ML24108A070

– Issuance of License Amendments to Revise the 10 CFR 50.69 Categorization Process to Reflect an Alternative Seismic Approach

ML24108A070
Person / Time
Site:	Brunswick
Issue date:	06/05/2024
From:	Luke Haeg NRC/NRR/DORL/LPL2-2
To:	Krakuszeski J Duke Energy Progress
References
EPID L-2023-LLA-0117
Download: ML24108A070 (1)

Text

June 5, 2024 Mr. John A. Krakuszeski Site Vice President Brunswick Steam Electric Plant Duke Energy Progress, LLC 8470 River Rd., SE (M/C BNP001)

Southport, NC 28461

SUBJECT:

BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 - ISSUANCE OF LICENSE AMENDMENTS TO REVISE THE 10 CFR 50.69 CATEGORIZATION PROCESS TO REFLECT AN ALTERNATIVE SEISMIC APPROACH (EPID L-2023-LLA-0117)

Dear John Krakuszeski:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment Nos. 313 and 341 to Renewed Facility Operating License (RFOL) Nos. DPR-71 and DPR-62 for the Brunswick Steam Electric Plant, Units Nos. 1 and 2 (Brunswick),

respectively. These amendments revise the Brunswick Units 1 and 2 RFOLs in response to your license amendment request (LAR) dated August 17, 2023, associated with Brunswicks adoption 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 revise the Brunswick Units 1 and 2 RFOL conditions to allow the use of an alternative approach for evaluating seismic risk for categorization of SSCs under Brunswicks previously approved 10 CFR 50.69 program and remove the requirements for the completion of certain pre-program implementation items that have been completed.

J. Krakuszeski A copy of the safety evaluation is also enclosed. A Notice of issuance will be included in the Commissions monthly Federal Register notice.

If you have any questions, please contact me at (301) 415-0272 or by e-mail at Lucas.Haeg@nrc.gov.

Sincerely,

/RA/

Lucas Haeg, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos.: 50-325 and 50-324

Enclosures:

1. Amendment No. 313 to DPR-71

2. Amendment No. 341 to DPR-62

3. Safety Evaluation cc: Listserv

DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 313 Renewed License No. DPR-71

1.

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

A.

The application for amendment filed by Duke Energy Progress, LLC (the licensee), dated August 17, 2023, 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, the license is amended by changes as indicated in the attachment to this license amendment, and paragraph 3 of Renewed Facility Operating License No. DPR-71 is hereby amended to read as follows:

Additional Conditions The Additional Conditions contained in Appendix B, as revised through Amendment No. 313, are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Additional Conditions.

3.

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

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

Changes to the Renewed Facility Operating License and Appendix B, Additional Conditions Date of Issuance: June 5, 2024 DAVID WRONA Digitally signed by DAVID WRONA Date: 2024.06.05 12:58:57 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 313 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-71 DOCKET NO. 50-325 Replace the following pages of Renewed Facility Operating License No. DPR-71 and Appendix B, Additional Conditions, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Renewed Facility Operating License Remove Page Insert Page 10 10 Appendix B, Additional Conditions Remove Page Insert Page App. B-5 App. B-5 Renewed License No. DPR-71 Amendment No. 313 3.

Additional Conditions The Additional Conditions contained in Appendix B, as revised through Amendment No. 313 are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Additional Conditions.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:

1.

Unit 1 - Technical Specifications - Appendices A and B Date of Issuance: June 26, 2006

Brunswick Unit 1 App. B-5 Amendment No. 313 Amendment Number Implementation Date 313 Additional Conditions Duke Energy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and high winds; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 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 updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in Unit 1 License Amendment No. 305 dated April 30, 2021.

In addition, Duke Energy is approved to implement 10 CFR 50.69 using the alternative seismic approach for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs as described in Duke Energy letter RA-23-0122, dated August 17, 2023, as specified in Unit 1 License Amendment No. 313 dated June 5, 2024.

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

Upon implementation of Amendment No. 313.

DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 341 Renewed License No. DPR-62

1.

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

A.

The application for amendment filed by Duke Energy Progress, LLC (the licensee), dated August 17, 2023, 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, the license is amended by changes as indicated in the attachment to this license amendment, and paragraph 3 of Renewed Facility Operating License No. DPR-62 is hereby amended to read as follows:

Additional Conditions The Additional Conditions contained in Appendix B, as revised through Amendment No. 341, are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Additional Conditions.

3.

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

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

Changes to the Renewed Facility Operating License and Appendix B, Additional Conditions Date of Issuance: June 5, 2024 DAVID WRONA Digitally signed by DAVID WRONA Date: 2024.06.05 12:59:29 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 341 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 FACILITY OPERATING LICENSE NO. DPR-62 DOCKET NO. 50-324 Replace the following pages of Renewed Facility Operating License No. DPR-71 and Appendix B, Additional Conditions, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Renewed Facility Operating License Remove Page Insert Page 10 10 Appendix B, Additional Conditions Remove Page Insert Page App. B-5 App. B-5 Renewed License No. DPR-62 Amendment No. 341 M.

Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(1)

Fire fighting response strategy with the following elements:

1.

Pre-defined coordinated fire response strategy and guidance 2.

Assessment of mutual aid fire fighting assets 3.

Designated staging areas for equipment and materials 4.

Command and control 5.

Training of response personnel (2)

Operations to mitigate fuel damage considering the following:

1.

Protection and use of personnel assets 2.

Communications 3.

Minimizing fire spread 4.

Procedures for implementing integrated fire response strategy 5.

Identification of readily-available pre-staged equipment 6.

Training on integrated fire response strategy 7.

Spent fuel pool mitigation measures (3)

Actions to minimize release to include consideration of:

1.

Water spray scrubbing 2.

Dose to onsite responders N.

The licensee shall implement and maintain all Actions required by Attachment 2 to NRC Order EA-06-137, issued June 20, 2006, except the last action that requires incorporation of the strategies into the site security plan, contingency plan, emergency plan and/or guard training and qualification plan, as appropriate.

3.

Additional Conditions The Additional Conditions contained in Appendix B, as revised through Amendment No. 341, are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Additional Conditions.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:

1.

Unit 2 - Technical Specifications - Appendices A and B Date of Issuance: June 26, 2006

Brunswick Unit 2 App. B-5 Amendment No. 341 Amendment Number Implementation Date 341 Additional Conditions Duke Energy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and high winds; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 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 updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in Unit 2 License Amendment No. 333 dated April 30, 2021.

In addition, Duke Energy is approved to implement 10 CFR 50.69 using the alternative seismic approach for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs as described in Duke Energy letter RA-23-0122, dated August 17, 2023, as specified in Unit 2 License Amendment No. 341 dated June 5, 2024.

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

Upon implementation of Amendment No. 341.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 313 AND 341 RENEWED FACILITY OPERATING LICENSE NOS. DPR-71 AND DPR-62 DUKE ENERGY PROGRESS, LLC BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 DOCKET NOS. 50-325 AND 50-324

1.0 INTRODUCTION

By letter dated August 17, 2023 (Reference 1), Duke Energy Progress, LLC (Duke Energy, the licensee) submitted a license amendment request (LAR) to Renewed Facility Operating License (RFOL) Nos. DPR-71 and DPR-62, for the Brunswick Steam Electric Plant, Units 1 and 2 (Brunswick), respectively, associated with Brunswicks adoption 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 LAR proposed amendments to allow the use of an alternative approach for evaluating seismic risk for categorization of SSCs under Brunswicks previously approved 10 CFR 50.69 program and to remove the requirements for the completion of certain pre-program implementation items that have been completed.

2.0 REGULATORY EVALUATION

2.1 Background

In the safety evaluation (SE) for License Amendment Nos. 305 and 333 to RFOL Nos. DPR-71 and DPR-62 for Brunswick dated April 30, 2021 (Reference 4), the U.S. Nuclear Regulatory Commission (NRC) staff concluded that the licensees 10 CFR 50.69 program was consistent with the NRC-endorsed guidance in the Nuclear Energy Institute (NEI) 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline (Reference 10), and thus satisfied the requirements of 10 CFR 50.69(c). In the previously approved categorization process, the licensee used the seismic margin analysis (SMA) for the consideration of seismic risk in the categorization process. A license condition incorporated into the license as part of the NRC staffs decision to approve the licensees original 10 CFR 50.69 LAR stated that [p]rior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

For the subject LAR (Reference 1), the licensee proposed to allow the use of an alternative seismic approach for evaluating seismic risk in addition to the SMA for categorization of SSCs described in the licensees previously approved 10 CFR 50.69 program.

2.2 Description of Changes A license condition was added to the Brunswick RFOLs when the NRC approved the licensees use of 10 CFR 50.69 on April 30, 2021 (Reference 4). As discussed in 10 CFR 50.69(b), a licensee may voluntarily comply with 10 CFR 50.69 as an alternative to compliance with the following requirements for Risk Informed Safety Class (RISC)-3 and RISC-4 SSCs after it submits, and the NRC approves, an application for a license amendment:

(i) 10 CFR Part 21, Reporting of Defects and Noncompliance; (ii)

A portion of 10 CFR 50.46a, Acceptance criteria for reactor coolant system venting systems, paragraph (b);

(iii) 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants; (iv) 10 CFR 50.55, Conditions of construction permits, early site permits, combined licenses and manufacturing licenses, paragraph (e);

(v)

Certain requirements of 10 CFR 50.55a, Codes and Standards; (vi) 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants, except for paragraph (a)(4);

(vii) 10 CFR 50.72, Immediate notification requirements for operating nuclear power reactors; (viii) 10 CFR 50.73, License event report system; (ix) 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants; (x)

Certain containment leakage testing requirements in 10 CFR Part 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors; and (xi)

Certain requirements of 10 CFR Part 100, Appendix A, Seismic and Geologic Siting Criteria for Nuclear Power Plants.

The current license condition for Units 1 and 2 (added by License Amendment Nos. 305 and 333) states the following [Unit 2 differences as noted]:

Duke Energy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and high winds; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non PRA evaluations that are based on the [individual plant examination of external events] IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in Unit 1 [2]

License Amendment No. 305 [333] dated April 30, 2021.

Duke Energy will complete the implementation items listed in Attachment 1 of Duke letter to the NRC dated November 24, 2020, prior to implementation of

10 CFR 50.69 in accordance with the categorization process described above.

The issues identified in the attachment will be addressed and any associated changes will be made prior to implementation of 10 CFR 50.69 in accordance with the categorization process described above.

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

The licensee proposed in the subject LAR to amend the above license condition for Unit Nos. 1 and 2 to read as follows:

Duke Energy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and high winds; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 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 updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in Unit 1 [2] License Amendment No. 305 [333] dated April 30, 2021.

In addition, Duke Energy is approved to implement 10 CFR 50.69 using the alternative seismic approach for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs as described in Duke Energy letter RA 23 0122, dated August 17, 2023, as specified in Unit 1 [2] License Amendment No. [XXX] dated [DATE].

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

In Section 3 of the LAR enclosure, the licensee stated that it would use a single approach (i.e.,

either the SMA or the proposed alternative seismic approach) for the categorization of an entire system. The licensee proposed that it could implement the alternate seismic approach for any system that was previously categorized, or for systems that will be categorized, and that it would not be required to re-categorize (with the alternative seismic approach) any system that it previously categorized. The licensee proposed that it may continue to use the processes identified in the current applicable license conditions.

2.3 Applicable Regulatory Requirements and Guidance 2.3.1 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), the requirements may not be changed.

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

Categorization of SSCs does not allow for the elimination of SSC functional requirements or allow equipment that is required by the deterministic design basis to be removed from the facility. Instead, 10 CFR 50.69 enables licensees to focus their resources on SSCs that make a significant contribution to plant safety. For SSCs that are categorized as HSS, existing treatment requirements are maintained or potentially enhanced. Conversely, for SSCs categorized as LSS that do not significantly contribute to plant safety on an individual basis, the regulation allows an alternative risk informed approach to treatment that provides a reasonable level of confidence that these SSCs will satisfy functional requirements. Implementation of 10 CFR 50.69 allows licensees to improve focus on equipment that is HSS.

2.3.2 Regulatory Guidance 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 (Reference 5);

RG 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk Informed Activities (Reference 6);

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 7);

and, NUREG 1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs [probabilistic risk assessments] in Risk Informed Decisionmaking (Reference 8).

3.0 TECHNICAL EVALUATION

The NRC staff reviewed the proposed change with respect to the previously-approved Brunswick 10 CFR 50.69 program. In its LAR, the licensee stated that, except for the proposed adoption of an alternate seismic approach, all other previously-approved 10 CFR 50.69 categorization methods would not be impacted by this LAR. The NRC staffs review confirmed that the LAR did not impact or change any other aspect of the licensees categorization process except for the addition of the alternative seismic approach to consider the seismic risk.

Therefore, the NRC staffs determinations in the letter dated April 30, 2021 (Reference 4),

concerning the licensees categorization process, other than the addition of the alternative seismic approach related to the consideration of the seismic risk, remain unchanged and valid.

Further, in Section 3 of the enclosure to the LAR, the licensee stated that it would use a single

approach (i.e., either the SMA or the proposed alternative seismic approach) for the categorization of an entire system. Consequently, the NRC staff did not separately review the licensees categorization process other than the change requested in the LAR.

As stated in Regulatory Guide (RG) 1.201 (Reference 5), if a licensee wishes to change its categorization approach, the staffs review of the resulting license amendment request will focus on the acceptability of the methodology and analyses relied upon in the application. Section 3.1 below summarizes the NRC staffs review of the acceptability of the proposed alternative seismic approach as described in the LAR.

In Section 2.3, Description of the Proposed Change, of the LAR, the licensee stated that the implementation items listed in the previously-approved amendment in Attachment 1 of Duke Energys letter to the NRC (Reference 13) were completed. Section 3.2 below summarizes the NRC staffs review of the PRA implementation items listed in the previously approved amendment that would be removed from the license condition as described in the LAR.

3.1 Alternative Seismic Approach In Section 3 of the LAR enclosure, the licensee stated that the proposed process allows for the following non-PRA methods for the risk characterization:

Seismic hazard is assessed using the seismic margins assessment in accordance with NEI 00-04, and as approved in the letter dated April 30, 2021 (Reference 4).

Seismic hazard is assessed using an alternative approach supported by Electric Power Research Institute (EPRI) Report No. 3002017583 (Reference 12), as described in the LAR.

The NRC staff notes that use of the alternative seismic approach is an alternative method not endorsed by the NRC in NEI 00-04 (Reference 10). A detailed NRC staff review of the licensees proposed alternative seismic approach follows.

As part of its proposed process to categorize SSCs according to safety significance, the licensee proposed to use a non-PRA method to consider seismic hazards. The regulation in 10 CFR 50.69(b)(2)(ii) and 50.69(c)(1)(ii) permit the use of systematic evaluation techniques in the risk informed categorization process. The licensee provided a description of its proposed alternative seismic approach for considering seismic risk in the categorization process and described how the proposed alternative seismic approach would be used in the categorization process in Section 3 of the LAR enclosure. In part, the licensee based its plant-specific evaluation on the case studies performed in EPRI Report No. 3002017583. The licensee stated that Brunswick is using test case information from EPRI Report No. 3002017583 in their alternative seismic approach and that the test case information is incorporated by reference into the license amendment. The licensee further stated how the proposed alternative seismic approach would be used in the categorization process and that the measures for assuring the quality and level of detail for the licensees proposed alternative seismic approach were adequate for the categorization of SSCs. Based on the above, the NRC staff finds that the requirements in 10 CFR 50.69(b)(2)(ii) for the proposed alternative seismic approach are met because the licensee adequately described the measures taken to assure the quality and level of detail of their proposed alternative seismic approach to evaluate the plant for the categorization of SSCs.

EPRI Report No. 3002017583 includes the results from case studies performed to determine the extent and type of unique HSS SSCs from seismic PRAs (SPRAs). In the LAR, the licensee indicated that aside from two exceptions (the site-specific LaSalle information and the configuration control checklist described in the LaSalle submittal), the licensee would follow the same alternative seismic approach in its proposed 10 CFR 50.69 categorization process as the approved LaSalle 10 CFR 50.69 amendment. Therefore, it is understood by the NRC staff that the technical criteria in EPRI Report No. 3002017583 are unchanged from its predecessor report EPRI Report No. 3002012988 (Reference 11), and that the case studies are applicable to Brunswick and are used in the alternative seismic approach. The NRC staffs review confirmed that the case studies in EPRI Report No. 3002017583 used by the licensee to support its proposed alternative seismic approach provided sufficient plant-specific evaluation of the applicability and differences for Brunswick as compared to the approach approved by the NRC (Reference 14) for LaSalle. The information presented in the LAR as well as in EPRI Report No. 3002017583 provided a sufficient description of, and basis for, acceptability of the evaluations to be conducted to satisfy 10 CFR 50.69(c)(1)(iv) for the alternative seismic approach because the NRC staff has reasonable confidence that evaluated low safety significant safety-related SSCs have sufficient safety margins maintained and that any potential increases in core damage frequency and large early release frequency resulting from the changes in SSC treatment are small. Therefore, the NRC staff finds that the requirements in 10 CFR 50.69(b)(2)(iv) are met for the proposed alternative seismic approach.

Based on the information provided in the LAR, the NRC staff finds it can base its approval of the licensees alternative seismic approach based on the reasons provided for its approval of LaSalles alternative seismic approach because (1) the differences between the licensees proposed alternative seismic approach and the alternative seismic approach previously approved are addressed; (2) there are no differences in the technical criteria used in EPRI Report No. 3002017583 and its predecessor, EPRI Report No. 3002012988, for use in this application; and (3) all references needed to support the NRC staffs finding on the proposed alternate seismic approach have been cited by the LAR.

3.1.1 Evaluation of Technical Acceptability of the PRAs Used for Case Studies Supporting the Proposed Alternative Seismic Approach In Section 3 of the LAR enclosure, the licensee included a discussion of the precedent including the case studies, mapping approach, and conclusions on the determination of unique HSS SSCs from the case studies which were used by the licensee to support its proposed alternative seismic approach. The licensee stated that Brunswick is using test case information from EPRI Report No. 3002017583 that demonstrated that seismic risk is adequately addressed for Tier 2 sites. Brunswick is a Tier 2 site by the results of additional qualitative assessments addressed in Section 3.1.2 below. The NRC staff reviewed and evaluated the technical acceptability of the PRAs used in the case studies for Plants A, C, and D in EPRI Report No. 3002017583 and the applicability of these test cases to Brunswick. The NRC staff also evaluated the peer review process, resolution of peer review findings, and key assumptions and sources of uncertainty for Plants A, C, and D.

Based on the above, the NRC staff finds the technical acceptability of PRAs used for the Plant A, C, and D case studies in EPRI Report No. 3002017583, the mapping approach used in those case studies, and the conclusions on the determination of unique HSS SSCs from the case studies in the precedent (Reference 14) are applicable to Brunswicks proposed plant specific alternative seismic approach. Therefore, the NRC staff concludes that the Plant A, C, and D PRAs were technically acceptable and applicable for use in support of the licensees proposed alternative seismic approach; the mapping of SSCs between the SPRA, the full power Internal

Events Probabilistic Risk Assessment (IEPRA), and, as applicable, the fire PRA (FPRA) for the Plant A, C, and D case studies. The licensees plant-specific evaluation is technically justifiable to support conclusions on the determination of unique HSS SSCs from SPRAs in Plant A, C, and D case studies in the EPRI Report No. 3002017583 and the licensees proposed alternative seismic approach is applicable to Brunswick.

3.1.2 Evaluation of the Criteria for the Proposed Alternative Seismic Approach In Section 3 of the LAR enclosure, the licensee stated that its basis for the classification of Brunswick as a Tier 2 plant is because the site ground motion response spectrum (GMRS) to safe shutdown earthquake (SSE) comparison is above the Tier 1 threshold in EPRI Report No. 3002017583, but not high enough that the NRC required the plant to perform a seismic PRA to respond to Recommendation 2.1 of the Near Term Task Force (NTTF) 10 CFR 50.54(f) letter (Reference 3). Additionally, consideration of the full range of the seismic hazard produces limited unique insights to the categorization process. The licensee explained that the basis for using the proposed alternative seismic approach is that the special seismic risk evaluation process for the proposed approach can identify the appropriate seismic insights to be considered with the other categorization insights by the Integrated Decision-making Panel (IDP) for the final HSS determinations.

The NRC staff notes that the licensees plant-specific evaluation is supported by its 10 CFR 50.54(f) response dated March 31, 2014 (Reference 9). The NRC staff reviewed the LAR and plant-specific evaluation and concludes that Brunswick meets the criteria for a Tier 2 plant.

Therefore, the licensees proposed criteria to determine the applicability and use of the proposed alternative seismic approach is acceptable.

3.1.3 Evaluation of Applicability of Criteria for this Application In Section 3 of the LAR enclosure, the licensee compared the GMRS from the reevaluated seismic hazard for Brunswick, developed and submitted by the licensee in response to NTTF Recommendation 2.1, against the sites design basis SSE in the March 31, 2014, response to a 10 CFR 50.54(f) letter associated with NTTF Recommendation 2.1 (Reference 9). This comparison was conducted to demonstrate that the site meets the criteria for application of the proposed alternative seismic approach. The licensee also stated in Section 3 of the LAR enclosure that the NRC concluded that the methodology used by the licensee in determining the GMRS was acceptable and that the GMRS determined by the licensee adequately characterized the reevaluated hazard for the Brunswick site. The NRC staffs review confirmed the licensees statements and the comparison of the GMRS from the reevaluated seismic hazard against the SSE. Based on its review, the NRC staff finds that the licensees seismic hazard meets the criteria for the proposed alternative seismic approach.

3.1.4 Evaluation of the Implementation of Conclusions from the Case Studies The categorization conclusions from the EPRI Report No. 3002017583 case studies indicated that seismic specific failure modes resulted in HSS categorization uniquely from SPRAs.

Therefore, seismic specific failure modes, such as correlated failures, relay chatter, and passive component structural failure modes, can influence the categorization process. The NRC staff reviewed the proposed alternative seismic approach to evaluate whether the categorization-related conclusions from EPRI Report No. 3002017583 were appropriately included and implemented.

In Section 3 of the LAR enclosure, the licensee discussed the proposed alternative seismic approach. The licensee stated that the proposed categorization approach for seismic hazards includes qualitative consideration of the mitigation capabilities of SSCs during seismically induced events and seismic failure modes, based on non-PRA based qualitative assessments in conjunction with any seismic insights provided by the Brunswick PRA. Section 3 of the LAR enclosure also provides additional information on the prior evaluations performed for Brunswick.

In Section 3 the LAR enclosure, the licensee stated that its plant-specific evaluation follows the same categorization approach for Tier 2 seismic risk as the proposed alternative seismic approach from the precedent previously reviewed and approved by the NRC staff (Reference 14). The NRC staffs review of the licensees proposed alternative seismic approach determined that the LaSalle approach is applicable to the proposed alternative seismic approach and the plant specific evaluation on the implementation of the alternative seismic approach is acceptable. Based on its review of the licensees proposed alternative seismic approach, in conjunction with the requirements in 10 CFR 50.69 and the corresponding statements of consideration (Reference 15), the NRC staff finds that the proposed alternative seismic approach provides reasonable confidence in the evaluations required by 10 CFR 50.69(c)(1)(ii) as well as 10 CFR 50.69(c)(1)(iv) because:

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

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

The insights presented to the IDP include potentially important seismically induced failure modes, as well as mitigation capabilities of SSCs during seismically induced design basis and severe accident events consistent with the conclusions on the determination of unique HSS SSCs from SPRAs in EPRI Report No. 3002017583. The insights will use prior plant specific seismic evaluations, and therefore, in conjunction with performance monitoring for the proposed alternative seismic approach, reasonably reflect the current plant configuration. Further, the recommendation for categorizing civil structures in the alternative seismic approach provides appropriate consideration of such failures from a seismic event.

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

3.1.5 Consideration of Changes to Seismic Hazard An important input to the NRC staffs evaluation of the proposed alternative seismic approach is the current knowledge of the seismic hazard at the plant. The possibility exists for the seismic hazard at the site to increase or decrease such that the criteria for use of the proposed alternative seismic approach are challenged. In such a situation, the categorization process may

be impacted from a seismic risk perspective either solely due to the seismic risk or by the integrated importance measure determination.

In Section 3 of the LAR enclosure, the licensee stated that if the Brunswick seismic hazard changes from medium risk (i.e., Tier 2) at some future time and the feedback process determines that a process different from the proposed alternative seismic approach is warranted for seismic risk consideration in categorization under 10 CFR 50.69, prior NRC approval, pursuant to 10 CFR 50.90, will be requested. The licensee also stated that aspects of their feedback and review process remain as stated in Reference 2, which states that periodic reviews are performed to ensure continued categorization validity and acceptable performance for those SSCs that have been categorized.

The NRC staff finds that the consideration of changes to the seismic hazard in the licensees proposed alternative seismic approach is consistent with the NRC staffs approval (Reference 14) of the LaSalle LAR. Therefore, the NRC staffs evaluation of the proposed changes to the seismic hazard against the requirements in 10 CFR 50.69(e)(1), 10 CFR 50.69(e)(3), and 10 CFR 50.69(d)(2)(ii), as well as the resulting conclusion on consideration of changes to the seismic hazard for LaSalle is applicable to this licensees proposed alternative seismic approach. Consequently, the NRC staff finds that the consideration of changes to the seismic hazard at Brunswick is acceptable because: (1) the criteria for use of the proposed alternative seismic approach are clear and traceable, (2) the proposed alternative seismic approach includes periodic reconsideration of the seismic hazard as new information becomes available, (3) the proposed alternative seismic approach satisfies the requirements in 10 CFR 50.69 discussed above, and (4) the licensee has included a proposed license condition in the LAR to require NRC approval for a change to the specified seismic categorization approach.

3.1.6 Monitoring of Inputs to and Outcome of Proposed Alternative Seismic Approach In Section 3 of the LAR enclosure, the licensee described its feedback and adjustment process.

The licensee provided a description of its periodic review process to review the impact of plant changes and a list of items included in the design effects and consideration review. Further, the licensee cited precedent for its proposed alternative seismic approach (Reference 14) with the exception that a specific configuration control checklist was not developed for 10 CFR 50.69 reviews.

The NRC staff found that consideration of the feedback and adjustment process in the licensees proposed alternative seismic approach is acceptable. The NRC staff finds that (1) the licensees programs provide reasonable assurance that the existing seismic capacity of LSS components would not be significantly impacted and (2) the monitoring and configuration control program ensures that potential degradation of seismic capacity would be detected and addressed before significantly impacting the plant risk profile. Therefore, the NRC staff finds that the potential impact of the seismic hazard on the categorization of RISC-3 and RISC-4 SSCs is maintained acceptably low and the requirements in 10 CFR 50.69(c)(1)(iv) are met for the proposed alternative seismic approach.

3.2 PRA Implementation Items In Section 2.3 of the LAR enclosure, the licensee indicated that it was necessary to disposition the PRA implementation items that are referenced in its existing 10 CFR 50.69 license condition before assessing the technical adequacy of the alternative seismic approach. The existing license condition states, in part, that:

Duke Energy will complete the implementation items listed in Attachment 1 of Duke letter to the NRC dated November 24, 2020, prior to implementation of 10 CFR 50.69 in accordance with the categorization process described above.

The implementation items in Attachment 1 of Reference 13, are as follows:

BSEP [Brunswick] will enhance the existing temporary passive interior door flood barriers (i.e., Cliff Edge Barriers) to provide protection for SSCs relied upon to protect KSFs [key safety functions] in the Reactor Building (interior to external doors D-2 and D-3), Diesel Generator [DG] Building (Exterior to external door D-6) and Control Building [CB] (At CB entrance doors interior to TB [Turbine Building] external doors D-19 and D-22) from the CE [combined effects] Storm Surge event. The top elevation of these enhanced temporary passive barriers will be at elevation 27.5 ft NGVD29 [National Geodetic Vertical Datum of 1929] to achieve minimum APM [available physical margin] = 0.7 ft.

For Control Building door locations, 2-CTB-DR-EL023-102 and 1-CTB-DR-EL023-105, BSEP modifications will replace the temporary barriers with permanent water-tight doors providing permanent passive flood protection.

For the Service Water Building at the locations near Doors D-13 and D-14, BSEP will modify the existing security delay gate doors to shield ventilation openings from waves.

For the DG Building north external personnel watertight door D-4, BSEP will provide a new debris barrier that protects this door subjected to debris loading associated with the CE Storm Surge event.

BSEP will confirm all barriers conform to the requirements for flood protection features specified in NEI 16-05, including Appendix B, and the time requirements of 0AI-68.

BSEP will provide a new debris barrier that protects all penetrations on the FOTC

[Fuel Oil Tank Chamber] Roof from debris loading associated with the CE Storm Surge event.

In Section 1 the LAR enclosure, the licensee confirmed that all the implementation items have been completed.

As discussed in the SE within Reference 4, the completion of the implementation items will provide adequate protection to the site during a CE storm surge event and that the adequacy of these protective features was evaluated as part of the staffs assessment of the licensees focused evaluation so these modifications by the licensee do not change or impact the bases for the safety conclusions made by the NRC staff. The NRC staff may choose to examine, through an onsite audit or future inspections, the closure of the implementation items with the expectation that any concerns regarding adequate completion of the implementation items would be tracked and dispositioned appropriately under the licensees corrective action program and could be subject to appropriate NRC enforcement action. Based on this consideration, as well as the licensees affirmation that the implementation items have been completed and the information provided by the licensee in the LAR, the NRC staff finds that the license condition requiring completion of implementation items can be removed.

3.3 Technical Evaluation Summary Based on the above, the NRC staff finds the licensees non-PRA methods for assessing risk for seismic hazards, a deviation from NEI 00-04, acceptable, and determines that the licensees proposed 10 CFR 50.69 program, with the proposed license condition, continues to meet the requirements in 10 CFR 50.69.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the State of North Carolina official was notified of the proposed issuance of the amendment on April 22, 2024. The State of North Carolina official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes the requirements with respect to installation or use of a facilitys components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration in the Federal Register on October 31, 2023 (88 FR 74530),

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 to 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.

Krakuszeski, J. A., Duke Energy Progress, LLC, letter to U.S. Nuclear Regulatory Commission, License Amendment Request to Revise the 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, Categorization Process to Reflect an Alternative Seismic Approach, August 17, 2023 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML23229A456).

2.

Duke Energy letter to U.S. Nuclear Regulatory Commission, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors, January 10, 2018 (ML18010A344).

3.

U.S. Nuclear Regulatory Commission letter to all Power Reactor Licensees, Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding

Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, March 12, 2012 (ML12053A340).

4.

Hon, A., U.S. Nuclear Regulatory Commission, letter to John A. Krakuszeski, Duke Energy Progress, LLC, Brunswick Steam Electric Plant, Units 1 and 2 - Issuance of Amendment Nos. 305 and 333 to Revise License Conditions to Modify Approved 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors Categorization Process (EPID L-2020-LLA-0152), April 30, 2021 (ML21067A224).

5.

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

6.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, December 2020 (ML20238B871).

7.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, January 2018 (ML17317A256).

8.

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

9.

Hamrick, G. T., Duke Energy Progress, Inc., letter to U.S. Nuclear Regulatory Commission, Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, March 31, 2014 (ML14106A461).

10.

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

11.

Electric Power Research Institute Report No. 3002012988, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization, July 2018 (ML21067A143; non-publicly available).

12.

Electric Power Research Institute Report No. 3002017583, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization, February 2020 (ML21082A170).

13.

Ratliff, J., Duke Energy Progress, LLC, letter to U.S. Nuclear Regulatory Commission, Response to Request for Additional Information (RAI) for License Amendment Request to Modify Approved 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors Categorization Process, November 24, 2020 (ML20329A466).

14.

Vaidya, B. K., U.S. Nuclear Regulatory Commission, letter to Rhoades, D. P., Exelon Generation Company, LLC, LaSalle County Station, Unit Nos. 1 and 2 - Issuance of Amendment Nos. 249 and 235 Related to Application to Adopt 10 CFR 50.69, Risk-

informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors (EPID L-2020-LLA-0017), May 27, 2021 (ML21082A422).

15.

U.S. Nuclear Regulatory Commission, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, Federal Register Vol. 69, No. 224, November 22, 2004.

Principal Contributors:

K. Tetter, NRR S. Alferink, NRR Date: June 5, 2024

ML24108A070 OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DRA/APLC/BC NAME LHaeg ABaxter SVasavada DATE 4/16/2024 4/22/2024 3/20/2024 OFFICE OGC - NLO NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM NAME RSiegman DWrona LHaeg DATE 5/15/2024 5/31/2024 6/5/2024