.201, Revision 1. All other previously approved screening and categorization methods are not affected by this LAR.

The NRC staffs review confirmed that the current LAR does not change any other aspect of the licensees categorization process except for the use of SPRA and FPRA instead of the SSSEL and FSSEL, respectively. Therefore, other than the use of SPRA and FPRA for the consideration of the seismic and fire risk, the NRC staffs previous review and decisions in the letter dated September 18, 2019 [10] on the licensees categorization process remain unchanged and valid. Consequently, the NRC staff did not separately review the licensees categorization process other than the changes requested in the LAR.

As stated in RG 1.201, Revision 1, if a licensee wishes to change its categorization approach, the staffs review of the resulting submittal will focus on the technical adequacy of the methodology and analyses relied upon in the application. Sections 3.1 and 3.2 summarize the NRC staffs review of the acceptability of the SPRA and FPRA, respectively, as described in the LAR and its supplement.

3.1 Seismic PRA Model Evaluation The NRC staff reviewed the proposed use of SPRA in lieu of the SSSEL for the previously approved 10 CFR 50.69 categorization process. The NRC staffs review of the licensees SPRA was based on the results of the peer review and the associated Independent Assessments for closure of facts and observations (F&Os) described in Section 3.1 of the LAR.

The licensee stated that the SPRA model was peer reviewed against the requirements of Part 5 of American Society of Mechnical Engineers (ASME)/American Nuclear Society (ANS)

RA-Sb-2013 (Addendum B of the PRA Standard) [11], which is not endorsed by RG 1.200, Revision 2. The licensee stated in its LAR that all but six of Addendum B supporting requirements (SRs) have been shown to either be equal to or envelop the corresponding Addendum A SRs of the ASME/ANS 2009 Standard [12], which is endorsed by RG 1.200, Revision 2. The licensee then provided the assessment of the remaining six SRs in Table E-1 of the LAR to show conformity with Addendum A SRs. Based on its review, the NRC staff finds that the licensees use of Part 5 of Addendum B adequately addresses the technical elements for the development of an SPRA. Therefore, the NRC staff concludes that the use of Part 5 of Addendum B is an acceptable alternative to the NRC-endorsed Addendum A approach for the licensees SPRA used to support this application.

The licensee stated that the SQN SPRA model received a full-scope peer review in April 2018 using NEI guidelines NEI 12-13, with a total of 53 unique Finding level F&Os generated.

Subsequently, in February 2019, the licensee conducted an Independent Assessment and Focused-Scope Peer Review for closure of all the 53 F&Os. The F&O closure review was performed in accordance with Appendix X to NEI 12-13. In response to RAI APLC-01 [2], the licensee further explained that four of the seven seismic hazard analysis (SHA) findings were assessed to be upgrades, which required a focused scope peer review. All four reviewed SRs were assessed to be met and three new F&Os were generated. These three F&Os were ultimately closed in the April 17, 2019, F&O closure review. Based on its review, the NRC staff finds that the SQN SPRA was appropriately peer reviewed consistent with RG 1.200, Revision 2, F&Os were reviewed consistent with NRC accepted guidance, and all F&O findings have been closed.

In summary, the NRC staff reviewed the SPRA model peer review history provided by the licensee in Section 3.1 of the LAR, as supplemented. The NRC staff finds that the licensee adequately applied the guidance in RG 1.200, Revision 2 for establishing SPRA technical acceptability for this application.

3.2 Internal Fire PRA Model Evaluation The NRC staff reviewed the proposed use of FPRA in lieu of the FSSEL for the previously approved 10 CFR 50.69 categorization process. The NRC staffs technical adequacy determination of the SQN FPRA was based on the results of the peer review and the associated Independent Assessments for closure of F&Os described in Section 3.2 of the LAR.

The licensee stated that the SQN FPRA had been peer reviewed using the ASME/ANS RA-Sa-2009 PRA Standard [12], NEI 07-12, Revision 1, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines [15], and RG 1.200, Revision 2. Part 4 (Fire) of the ASME/ANS PRA Standard contains a total of 173 SRs under 13 technical elements. In its LAR, the licensee also stated that a total of 25 of the SRs were judged by the peer review team to be not applicable to SQN, and therefore, the remaining 148 SRs were reviewed. The SQN FPRA was reviewed against capability category II (CC-II) of the PRA Standard for all applicable SRs. In this review, the licensee concluded that all but 9 of the 148 SRs were met at CC-II or higher. Six SRs were not met and three were met at CC-I.

The F&Os identified from the above SQN FPRA peer review were addressed during a subsequent update of the FPRA, and in April 2020, the licensee conducted an Independent Assessment for closure of the F&Os in accordance with NEI 07-12, Appendix X. The assessment resolved all F&Os that were partially closed or open, and the nine SRs previously found to not be met or met at CC-I were assessed as met at CC-II or better. No upgrades were required as a part of the F&O closure review.

Based on the above, the NRC staff finds the SQN FPRA was appropriately peer reviewed consistent with RG 1.200, Revision 2, fire methodologies were appropriately considered and implemented, and all F&O findings were closed consistent with NEI 07-12, Appendix X.

Therefore, the NRC staff concluded that the licensee adequately applied the guidance for establishing SQN FPRA technical quality and is acceptable for use in lieu of the FSSEL for the 10 CFR 50.69 categorization process.

3.3 PRA Model Maintenance and Updates In Section 3.3 of the LAR, the licensee stated that the applicable PRA models, including the SPRA and FPRA models, used in this application continue to reflect the as-built and as-operated plant for each of the SQN units, and if there is a significant impact on the PRA model, the SSC categorization will be re-evaluated. In addition, the licensee stated that any PRA model upgrades to any of the PRA models used in support of the SQN 10 CFR 50.69 process will be peer reviewed prior to implementing those changes in the PRA model used for categorization. The NRC staff finds that this process is consistent with NEI 00-04 [9], and, therefore, it is acceptable for this application.

3.4 Modeling of FLEX Equipment In Section 3.4 of the LAR, the licensee discussed PRA modeling of FLEX equipment, including permanently installed diesel generators. The licensee stated that no portable FLEX equipment is credited in the models. The NRC staff finds that the credit for FLEX equipment is appropriate because the licensee included the permanently installed equipment with operator actions in the SQN PRA models for this application.

3.5 Key Assumptions and Sources of Uncertainty Section 3.6 of the LAR describes how the licensee assessed assumptions and sources of uncertainty to identify key assumptions and key sources of uncertainty for this application. The licensee explains the two-step process, including a review of plant-specific assumptions and sources of uncertainty identified from the PRAs, and a review of the generic sources of uncertainty from Electric Power Research Institute (EPRI) 1016737 [13] and EPRI 1026511 [14].

This process is consistent with NUREG-1855, Revision 1 [7], Stage E: Assessing Model Uncertainty, which is to identify, screen, and characterize those sources of model uncertainty and related assumptions in the base PRA.

The licensee stated that to determine whether each assumption or uncertainty for the 10 CFR 50.69 application is applicable, the assumption or source of uncertainty was assessed against the five criteria listed in Section 3.6 of the LAR. In response to RAI APLC-03 [2], the licensee stated that based on the review performed, none of the assumptions or uncertainties were identified as being key with respect to the 10 CFR 50.69 application. The NRC staff finds that the five criteria are based on NUREG-1855 and RG 1.200, Revision 2. The NRC staff finds that the assessment performed by the licensee to identify the key assumptions and sources of uncertainty is consistent with NEI 00-04 and is acceptable for this application.

In Section 3.7.2 of the LAR, the licensee discussed the sensitivity study which utilized a factor of 3 to increase the unavailability or unreliability of LSS components modeled in PRAs, including the SPRA and FPRA for this application. The licensee also discussed the seismic capacity of LSS components in Section 3.7.3 of the LAR. The licensee stated that TVA has a program for monitoring degradation that could affect the seismic capacity of components at a periodic frequency, and if such degradation is identified that an evaluation would be performed to determine if the original categorization remains valid. The NRC staff finds that this process is consistent with the NEI 00-04 guidance and, therefore, it is acceptable for this application.

In Attachment 1 of the LAR, the licensee provided total baseline core damage frequency (CDF) and large early release frequency (LERF) values for the internal events including internal flooding, fire, and seismic PRAs based on point estimate values. The NRC staff confirms that the total CDF and LERF meet the RG 1.174, Revision 2 risk acceptance guidelines.

3.6 Integrated Importance Measures Paragraph (c)(1)(ii) of 10 CFR 50.69 requires, in part, that the SSC functional importance be determined using an integrated, systematic process. Section 5.6 of NEI 00-04, Revision 0 [9],

Integral Assessment, discusses the need for an integrated computation using the available importance measures. The guidance in NEI 00-04, Revision 0 further states, in part, that the integrated importance measure essentially weights the importance from each risk contributor (e.g., internal events, fire, seismic [and high wind] PRAs) by the fraction of the total core damage frequency [or large early release frequency] contributed by that contributor. The guidance also provides formulas to compute the Integrated Fussell-Vesely Importance, and Integrated Risk Achievement Worth.

In Section 3.7.1 of the LAR, as supplemented, the licensee stated that the integrated importance measures for a large majority of SPRA and FPRA basic events are calculated using the equations presented in NEI 00-04 Section 5.6, and the resulting integrated importance measures are compared against the screening criteria in NEI 00-04. These equations essentially weight the importance from each risk contributor (i.e., internal events including flooding, fire, and seismic) by the fraction of that contributor to the total CDF or total LERF.

There are some SSCs in the SPRA and/or FPRA that are not directly included in the other PRA models. In response to RAI APLC-02 [2], the licensee further explained the integral assessment process for those SSCs by using examples to demonstrate how SSCs only in the SPRA and FPRA models would be treated for the importance analysis. The licensee stated that the safety significance would be presented to the Integrated Decision-making Panel for consideration in the decision-making process. The NRC staff finds that the licensees use and treatment of importance measures is consistent with the guidance in NEI 00-04, Revision 0 [9], as endorsed in RG 1.201, Revision 1.

3.7 Technical Evaluation Summary Pursuant to 10 CFR 50.69(c)(1)(i), the categorization process must consider results and insights from the plant-specific PRA. The use of the FPRA and SPRA to support SSC categorization is endorsed by RG 1.201, Revision 1. The PRAs 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 that is endorsed by the NRC.

The staff reviewed the peer-review history, the licensees resolution of peer-review findings, and the identification and disposition of key assumptions and sources of uncertainty. The staff concludes that (1) the licensees SPRA and FPRA are acceptable to support the categorization of SSCs using the process endorsed by RG 1.201, Revision 1, and (2) the key assumptions for the PRAs have been identified consistent with the guidance in RG 1.200, Revision 2 [5] and NUREG-1855 [7], as applicable, and addressed appropriately for this application. The staff finds that the licensee provided the required information and that the SPRA and FPRA are acceptable and, therefore, meet the requirements set forth in 10 CFR 50.69(c)(1)(i) and (ii). In addition, the staff finds that the licensees proposed changes to paragraph (3) of the approved 10 CFR 50.69 License Conditions are administrative in nature and require no further evaluation.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Tennessee State official was notified of the proposed issuance of the amendments on November 28, 2022. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement 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, published in the Federal Register on April 19, 2022 (87 FR 23274), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

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

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions 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] Polickoski, James, Tennessee Valley Authority to U.S. Nuclear Regulatory Commission, Docket Numbers 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Renewed Facility Operating License Nos. DPR-77 and DPR-79, License Amendment Request to Modify Approved 10 CFR 50.69 Categorization Process (SQN-TS-21-07), dated February 24, 2022 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML22055A625).

[2] Barstow, James, Tennessee Valley Authority to U.S. Nuclear Regulatory Commission, Docket Numbers 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Renewed Facility Operating License Nos. DPR-77 and DPR-79, Response to Request for Additional Information Regarding License Amendment Request to Modify Approved 10 CFR 50.69 Categorization Process (SQN-TS-21-07), dated September 16, 2022 (ML22259A127).

[3] Buckberg, Perry, U.S. Nuclear Regulatory Commission, to Taylor, Andrew Charles, Tennessee Valley Authority, Request for Additional Information Related to SQN Request for Fire and Seismic PRA Modification to 50.69 L-2022-LLA-0033 dated August 2, 2022 (ML22214A158).

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

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

[6] Regulatory Guide 1.174, Revision 2, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, dated May 2011 (ML100910006).

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

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

[9] NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0, dated July 2005 (ML052910035).

[10] Hon, Andrew, U.S. Nuclear Regulatory Commission, to Barstow, James, Tennessee Valley Authority, Sequoyah Nuclear Plant, Units 1 and 2 - Issuance of Amendment Nos. 346 and 340 Re: Request to Adopt CFR 50.69, Risk-Informed Categorization and Treament of Structures, Systems, and Components for Nuclear Power Reactors (EPID L-2018-LLA-0066), dated September 18, 2019 (ML19179A135).

[11] ASME/ANS RA-Sb-2013, Standard for Level l/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum B to RA-S-2008, American Society of Mechanical Engineers, New York, NY, American Nuclear Society, La Grange Park, Illinois, dated July 2013.

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

PRA standard ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, dated February 2009, New York, NY (Copyright).

[13] Electric Power Research Institute (EPRI), Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, EPRI Technical Report 1016737, dated December 2008. (Access at https://www.epri.com/research/products/1016737.)

[14] Electric Power Research Institute (EPRI), Practical Guidance on the Use of Probabilistic Risk Assessments in Risk-Informed Applications with a Focus on the Treatment of Uncertainty, Technical Update Report 1026511, dated December 2012. (Access at https://www.epri.com/research/products/1026511.)

[15] Nuclear Energy Institute, NEI 07-12, Revision 1, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, dated June 2010 (ML102230070).

Principal Contributors: D. W. Wu, NRR T. Dinh, NRR K. Tetter, NRR J. Robinson, NRR Date: February 6, 2023

ML22334A073 OFFICE NRR/DORL/LPLII-2/PM NRR/DORL/LPLII-2/LA NRR/DRA/APLC/BC OGC NAME PBuckberg RButler SVasavada MFWoods DATE 12/8/2022 12/8/2022 11/22/2022 1/10/2023 OFFICE NRR/DORL/LPLII-2/BC NRR/DORL/LPLII-2/PM NAME DWrona PBuckberg DATE 2/6/2023 2/6/2023