L-2024-120, LAR 301 - Revise Renewed Facility Operating Licenses to Adopt an Alternative Approach for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization

LAR 301 - Revise Renewed Facility Operating Licenses to Adopt an Alternative Approach for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization

ML24282A760
Person / Time
Site:	Point Beach
Issue date:	10/08/2024
From:	Mack K Point Beach
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
L-2024-120
Download: ML24282A760 (1)

Text

NEXTeraM ENERGY~

~

Attn: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 RE:

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 Renewed Facility Operating Licenses DPR-24 and DPR-27 October 08, 2024 L-2024-120 10 CFR 50.90 10 CFR 50.69 LAR 301 - Revise Renewed Facility Operating Licenses to Adopt an Alternative Approach for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), NextEra Energy Point Beach, LLC (NextEra) hereby requests an amendment to Renewed Facility Operating License (RFOL) Nos. DPR-24 and DPR-27 for Point Beach Nuclear Plant (PBNP), Units 1 and 2, respectively.

The proposed change would revise the license condition associated with the adoption of 10 CFR 50.69, "Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors," that was added to the PBNP RFOL upon issuance of Amendments 262 and 265. Specifically, the proposed change would allow for an alternative to the approaches provided in Nuclear Energy Institute (NEI) 00-04, "1 O CFR 50.69 SSC Categorization Guideline," Revision 0, for evaluating the impact of the seismic hazard in the 10 CFR 50.69 categorization process. Additionally, the proposed license amendment would make editorial corrections to Unit 1 License Condition M and Unit 2 License Condition L, "Additional Conditions," Functions 5 and 6 on Technical Specification (TS) Table 3.3.1-1, and TS Section 5.5.18.h.

The proposed alternative seismic approach is described in Electric Power Research Institute (EPRI) 3002017583, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," and is a risk-informed, graded approach that has demonstrated categorization insights equivalent to a seismic probabilistic risk assessment.

The Enclosure provides a description and assessment of the proposed changes. Attachment 1 provides the license conditions from the PBNP RFOLs marked to show the proposed changes. provides the existing TS pages marked to show the proposed changes. The TS Bases are not affected by the proposed changes.

NextEra has determined that the proposed change does not involve a significant hazards consideration pursuant to 10 CFR 50.92(c), and that there are no significant environmental impacts associated with the change. The Point Beach Onsite Review Group has reviewed the enclosed amendment request.

NextEra Energy Point Beach, LLC 6610 Nuclear Road, Two Rivers, WI 54241

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Page 2 of 2 In accordance with 1 O CFR 50.91 (b)(1 ), a copy of this license amendment request is being provided to the designated State of Wisconsin official.

NextEra requests approval of the proposed license amendment within one year of completion of the Nuclear Regulatory Commission's acceptance review. Once approved, the amendment shall be implemented within 60 days.

There are no regulatory commitments made in this submittal.

Should you have any questions regarding this submission, please contact Ms. Maribel Valdez, Fleet Licensing Manager, at 561-904-5164.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on the 81h day of October 2024.

Sincerely, Kenn~-¥ Director, Regulatory Affairs

Enclosure:

Description and Assessment Attachments: 1. Mark-ups of the PBNP Units 1 and 2 Renewed Facility Operating Licenses

2. Proposed Technical Specification Changes (Mark-ups) cc:

USNRC Regional Administrator, Region Ill Project Manager, USNRC, Point Beach Nuclear Plant Resident Inspector, USNRC, Point Beach Nuclear Plant Public Service Commission of Wisconsin

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 ENCLOSURE DESCRIPTION AND ASSESSMENT 1

SUMMARY

DESCRIPTION L-2024-120 Enclosure Page 1 of 29 NextEra Energy Point Beach, LLC (NextEra) hereby requests an amendment to Renewed Facility Operating License (RFOL) Nos. DPR-24 and DPR-27 for the Point Beach Nuclear Plant (PBNP), Units 1 and 2, respectively.

The proposed change would revise the license condition associated with the adoption 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," that was added to the PBNP RFOL upon the issuance of Amendments 262 and 265 (Reference 1 ).

Specifically, the proposed change would allow for an alternative to the approaches provided in Nuclear Energy Institute (NEI) 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision 0 (Reference 2), for evaluating the impact of the seismic hazard in the 10 CFR 50.69 categorization process.

The proposed alternative seismic approach for Tier 2 plants (e.g., PBNP) is described in Electric Power Research Institute (EPRI) 3002017583, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," (Reference 3) and is a risk-informed, graded approach that has demonstrated categorization insights equivalent to a seismic probabilistic risk assessment (PRA). For Tier 2 plants such as PBNP, the EPRI approach relies on the insights gained from the seismic PRAs examined in Reference 3 and plant specific insights considering seismic correlation effects and seismic interactions.

Furthermore, the license condition issued as part of Amendments 262 and 265 required that certain items be completed prior to implementing 10 CFR 50.69 at PBNP. NextEra has completed those implementation items (Item A in Attachment 1 of Reference 4 and of Reference 5); therefore, it is proposed that these items be removed from the license condition. Additionally, the proposed license amendment would make editorial corrections to Unit 1 License Condition M and Unit 2 License Condition L, "Additional Conditions," Functions 5 and 6 on Technical Specification (TS) Table 3.3.1-1, and TS Section 5.5.18.h.

2 DETAILED DESCRIPTION 2.1 CURRENT REGULATORY REQUIREMENTS The provisions of 10 CFR 50.69 allow adjustment of the scope of structures, systems, and components (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.

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 2 of 29 The regulation in 10 CFR 50.69 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.

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.

In Reference 1, the Nuclear Regulatory Commission (NRC) issued Amendments 262 and 265 to the RFOL for PBNP, which added a new license condition to allow for the implementation of the provisions of 1 O CFR 50.69. Consistent with the guidance in NEI 00-04, as endorsed by Regulatory Guide (RG) 1.201, Revision 1, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," (Reference

6) the existing PBNP categorization process uses probabilistic risk assessment (PRA) models to assess risk from internal events, including internal flooding, and internal fire. PBNP currently uses the Seismic Safe Shutdown Equipment List (SSEL) from the seismic margin analysis (SMA) to assess seismic risk, the individual plant examination of external events (IPEEE) screening process to assess other external hazards, and a qualitative defense-in-depth shutdown model to assess shutdown risk.

Regarding the subject license amendment request and the assessment of seismic risk in the 10 CFR 50.69 categorization process, PBNP currently uses the SMA screening method.

PBNP currently follows the approach in Reference 2 using the SSEL to identify credited equipment as HSS, regardless of the equipment's capacity, frequency of challenge or level of functional diversity. Consistent with Reference 2, the PBNP 10 CFR 50.69 categorization process considers all components on the SSEL as HSS based on seismic risk. All components not listed in the SSEL are considered preliminary LSS with respect to seismic risk.

2.2 REASON FOR PROPOSED LICENSE CONDITION CHANGE The NRC staff expects that licensees proposing to use non-PRA approaches in the 1 O CFR 50.69 categorization process provide a basis in the submittal for why the approach and the accompanying method employed to assign safety significance to SSCs is technically adequate (Reference 6). The guidance further states that as part of the NRC's review and approval of an application requesting to implement 1 O CFR 50.69, the NRC staff intends to impose a license condition that will explicitly address the scope of the PRA and non-PRA methods used in the categorization approach. To that end, the NRC imposed a license condition on PBNP (Reference 1) that requires use of the SMA to evaluate seismic risk in the 10 CFR 50.69 categorization process.

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 3 of 29 Since NextEra desires to change the categorization approach specified in the PBNP RFOL with respect to the assessment of seismic risk (i.e., to allow use of the EPRI alternative seismic approach in Reference 3), NRC approval of the new approach in the categorization process must be requested pursuant to 10 CFR 50.90, in accordance with the license condition that was added upon issuance of PBNP Amendment Nos. 262 and 265.

This license amendment request follows the same categorization approach for Tier 2 seismic risk as approved for LaSalle County Station, Units 1 and 2 (References 7, 8, 9, 10, and 11),

which is further discussed in Section 3.1 of this submittal.

2.3 DESCRIPTION

OF THE PROPOSED LICENSE CONDITION CHANGE NextEra proposes to revise the 10 CFR 50.69 license condition, that was added to the PBNP RFOL by Amendment 262 (Unit 1) and Amendment 265 (Unit 2). The following changes are proposed to the license condition; bracketed text is used to identify the unit-specific license amendment number(s) and the issuance date of the approved license amendment, when issued.

NextEra Energy Point Beach is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using:

Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (AN0-2) 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 License Amendment No. [262/265] dated November 26, 2018.

In addition, NextEra Energy Point Beach is approved to implement 10 CFR 50.69 using the alternative seismic approach described in NextEra Energy Point Beach letter L-2024-120, dated September X, 2024, for categorization of RISC-1, RISC-2, RISC 3, and RISC-4 SSCs, as specified in License Amendment No. [XXX] dated

[DATE].

Prior to implementation of the provisions of 10 CFR 50.69, NextEra Energy Point Beach shall complete the items below:

a. Item A in Attachment 1, List of Categorization Prerequisites, to NextEra Energy Point Beach letter NRG 2017 0043, "License Amendment Request 287, Application to Adopt 1 O CFR 50.69, 'Risk Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Plants,' " dated August 31, 2017; and

b. Attachment 1, Point Beach 10 CFR 50.69 PRA Implementation Items, in

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 4 of 29 NextEra Energy Point Beach letter NRG 2018 0044, "Supplement to Response to Request for /\\dditional Information Regarding License Amendment Request 287, Application to Adopt 10 CFR 50.69, 'Risk informed Categorization and Treatment of Structures, System, and Components (SSCs) for Nuclear Power Plants,'" dated September 28, 2018.

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 second paragraph (underlined text) provides the proposed requirements for the alternate seismic categorization process, which is the subject of this request.

The implementation items were completed prior to the implementation of the 1 O CFR 50.69 categorization process at PBNP, as required by the license condition. Therefore, it is proposed that the associated text be deleted (stricken through text).

The proposed license condition markups are provided in Attachment 1 to this submittal.

2.4 EDITORIAL CORRECTION TO UNIT 1 LICENSE CONDITION M AND UNIT 2 LICENSE CONDITION L, "ADDITIONAL CONDITIONS" Unit 1 License Condition M states, "The additional conditions contained in Appendix C, as revised through Amendment No. 241, are hereby incorporated into this license."

Unit 1 License Condition M would be revised to state, 'The additional conditions contained in Appendix C, as revised through Amendment No. 258, are hereby incorporated into this license."

Unit 2 License Condition L states, "The additional conditions contained in Appendix C, as revised through Amendment No. 245, are hereby incorporated into this license."

Unit 2 License Condition L would be revised to state, "The additional conditions contained in Appendix C, as revised through Amendment No. 262, are hereby incorporated into this license."

2.5 REASON FOR EDITORIAL CORRECTION TO UNIT 1 LICENSE CONDITION M AND UNIT 2 LICENSE CONDITION L, "ADDITIONAL CONDITIONS" Reference 51 requested changes to completed license conditions in the RFOL Nos. DPR-24 and DPR-27 for Point Beach Units 1 and 2, respectively, Appendix C, Additional Conditions Operating License DPR-24, and Appendix C, Additional Conditions Operating License DPR-

27. Reference 51 omitted inclusion of corresponding changes to Unit 1 License Condition M and Unit 2 License Condition L to reflect NRC issuance of Amendments approving LAR 280.

Reference 52 issued Amendment Nos. 258 and 262 to RFOL Nos. DPR-24 and DPR-27 for Point Beach Units 1 and 2, respectively, which included Appendix C, Additional Conditions Operating License DPR-24 being issued through Amendment 258 and Appendix C, Additional Conditions Operating License DPR-27 being issued through Amendment 262. This resulted in

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 5 of 29 an inadvertent inconsistency between Unit 1 License Condition M and Additional Conditions Operating License DPR-24 and between Unit 2 License Condition L and Appendix C, Additional Conditions Operating License DPR-27.

2.6 EDITORIAL CORRECTION TO FUNCTIONS 5 AND 6 ON TS TABLE 3.3.1-1 The Allowable Value for Function 5 on TS Table 3.3.1-1 states, "Refer to Note 1 (Page 3.3.1-18)."

The Allowable Value for Function 5 on TS Table 3.3.1-1 would be revised to state, "Refer to Note 1 (Page 3.3.1-20)."

The Allowable Value for Function 6 on TS Table 3.3.1-1 states, "Refer to Note 2 (Page 3.3.1-19)."

The Allowable Value for Function 6 on TS Table 3.3.1-1 would be revised to state, "Refer to Note 2 (Page 3.3.1-21)."

2.7 REASON FOR EDITORIAL CORRECTION TO FUNCTIONS 5 AND 6 ON TS TABLE 3.3.1-1 Reference 53 requested changes to TS to permit use of Risk Informed Completion Times in accordance with TSTF-505, Revision 2, "Provide Risk Informed Extended Completion Times

- RITSTF Initiative 4b". Reference 53 omitted inclusion of corresponding changes to the Allowable Value for Functions 5 and 6 on TS Table 3.3.1-1 because of failure to account for additional pages added to TS Section 3.3.1 because of changes to Actions for Limiting Condition for Operation 3.3.1. Reference 54 issued Amendment Nos. 271 and 273 to RFOL Nos. DPR-24 and DPR-27 for Point Beach Units 1 and 2, respectively. This resulted in an inadvertent inconsistency between the page numbers listed for Notes 1 and 2 in the Allowable Value column for Functions 5 and 6 on TS Table 3.3.1-1 and the actual TS Section 3.3.1 pages containing Notes 1 and 2.

2.8 EDITORIAL CORRECTION TO TS SECTION 5.5.18.h TS Section 5.5.18.h states, "Portable smoke ejection equipment per the Fire Protection Evaluation Report and Safe Shutdown Analysis Report to address a potential smoke challenge."

TS Section 5.5.18.h would be revised to state, "Portable smoke ejection equipment per the Fire Protection Program Design Document to address a potential smoke challenge."

2.9 REASON FOR EDITORIAL CORRECTION TO TS SECTION 5.5.18.h Reference 55 requested changes to License Condition F in the RFOL Nos. DPR-24 and DPR-27 for Point Beach Units 1 and 2, respectively, and to TS Section 5.4.1. Reference 55 omitted inclusion of corresponding changes to TS Section 5.5.18.h. Reference 56 issued Amendment Nos. 256 and 260 to RFOL Nos. DPR-24 and DPR-27 for Point Beach Units 1 and 2,

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 6 of 29 respectively. This resulted in an inadvertent inconsistency between License Condition F, which refers to the approved fire protection program that complies with 10 CFR 50.48(a) and 1 O CFR 50.48(c) and TS Section 5.5.18.h that refers to the Fire Protection Evaluation Report and Safe Shutdown Analysis Report, which were part of the Point Beach licensing basis per 10 CFR 50.48(b) and 10 CFR 50, Appendix R.

3 TECHNICAL EVALUATION 1 O CFR 50.69 specifies the information to be provided by a licensee requesting adoption of the regulation. This request conforms to the requirements of 10 CFR 50.69(b)(2), which states:

A licensee voluntarily choosing to implement this section shall submit an application for license amendment under§ 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 probabilistic risk assessment (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§ 50.69(c)(1)(i).

(iv) A description of, and basis for acceptability of, the evaluations to be conducted to satisfy§ 50.69(c)(1)(iv). The evaluations must include the effects of common cause interaction susceptibility, and the potential impacts from known degradation mechanisms for both active and passive functions, and address internally and externally initiated events and plant operating modes (e.g., full power and shutdown conditions).

The above information was previously provided to the NRC with the PBNP application to allow implementation of the provisions of 10 CFR 50.69 (Reference 4). The PBNP 10 CFR 50.69 categorization process (overall process, including active and passive categorization elements) has been reviewed and approved by the NRC (Reference 1 ).

In its review and approval of that application, the NRC staff reviewed the technical adequacy of the PBNP internal events, including internal flooding, at power and fire PRA models. The NRC concluded that the quality and level of detail of the PRA models, with the completion of the implementation items specified in the license condition, was sufficient to support 10 CFR 50.69 categorization of SSCs (Reference 1). NextEra completed the implementation items (as required by the license condition) and implemented the PBNP 10 CFR 50.69 program on December 14, 2018. The NRC staff later reviewed the PRA models' peer review history as part of an application to adopt the risk-informed completion time (RICT) program (Reference 12). The NRC determined that NextEra adequately applied the guidance to

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 7 of 29 establish PRA technical adequacy and that the PBNP internal events, including internal flooding, and fire PRA models were acceptable for the RICT program (Reference 13).

The PBNP categorization process currently addresses seismic risk using SMA, following the process defined in NEI 00-04 (Reference 2) and endorsed in RG 1.201 (Reference 6). The purpose of this license amendment request is to allow, within the approved PBNP 50.69 program, the use of the alternative seismic method documented in EPRI Report 3002017583 for Tier 2 sites and as described in this request, in addition to use of the SMA, to assess seismic risk. Therefore, the remainder of this technical evaluation is focused on describing and evaluating the categorization process considering the addition of the EPRI alternative seismic method for Tier 2 sites, as described in Reference 3 (also referred to as "EPRI 3002017583" throughout the remainder of this enclosure).

3.1 CATEGORIZATION PROCESS DESCRIPTION (10 CFR 50.69(b)(2)(i))

For the proposed change, the process to categorize each system will continue to be consistent with the guidance in NEI 00-04, as endorsed by RG 1.201, except for evaluation of the impact of the seismic hazard, which will use EPRI 3002017583 for Tier 2 sites. The inclusion of additional steps to address seismic considerations will ensure compliance with 10 CFR 50.69(c)(1).

Categorization Process Using Alternative Seismic Method for Tier 2 Sites NextEra proposes that the alternate Seismic Tier 2 categorization process may be implemented for any PBNP system that was previously categorized or for systems that will be categorized. However, any system that has been previously categorized is not required to be re-categorized with the alternate Seismic Tier 2 categorization process. The processes identified in the existing NRG-approved 10 CFR 50.69 license condition may continue to be used. With the proposed change, PBNP will use a single approach for a given system categorization (e.g., either SMA or alternate seismic approach described herein).

Table 1 is an update to the table that was provided in NextEra's response to an NRC request for additional information (RAI) (Reference 14), which supplemented the original submittal to adopt 10 CFR 50.69 (Reference 4). The proposed changes are marked by bold, italic font to highlight the proposed alternative approach for evaluating the seismic hazard.

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 Table 1: Categorization Evaluation Summary Categorization Step -

IDP Change Element Evaluation Level NEI 00-04 Section HSS to LSS Internal Events Base Case -

Not Allowed Section 5.1 Fire, Seismic and Other Allowable Risk (PRA External Events Base Case Modeled)

Component PRA Sensitivity Studies Allowable Integral PRA Assessment -

Not Allowed Section 5.6 Fire and Other External Component Not Allowed Hazards Risk (Non-Seismic - SMA Process Component Not Allowed modeled)

Seismic - Alternative Tier 2 Approach Function/Component Allowed 2 Shutdown - Section 5.5 Function/Component Not Allowed Defense-Core Damage - Section 6.1 Function/Component Not Allowed in-Depth Containment - Section 6.2 Component Not Allowed Qualitative Considerations - Section 9.2 Function Allowable 1 Criteria Passive Passive - Section 4 Segment/Component Not Allowed Notes:

1 Note 1 remains as it was in Table 1 of Reference 14.

L-2024-120 Enclosure Page 8 of 29 Drives Associated Functions Yes No No Yes No No No No Yes Yes N/A No 2 Integrated Decision-Making Panel (IDP) consideration of seismic insights can also result in an LSS to HSS determination.

10 CFR 50.69(c)(1) requires the use of PRA to assess risk from internal events. For other risk hazards, such as seismic, 10 CFR 50.69(b)(2) allows, and NEI 00-04 (Reference 2) summarizes, the use of other methods for determining SSC functional importance in the absence of a quantifiable PRA (such as SMA or IPEEE Screening) as part of an integrated, systematic process. For the PBNP seismic hazard assessment, NextEra proposes to use a risk-informed graded approach that meets the requirements of 10 CFR 50.69(b)(2) as an alternative to those listed in NEI 00-04 sections 1.5 and 5.3. This approach is specified in

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 9 of 29 EPRI 3002017583 with the markups provided in Attachment 2 of References 9 and 10 and includes additional considerations that are discussed in this section.

Note: The discussion below pertaining to EPRI 3002017583 includes the markups provided in Attachment 2 of References 9 and 10.

EPRI 3002017583 (Reference 3) is an update to EPRI 3002012988, "Alternative Approaches for Addressing Seismic Risk in 1 O CFR 50.69 Risk-Informed Categorization,"

July 2018 (Reference 15) which was referenced in the NRG-issued amendment and Safety Evaluation for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, to implement 10 CFR 50.69, as noted in Reference 16. The technical criteria in EPRI 3002017583 are unchanged from its predecessor report EPRI 3002012988.

This license amendment request incorporates by reference the Clinton Power Station, Unit 1 response to RAI 'DRA/APLC RAI 03 - Alternate Seismic Approach' included in the letter dated November 24, 2020 (Reference 17), in particular, the response to the question regarding the differences between the initial EPRI report (3002012988) and EPRI 3002017583.

The proposed categorization approach for PBNP is a risk-informed graded approach that is demonstrated to produce categorization insights equivalent to a seismic PRA (SPRA). This approach relies on the insights gained from the SPRAs examined in EPRI 3002017583 and plant specific insights considering seismic correlation effects and seismic interactions.

Following the criteria in EPRI 3002017583, PBNP is considered a Tier 2 site because the site Ground Motion Response Spectrum (GMRS) to SSE (Safe Shutdown Earthquake) comparison is above the Tier 1 threshold, but not high enough that the NRC required the plant to perform an SPRA to respond to Recommendation 2.1 of the Near-Term Task Force 50.54(f) letter (Reference 18). EPRI 3002017583 also demonstrates that seismic risk is adequately addressed for Tier 2 sites by the results of additional qualitative assessments discussed in this section and existing elements of the 10 CFR 50.69 categorization process specified in NEI 00-04.

The trial studies in EPRI 3002017583, as amended by the RAI responses and NRC issued amendments (References 19, 20, 21, 22, 23, 24, 25, 26, and 27) demonstrate that seismic categorization insights are overlaid by other risk insights even at plants where the GMRS is far beyond the seismic design basis. Therefore, the basis for the Tier 2 classification and resulting criteria is that consideration of the full range of the seismic hazard produces limited unique insights to the categorization process. That is the basis for the following statements in Table 4-1 of EPRI 3002017583.

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

At sites with moderate seismic demands (i.e., Tier 2 range) such as PBNP, there is no need to perform more detailed evaluations to demonstrate the inherent seismic capacities

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 10 of 29 documented in industry sources such as Reference 28. Tier 2 seismic demand sites have a lower likelihood of seismically induced failures and less challenges to plant systems than trial study plants. This, therefore, provides the technical basis for allowing use of a graded approach for addressing seismic hazards at PBNP.

Test cases described in Section 3 of EPRI 3002017583, as amended by their RAI responses and NRC issued amendments (References 19, 20, 21, 22, 23, 24, 25, 26, and 27) demonstrated that there are very few, if any, SSCs that would be designated HSS for seismic unique reasons. The test cases identified that the unique seismic insights were typically associated with seismically correlated failures and led to unique HSS SSCs. While it would be unusual even for moderate hazard plants to exhibit any unique seismic insights, it is prudent and recommended by EPRI 3002017583 to perform additional evaluations to identify the conditions where correlated failures and seismic interactions may occur and determine their impact in the 1 O CFR 50.69 categorization process. The special sensitivity study recommended in EPRI 3002017583 uses common cause failures, similar to the approach taken in a full power internal events (FPIE) PRA and can identify the appropriate seismic insights to be considered by the IDP along with the other categorization insights for the final HSS determinations.

The test case information from EPRI 3002017583, developed by other licensees, including Case Study A (Reference 29), Case Study C (Reference 30), and Case Study D (Reference 31), as well as RAI responses and amendments (References 19, 20, 21, 22, 23, 24, 25, 26, and 27) clarify aspects of these case studies and provide additional supporting bases for this application. Therefore, these case studies, RAI responses, and amendments are incorporated by reference into this amendment request.

Basis for PBNP being a Tier 2 Plant As defined in EPRI 3002017583, PBNP meets the Tier 2 criteria for a "Moderate Seismic Hazard / Moderate Seismic Margin" site. The Tier 2 criteria are as follows:

Tier 2: Plants where the GMRS to SSE comparison between 1.0 Hz and 10 Hz is greater than in Tier 1 but not high enough to be treated as Tier 3. At these sites, the unique seismic categorization insights are expected to be limited.

Note: EPRI 3002017583 applies to the Tier 2 sites in its entirety except for Sections 2.2 (Tier 1 sites) and 2.4 (Tier 3 sites).

For comparison, Tier 1 plants are defined as having a GMRS peak acceleration at or below approximately 0.2g or where the GMRS is below or approximately equal to the SSE between 1.0 Hz and 1 O Hz. Tier 3 plants are defined where the GMRS to SSE comparison between 1.0 Hz and 10 Hz is high enough that the NRC required the plant to perform an SPRA to respond to the Fukushima 10 CFR 50.54(f) letter (Reference 18).

The NRC issued its final determination of licensee SPRAs in a letter dated October 27, 2015 (Reference 32). The letter informed power reactor licensees of the remaining seismic evaluations to be performed and specifically informed those licensees that would perform an SPRA. In the letter, NRC stated:

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 11 o'f 29 If the seismic hazard exceedance, peak of the spectral acceleration, and the general estimation of the [seismic core damage frequency] SCDF were judged to be not significant, then the NRC staff concluded that a SPRA is not necessary for NRC's 50.54(f) letter-related regulatory decisions. Based on this additional assessment, the NRC staff has determined that SPRA are not warranted for 13 sites listed in Table 1 a in Enclosure 1.

Note 3 of Table 1 a identifies PBNP as a site where an SPRA or SMA were no longer expected.

As shown in Figure 1, comparing the PBNP GMRS (derived from the seismic hazard) to the SSE (seismic design basis capability) between 1.0 Hz and 10 Hz, the GMRS exceeds the SSE above approximately 3 Hz. Note that the figure also shows the GMRS as determined by the NRC, when the staff reviewed information related to the reevaluated seismic hazard for PBNP (References 33 and 34). NextEra's GMRS curve for PBNP ("Licensee" in Figure 1) is generally similar to that determined by the NRC staff and both GMRS curves exceed the SSE in a portion of the response spectrum between 1.0 and 10 Hz.

Therefore, the PBNP seismic hazard meets the criteria for Tier 2 from EPRI 3002017583. The basis for PBNP being classified as a Tier 2 site will be documented and presented to the IDP for each system that is categorized.

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Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 12 of 29 The following paragraphs describe additional background and the process to be utilized for the graded approach to categorize the seismic hazard for a Tier 2 plant.

Implementation of the Recommended Process EPRI 3002017583 recommends a risk-informed graded approach for addressing the seismic hazard in the 1 O CFR 50.69 categorization process. There are a number of seismic fragility fundamental concepts that support a graded approach and there are important characteristics about the comparison of the seismic design basis (represented by the SSE) to the site-specific seismic hazard (represented by the GMRS) that support the selected thresholds between the three evaluation Tiers in the report. The coupling of these concepts with the categorization process in NEI 00-04 are the key elements of the approach defined in EPRI 3002017583 for identifying unique seismic insights.

The seismic fragility of an SSC is a function of the margin between an SSC's seismic capacity and the site-specific seismic demand. References such as EPRI NP-6041 (Reference 28) provide inherent seismic capacities for most SSCs that are not directly related to the site-specific seismic demand. This inherent seismic capacity is based on the non-seismic design loads (pressure, thermal, dead weight, etc.) and the required functions for the SSC. For example, a pump has a relatively high inherent seismic capacity based on its design and that same seismic capacity applies at a site with a very low demand and at a site with a very high demand.

There are some plant features such as equipment anchorage that have seismic capacities more closely associated with the site-specific seismic demand since those specific features are specifically designed to meet that demand. However, even for these features, the design basis criteria have intended conservatisms that result in significant seismic margins within SSCs. These conservatisms are reflected in key aspects of the seismic design process. The SSCs used in nuclear power plants are intentionally designed using conservative methods and criteria to ensure that they have margins well above the required design bases.

Experience has shown that design practices result in margins to realistic seismic capacities of 1.5 or more.

In applying the EPRI 3002017583 process for Tier 2 sites to the PBNP 10 CFR 50.69 categorization process, the IDP will be provided with the rationale for applying the guidance and informed of plant SSC-specific seismic insights that the IDP may choose to consider in their HSS/LSS deliberations. As part of the categorization team's preparation of the System Categorization Document (SCD) that is presented to the IDP, a section will be included that provides identified plant seismic insights as well as the basis for applicability of the EPRI 3002017583 study and the bases for PBNP being a Tier 2 plant. The discussion of the Tier 2 bases will include such factors as:

The moderate seismic hazard for the plant, The definition of Tier 2 in the EPRI study, and The basis for concluding PBNP is a Tier 2 plant.

At several steps of the categorization process the categorization team will consider the available seismic insights relative to the system being categorized and document their conclusions in the SCD. Integrated importance measures over all the modeled hazards (i.e.,

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 13 of 29 internal events, including internal flooding, and internal fire for PBNP) are calculated per Section 5.6 of NEI 00-04, and components for which these measures exceed the specified criteria are preliminary HSS which cannot be changed to LSS. For HSS SSCs uniquely identified by the PBNP PRA models but having design-basis functions during seismic events or functions credited for mitigation and prevention of severe accidents caused by seismic events, these will be addressed using non-PRA based qualitative assessments in conjunction with any seismic insights provided by the PRA.

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

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

• Impact of relay chatter

• Implications related to potential seismic interactions such as with block walls

• Seismic failures of passive SSCs such as tanks and heat exchangers

• Any known structural or anchorage issues with a particular SSC

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

For each system categorized, the categorization team will evaluate correlated seismic failures and seismic interactions between SSCs. This process is detailed in Section 2.3.1 of EPRI 3002017583, including the markups provided in Attachment 2 of References 9 and 10, and as described in this request. The process is summarized in Figure 2.

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Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 15 of 29 Determination of seismic insights will make use of the FPIE PRA model supplemented by focused seismic walkdowns. An overview of the process to determine the importance of SSCs for mitigating seismic events follows and is utilized on a system basis:

• Gather the population of SSCs in the system being categorized and review existing seismic information (Step 1 of Figure 2). This step may use the results of the required Tier 1 assessment that is performed along with the Tier 2 assessment. As stated in EPRI 3002017583, the technical basis for the Tier 1 approach generally applies for Tier 2 plants in addition to the additional sensitivity and walkdowns described herein.

• Assign seismic based SSC equipment class and distributed system IDs, as used for SPRAs, for SSCs in the system being categorized (Step 2 of Figure 2).

• Perform a series of screenings to refine the list of SSCs subject to correlation sensitivity studies. Screens will identify (Steps 3a/3b/3c of Figure 2):

o Inherently rugged SSCs o SSCs not in Level 1 or Level 2 PRAs o Components already identified as HSS components from the internal events PRA or integrated assessment o The above screened SSCs will still be evaluated for seismic interactions (Step 1 to Step Sb in Figure 2).

• SSCs identified in the above screening can be screened from consideration as functional correlation surrogate events. They are removed from the remainder of the process ( can be considered LSS) unless they are subject to interaction source considerations (Step 4 of Figure 2).

• Perform Tier 2 walkdown(s) focusing on identifying seismic correlated or interaction SSC failures for SSCs that were not previously walked down (Steps Sa/Sb of Figure 2).

• Screen out from further seismic considerations SSCs that are determined through the walkdown to be of high seismic capacity and not included in seismically correlated groups or correlated interaction groups since their non-seismic failure modes are already addressed for 50.69 categorization in the FPIE PRA and fire PRA. Those remaining components proceed forward for inclusion of associated seismic surrogate events in the Tier 2 Adjusted PRA Model (Steps Sc/6 of Figure 2).

• Develop a Tier 2 Adjusted PRA Model and incorporate seismic surrogate events into the model to reflect the potential seismically correlated and interaction conditions identified in prior steps (Steps 6/7 of Figure 2). The seismic surrogate basic events shall be added to the PRA under the appropriate areas in the logic model (e.g., given that the Tier 2 Adjusted PRA Model uses only loss of offsite power (LOOP) and small loss of coolant accident (LOCA) sequences, the seismic surrogate events should be added to system and/or nodal fault tree structures that tie into these sequence types). The probability of each seismic surrogate basic event added to the model should be set to 1.0E-04 (based on guidance in EPRI 3002017583).

• Quantify only the LOOP and small LOCA initiated accident sequences of the Tier 2 Adjusted PRA Model (Step 8 of Figure 2). The event frequency of the LOOP initiator shall be set to a value of 1.0 and the event frequency for the small LOCA initiator shall be set to a value of 1.0E-02. Remove credits for restoration of offsite power and other

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 16 of 29 functional recoveries (e.g., Emergency Diesel Generator (EOG) and DC power recovery).

• Utilize the importance measures from the quantification of the Tier 2 Adjusted PRA Model to identify appropriate SSCs (in the system being categorized) that should be HSS due to correlation or seismic interactions (Step 9 of Figure 2).

• SSCs screened out in Steps 5c, 6, or 9 in Figure 2 can be considered LSS (Step 10 of Figure 2).

• Prepare documentation of the Tier 2 analysis results, including identification of seismic unique HSS SSCs, for presentation to the IDP (Step 11 of Figure 2).

Seismic impacts would be compiled on an SSC basis. As each system is categorized, the system-specific seismic insights will be documented in the categorization report and provided to the IDP for consideration as part of the IDP review process. The IDP cannot challenge any candidate HSS recommendation for any SSC from a seismic perspective if they believe there is a basis, except for certain conditions identified in Step 10 of Section 2.3.1 of EPRI 3002017583 (including markups in References 9 and 10). Any decision by the IDP to downgrade preliminary HSS components to LSS will consider the applicable seismic insights in that decision. SSCs identified from the fire PRA as candidate HSS, which are not HSS from the internal events PRA or integrated importance measure assessment, will be reviewed for their design basis function during seismic events or functions credited for mitigation and prevention of severe accidents caused by seismic events. These insights will provide the IDP with a means to consider potential impacts of seismic events in the categorization process.

If the PBNP 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 1 O CFR 50.90, will be requested. Upon receipt of NRC approval for such a change, NextEra will follow its categorization review and adjustment process to review the changes to the plant and update, as appropriate, the SSC categorization in accordance with 10 CFR 50.69(e) and the EPRI 3002017583 SSC categorization criteria for the updated Tier. This includes use of the NextEra corrective action process.

If the seismic hazard is reduced such that it meets the criteria for Tier 1 in EPRI 3002017583, NextEra will implement the following process:

a) For previously completed system categorizations, NextEra may review the categorization results to determine if use of the criteria in EPRI 3002017583 Section 2.2, "Tier 1 - Low Seismic Hazard/ High Seismic Margin Sites," would lead to categorization changes. If changes are warranted, they will be implemented through the NextEra corrective action program and NEI 00-04, Section 12.

b) Seismic considerations for subsequent system categorization activities will be performed in accordance with the guidance in 3002017583 Section 2.2, "Tier 1 - Low Seismic Hazard/ High Seismic Margin Sites."

If the seismic hazard increases to the degree that an SPRA becomes necessary to demonstrate adequate seismic safety, NextEra will implement the following process following completion of the SPRA, including adequate closure of Peer Review Findings and Observations:

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 17 of 29 a) For previously completed system categorizations, NextEra will review the categorization results using the SPRA insights as prescribed in NEI 00-04 Section 5.3, "Seismic Assessment" and Section 5.6, "Integral Assessment." If categorization changes are warranted, they will be implemented through the NextEra corrective action program and NEI 00-04 Section 12.

b) Seismic considerations for subsequent system categorization activities will follow the guidance in NEI 00-04, as recommended in EPRI 3002017583 Section 2.4, "Tier 3 -

High Seismic Hazard/ Low Seismic Margin Sites."

Historical Seismic References for PBNP The PBNP GMRS and SSE curves from the seismic hazard and screening response are shown in Figure 1, as replicated from the seismic hazard and screening report (Reference 33). The NRC staff assessment of the PBNP seismic hazard and screening response is documented in Reference 34. In Section 3.3.3 of Reference 34, the NRC concluded that the methodology used by NextEra adequately characterizes the seismic hazard for the PBNP site.

Section 1.1.3 of EPRI 3002017583 cites various post-Fukushima seismic reviews performed for the U.S. fleet of nuclear power plants. For PBNP, the specific seismic reviews prepared by NextEra and the NRC's staff assessments of those reviews are provided in the following licensing documents.

1. NTTF Recommendation 2.1 Seismic Hazard Screening (References 33 and 34).

2. NTTF Recommendation 2.1 Spent Fuel Pool assessment (References 35 and 36).

3. NTTF Recommendation 2.3 Seismic Walkdowns (References 37, 38, 39, and 40).

4. NTTF Recommendation 4.2 Seismic Mitigation Strategy Assessment (S-MSA)

(References 41 and 42).

The following additional post-Fukushima seismic reviews were performed for PBNP:

5. NTTF Recommendation 2.1 Expedited Seismic Evaluation Process (ESEP)

(References 43 and 44).

6. NTTF Recommendation 2.1 Seismic High Frequency Evaluation (References 45 and 46).

Technical Information Precedent By letter dated January 31, 2020, Exelon Generation Company, LLC (EGC) submitted a license amendment request (Reference 7) to allow for the implementation of the provisions of 10 CFR 50.69 for LaSalle County Station (LaSalle), Units 1 and 2. Following the criteria in EPRI report 3002012988 (Reference 15), based on the GMRS-to-SSE comparison, the LaSalle site is considered a Tier 2 site, similar to the PBNP site. For the LaSalle seismic hazard assessment, EGC also proposed the use of a risk-informed graded approach that meets the requirements of 1 O CFR 50.69(b )(2) as an alternative to those listed in NEI 00-04.

EGC provided responses to NRC RAls pertaining to the alternative seismic approach in letters

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 18 of 29 dated October 1, 2020, October 16, 2020, and January 22, 2021 (References 8, 9, and 10, respectively). Based on information provided in Reference 7, as modified by the EGC RAI responses.in References 8, 9, and 10, the NRC issued the license amendments, approving the EGC request, on May 27, 2021 (Reference 11 ).

NextEra will follow the same alternative seismic approach in the 10 CFR 50.69 categorization process for PBNP as that which was approved by the NRC staff for LaSalle (Reference 11 ),

except for the site-specific LaSalle information (e.g., seismic capacity discussions, etc.).

PBNP site-specific seismic capacity information is described above herein. References 8, 9, and 10 are incorporated by reference into this amendment request as they provide additional supporting bases for Tier 2 plants, such as PBNP, to adopt the alternative seismic methodology for use in the 10 CFR 50.69 categorization process. Note that Reference 8 included a response to RAI 'APLC 50.69-RAI No. 12 (a)', which is not relevant to this submittal because it addresses external hazards that are outside the scope of this request.

In addition, References 16, 47, 48, and 49 are incorporated by reference into this amendment request as they provide additional supporting bases for Tier 1 plants that are also used for Tier 2 plants.

Summary PBNP is a Tier 2 plant for which there may be a limited number of unique seismic insights, most likely attributed to the possibility of seismically correlated failures, appropriate for consideration in determining HSS SSCs.

The special sensitivity study recommended using common cause failures, similar to the approach taken in a FPIE PRA, can identify the appropriate seismic insights to be considered with the other categorization insights by the IDP for the final HSS determinations. Use of the approach outlined in EPRI 3002017583 to assess seismic hazard risk for 1 O CFR 50.69 with the additional reviews discussed above will provide a process for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs that satisfies the requirements of 10 CFR 50.69(c).

3.2 TECHNICAL ADEQUACY EVALUATION (10 CFR 50.69(b)(2)(ii))

Section 3.2 of the NextEra LAR to adopt 10 CFR 50.69 (Reference 4) addressed the quality and level of detail of the systematic processes used at PBNP to evaluate internal and external events during normal operation, low power, and shutdown. These processes were reviewed and confirmed to be acceptable for performing SSC categorization, as documented in the License Amendment and associated Safety Evaluation approving use of 10 CFR 50.69 at PBNP (Reference 1 ). Those evaluations are not impacted by the changes proposed in this request.

3.3 PRA REVIEW PROCESS RESULTS (10 CFR 50.69(b)(2)(iii))

Following submittal of the NextEra LAR (Reference 4) to allow PBNP to implement the risk-informed categorization and treatment provisions of *10 CFR 50.69, periodic model

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 19 of 29 maintenance has been performed on the PBNP PRA to confirm that the models continue to represent the as-built, as-operated plant.

Since approval to adopt 10 CFR 50.69 at PBNP, NextEra submitted a license amendment request to adopt the RICT program (Reference 12). In Enclosure 2 of that request, information was provided supporting PRA consistency with RG 1.200, Revision 2, which included the peer review history of the PBNP internal events, including internal flooding, and fire PRA models (Reference 50). The NRC staff found that the PBNP internal events (including internal flooding) PRA and fire PRA were acceptable commensurate with the RICT application because NextEra's use of the PRA models in the integrated decision-making process was consistent with RG 1.17 4, Revision 3 (Reference 13).

In summary, the PRA models utilized at PBNP to implement the RICT program and 10 CFR 50.69 have been peer reviewed and remain adequate to support the SSC categorization process. Previous conclusions regarding the technical adequacy of the PBNP PRA models are unaffected by this request.

3.4 RISK EVALUATIONS (10 CFR 50.69(b)(2)(iv))

As described in Section 3.1 of this submittal, PBNP may implement the alternative seismic categorization process. The processes identified in the current license condition may continue to be used.

The overall risk evaluation process described in NEI 00-04 (Reference 2) addresses both known degradation mechanisms and common cause interactions and meets the requirements of§ 50.69(b)(2)(iv). The sensitivity studies discussed in Section 8 of NEI 00-04, will be used to confirm that the categorization process results in acceptably small increases to core damage frequency and large early release frequency.

The SSC failure rates and initiating event frequencies used in the PBNP PRA include the quantifiable impacts from known degradation mechanisms, as well as other mechanisms (e.g., design errors, manufacturing deficiencies, human errors, etc.).

Subsequent performance monitoring and PRA updates as required by 10 CFR 50.69 will continue to include this data and provide timely insights into the need to account for any important new degradation mechanisms.

3.5 FEEDBACK AND ADJUSTMENT PROCESS If significant changes to the plant risk profile are identified, or if it is identified that a RISC-3 or RISC-4 SSC can (or actually did) prevent a safety significant function from being satisfied, a prompt evaluation and review will be performed prior to the normally scheduled periodic review. Otherwise, the assessment of potential equipment performance changes and new technical information will be performed during the normally scheduled periodic review cycle.

To more specifically address the feedback and adjustment (i.e., performance monitoring) process as it pertains to the proposed alternative seismic method for Tier 2 sites discussed in

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 20 of 29 Section 3.1 of this submittal, implementation of the NextEra design control and corrective action programs provide assurance that the inputs for the qualitative determinations for seismic continue to remain valid to maintain compliance with the requirements of 10 CFR 50.69(e).

The performance monitoring process will be described in NextEra's 10 CFR 50.69 program documents. The program requires that the periodic review assess changes that could impact the categorization results and provides the IDP with an opportunity to recommend categorization and treatment adjustments. Personnel from engineering, operations, risk management, regulatory affairs, and others have responsibilities for preparing and conducting various performance monitoring tasks that feed into this process. The intent of the performance monitoring reviews is to discover trends in component reliability; to help catch and reverse negative performance trends and take corrective action if necessary.

The NextEra configuration control process ensures that changes to the plant, including a physical change to the plant and changes to documents, are evaluated to determine the impact to drawings, design bases, licensing documents, programs, procedures, and training.

The configuration control program will include a checklist of configuration activities to recognize those systems that have been categorized in accordance with 10 CFR 50.69 to ensure that any physical change to the plant or change to plant documents is evaluated prior to implementing those changes.

The checklist will include:

• A review of the impact on the SCD for configuration changes that may impact a categorized system under 10 CFR 50.69.

• Steps to be performed if redundancy, diversity, or separation requirements are identified or affected. These steps include identifying any potential seismic interaction between added or modified components and new or existing safety related or safe shutdown components or structures.

• Review of impact to seismic loading and SSE seismic requirements, as well as the method of combining seismic components.

• Review of seismic dynamic qualification of components if the configuration change adds, relocates, or alters Seismic Category I mechanical or electrical components.

NextEra has a comprehensive problem identification and corrective action program that requires the identification and resolution of issues. Issues that may impact the 1 O CFR 50.69 categorization process will be identified and addressed through the problem identification and corrective action program, including seismic-related issues.

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

Scheduled periodic reviews no more frequent than once every two refueling outages will evaluate new insights resulting from available risk information (i.e., PRA model or other analysis used in the categorization) changes, design changes, operational changes, and SSC performance. If it is determined that these changes have affected the risk information or other

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 21 of 29 elements of the categorization process such that the categorization results are more than minimally affected, then the risk information and the categorization process will be updated.

This scheduled review will include:

A review of plant modifications since the last review that could impact the SSC categorization, A review of plant specific operating experience that could impact the SSC categorization, A review of the impact of the updated risk information on the categorization process

results, A review of the importance measures used for screening in the categorization process, and An update of the risk sensitivity study performed for the categorization.

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

The periodic monitoring requirements of the 10 CFR 50.69 process will capture these issues and address them at a frequency commensurate with the issue severity. The 10 CFR 50.69 periodic monitoring program will include immediate and periodic reviews, that include the requirements of the regulation, to provide assurance that issues that could affect 10 CFR 50.69 categorization are addressed. The periodic monitoring process will also monitor the performance and condition of categorized SSCs such that the assumptions for reliability in the categorization process are maintained.

4 REGULATORY EVALUATION 4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA The following NRC requirements and guidance documents are applicable to the proposed change.

The regulations at 10 CFR 50.69, "Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors."

NRC Regulatory Guide 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," Revision 1, May 2006.

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

Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, March 2009.

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 22 of 29 The alternate Seismic Tier 2 categorization process is consistent with the applicable regulations in 10 CFR 50.69. The proposed change represents a deviation from the

. NEI 00-04 guidance endorsed in RG 1.201, Revision 1. However, the NRC staff specifies that licensees may propose alternative approaches for the 10 CFR 50.69 categorization process and should provide a basis in the submittal explaining why the approach and the accompanying method employed to assign safety significance to SSCs is technically acceptable. The intent of the technical evaluation provided in Section 3 of this submittal is to satisfy the NRC expectation cited in RG 1.201, Revision 1 regarding the proposed alternate Seismic Tier 2 approach to be used in the PBNP 10 CFR 50.69 categorization process. The proposed change is consistent with the applicable regulations and regulatory guidance.

4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS NextEra Energy Point Beach, LLC (NextEra) requests an amendment to the Point Beach Nuclear Plant (PBNP), Unit Nos. 1 and 2 Renewed Facility Operating Licenses (RFOL). The proposed change would revise the license condition associated with the adoption 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," that was added to the PBNP RFOL upon the issuance of Amendments 262 and 265. Specifically, the proposed change would revise the license condition for each unit to reflect an alternative to the approaches provided in Nuclear Energy Institute (NEI) 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision 0, to allow the use of an alternate Seismic Tier 2 approach in the 1 O CFR 50.69 categorization process.

NextEra has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes to Unit 1 License Condition M and Unit 2 License Condition L, "Additional Conditions," Functions 5 and 6 on Technical Specification (TS) Table 3.3.1-1, and TS Section 5.5.18.h are editorial in nature and have no effect on accident scenarios previously evaluated. The proposed change would also revise the PBNP license condition for each unit that was added with the issuance of PBNP license Amendments 262 and 265 to reflect use of an alternate Seismic Tier 2 approach in the 1 O CFR 50.69 categorization process. The approach is described in EPRI 3002017583, "Alternative Approaches for Addressing Seismic Risk in 1 O CFR 50.69 Risk-Informed Categorization."

With the proposed change, PBNP will continue to be permitted to use a risk-informed categorization process to modify the scope of structures, systems and components (SSCs) subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The process used to evaluate SSCs for changes to NRC special treatment requirements and the use of alternative requirements will continue to ensure the ability of the SSCs to perform their design function. The potential change to special treatment requirements using the alternate Seismic Tier 2 categorization process does not change the design and operation of the SSCs. As a result, the proposed change to revise the 10 CFR 50.69 categorization process to reflect an alternate Seismic Tier 2 methodology does not significantly affect any initiators to accidents previously evaluated

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 23 of 29 or the ability to mitigate any accidents previously evaluated. The consequences of the accidents previously evaluated are not affected because the mitigation functions performed by the SSCs assumed in the safety analysis are not being modified. The SSCs required to safely shut down the reactor and maintain it in a safe shutdown condition following an accident will continue to perform their design functions.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes to Unit 1 License Condition M and Unit 2 License Condition L, "Additional Conditions," Functions 5 and 6 on TS Table 3.3.1-1, and TS Section 5.5.18.h are editorial in nature and do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the method governing normal plant operation. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of this administrative change. The proposed change would also revise the PBNP license condition for each unit that was added with the issuance of license Amendments 262 and 265 to reflect use of an alternate Seismic Tier 2 approach in the 10 CFR 50.69 categorization process. The approach is described in EPRI 3002017583. With the proposed change, PBNP will continue to be permitted to use a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not change the functional requirements, configuration, or method of operation of any SSC. Under the proposed change, no additional plant equipment will be installed.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed changes to Unit 1 License Condition M and Unit 2 License Condition L, "Additional Conditions," Functions 5 and 6 on TS Table 3.3.1-1, and TS Section 5.5.18.h are editorial in nature and do not alter setpoints at which protective actions are initiated or the manner in which safety limits are determined. The proposed change would also revise the PBNP license condition for each unit that was added with the issuance of license Amendments 262 and 265 to reflect use of an alternate Seismic Tier 2 approach in the 10 CFR 50.69 categorization process. The approach is described in EPRI 3002017583. With the proposed change, PBNP will continue to be permitted to use a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not affect any Safety Limits or operating parameters used to establish the safety margin. The safety margins included in analyses of accidents are not affected by the proposed change. The regulation requires that there be no significant effect on plant risk due to any change to the special treatment requirements for SSCs and that the SSCs continue to be capable of performing their design basis functions, as well as to perform any beyond design basis functions consistent with the categorization process and results.

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 24 of 29 Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, NextEra concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92( c), and accordingly, a finding of "no significant hazards consideration" is justified.

4.3 CONCLUSION

S In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5 ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 6

REFERENCES L-2024-120 Enclosure Page 25 of 29

1.

NRC Letter to NextEra, "Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Amendments to Adopt 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'," November 26, 2018 (ML18289A378).

2.

Nuclear Energy Institute (NEI) 00-04, "10 CFR 50.69 SSC Categorization Guideline,"

Revision 0, July 2005 (ML052910035).

3.

Electric Power Research Institute (EPRI) 3002017583, "Alternative Approaches for Addressing Seismic Risk in 1 O CFR 50.69 Risk-Informed Categorization," Technical Update, February 2020 (ML21082A170).

4.

NextEra (Point Beach) Letter 2017-0043 to NRC, "License Amendment Request 287, Application to adopt 10 CFR 50.69, 'Risk-informed Categorization and Treatment of Structures, System, and Components (SSCs) for Nuclear Power Plants'," August 31, 2017 (ML17243A201).

5.

NextEra Letter (Point Beach) 2018-0044 to NRC, "Supplement to Response to Request for Additional Information Regarding License Amendment Request 287, Application to Adopt 1 O CFR 50.69, 'Risk informed Categorization and Treatment of Structures, System, and Components (SSCs) for Nuclear Power Plants'," September 28, 2018 (ML18271A114).

6.

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

7.

Exelon Generation Company (LaSalle) Letter to NRC, "Application to Adopt 10 CFR 50.69, 'Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors'," January 31, 2020 (ML20031E699).

8.

Exelon Generation Company (LaSalle) Letter to NRC, "Response to Request for Additional Information regarding LaSalle License Amendment Request to Renewed Facility Operating Licenses to Adopt 1 O CFR 50.69, 'Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors'," October 1, 2020 (ML20275A292).

9.

Exelon Generation Company (LaSalle) Letter to NRC, "Response to Request for Additional Information regarding LaSalle License Amendment Request to Renewed Facility Operating Licenses to Adopt 1 O CFR 50.69, 'Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors',"

October 16, 2020 (ML20290A791).

10. Exelon Generation Company (LaSalle) Letter to NRC, "Response to Request for Additional Information Regarding the License Amendment Request to Adopt 1 O CFR 50.69," January 22, 2021 (ML21022A130).

11. NRC Letter to Exelon Generation Company, "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'," May 27, 2021 (ML21082A422).

12. NextEra (Point Beach) Letter 2022-0007 to NRC, "License Amendment Request 297, Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505,

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 26 of 29 Revision 2, 'Provide Risk Informed Extended Completion Times - RITSTF Initiative 4b',"

May 20, 2022 (ML22140A132).

13. NRC Letter to NextEra, "Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Amendment Nos. 271 and 273 Regarding Technical Specifications to Adopt TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b',"

June 1, 2023 (ML23103A133).

14. NextEra (Point Beach) Letter 2018-0038 to NRC, "Response to Request for Additional Information Regarding License Amendment Request 287, Application to Adopt 10 CFR 50.69, 'Risk informed Categorization and Treatment of Structures, System, and Components (SSCs) for Nuclear Power Plants'," August 10, 2018 (ML18222A539).

15. Electric Power Research Institute (EPRI) 3002012988, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," July 2018.

16. NRC letter to Exelon Generation Company, "Calvert Cliffs Nuclear Power Plant, Units 1 and 2 - Issuance of Amendment Nos. 332 and 310 Re: Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, February 28, 2020 (ML193300909).

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

18. NRC 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).

19. Exelon Generation Company (Peach Bottom) Letter to NRC, "Seismic Probabilistic Risk Assessment Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," August 28, 2018 (ML18240A065).

20. NRC Letter to Exelon Generation Company, "Peach Bottom Atomic Power Station, Units 2 and 3 - Staff Review of Seismic Probabilistic Risk Assessment Associated With Reevaluated Seismic Hazard Implementation of the Near-Term Task Force Recommendation 2.1: Seismic," June 10, 2019 (ML19053A469).

21. NRC Letter to Exelon Generation Company, "Peach Bottom Atomic Power Station, Units 2 and 3 - Correction Regarding Staff Review of Seismic Probabilistic Risk Assessment Associated With Reevaluated Seismic Hazard Implementation of the Near-Term Task Force Recommendation 2.1: Seismic," October 8, 2019 (ML19248C756).

22. Southern Nuclear Operating Company Letter to NRC, "Vogtle Electric Generating Plant -

Units 1 and 2 License Amendment Request to Modify Approved 1 O CFR 50.69 Categorization Process," June 22, 2017 (ML17173A875).

23. NRC Letter to Southern Nuclear Operating Company, "Vogtle Electric Generating Plant, Units 1 and 2 - Issuance of Amendments Regarding Application of Seismic Probabilistic Risk Assessment into the Previously Approved 10 CFR 50.69 Categorization Process,"

August 10, 2018 (ML18180A062).

24. Tennessee Valley Authority Letter to NRC, "Seismic Probabilistic Risk Assessment for Watts Bar Nuclear Plant, Units 1 and 2 - Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 27 of 29 Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," June 30, 2017 (ML17181A485).

25. Tennessee Valley Authority Letter to NRC, "Tennessee Valley Authority (TVA) - Watts Bar Nuclear Plant Seismic Probabilistic Risk Assessment Supplemental Information,"

April 10, 2018 (ML18100A966).

26. NRC Letter to Tennessee Valley Authority, "Watts Bar Nuclear Plant, Units 1 and 2 -

Staff Review of Seismic Probabilistic Risk Assessment Associated with Reevaluated Seismic Hazard Implementation of the Near-Term Task Force Recommendation 2.1:

Seismic," July 10, 2018 (ML18115A138).

27. NRC letter to Tennessee Valley Authority, "Watts Bar Nuclear Plant, Units 1 and 2 -

Issuance of Amendment Nos. 134 and 38 Regarding Adoption of Title 10 of the Code of Federal Regulations Section 50.69, 'Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Plants'," April 30, 2020 (ML20076A194).

28. Electric Power Research Institute (EPRI) NP-6041-SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Revision 1," August 1991.

29. Exelon Generation Company (Peach Bottom) Letter to NRC, "Supplemental Information to Support Application to Adopt 10 CFR 50.69, 'Risk-informed categorization and treatment of structures, systems, and components for nuclear power plants'," June 6, 2018 (ML18157A260).

30. Southern Nuclear Operating Company Letter to NRC, "Vogtle Electric Generating Plant -

Units 1 & 2 License Amendment Request to Incorporate Seismic Probabilistic Risk Assessment into the 1 O CFR 50.69 Categorization Process Response to Request for Additional Information (RAls 4-11 )," February 21, 2018 (ML18052B342).

31. Tennessee Valley Authority Letter to NRC, "Watts Bar Nuclear Plant, Units 1 and 2, Application to Adopt 1 O CFR 50.69, 'Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors'," November 29, 2018 (ML18334A363).

32. NRC Letter to Power Reactor Licensees, "Final Determination of Licensee Seismic Probabilistic Risk Assessments Under the Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendation 2.1 'Seismic' of the Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident,"

October 27, 2015 (ML15194A015).

33. NextEra Letter 2014-0024 to NRC, "NextEra Energy Point Beach, LLC Seismic Hazard and Screening Report (CEUS Sites), Response NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," March 31, 2014 (ML14090A275).

34. NRC Letter to NextEra, "Point Beach Nuclear Plant, Units 1 and 2 - Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," August 3, 2015 (ML15211A593).

35. NextEra (Point Beach) Letter to NRC, "Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 1 O CFR 50.54(f) Regarding

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 28 of 29 Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, November 30, 2016 (ML16335A143).

36. NRC Letter to NextEra, "Point Beach Nuclear Plant, Units 1 and 2 - Staff Review of Spent Fuel Pool Evaluation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1," January 6, 2017 (ML16349A572).

37. NextEra Letter 2012-0101 to NRC, "NextEra Energy Point Beach, LLC Response to 10 CFR 50.54(f) Request for Information Regarding Near-Term Task Force Recommendation 2.3, Seismic," November 26, 2012 (ML12332A070).

38. NextEra Letter 2013-0025 to NRC, "Update to NextEra Energy Point Beach, LLC Response to 10 CFR 50.54(f) Request for Information Regarding Near-Term Task Force Recommendation 2.3, Seismic," October 3, 2013 (ML13277A109).

39. NextEra Letter 2013-0108 to NRC, "NextEra Energy Point Beach, LLC Response to 10 CFR 50.54(f) Request for Additional Information Associated with Near-Term Task Force Recommendation 2.3, Seismic Walkdowns," November 27, 2013 (ML13331A912).

40. NRC Letter to NextEra, "Point Beach Nuclear Plant, Units 1 and 2 - Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-ichi Nuclear Power Plant Accident,"

April 11, 2014 (ML14072A253).

41. NextEra Letter 2017-0040 to NRC, "NextEra Energy Point Beach, LLC, Seismic Mitigating Strategies Assessment (MSA) Report for the Reevaluated Seismic Hazard Information - NEI 12-06, Appendix H, Section H.4.4, Path 4: GMRS ~ 2X SSE," August 17, 2017 (ML172298210).

42. NRC Letter to NextEra, "Point Beach Nuclear Plant, Units 1 and 2 - Staff Review of Mitigating Strategies Assessment Report of the Impact of the Reevaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(f) Letter," November 16, 2017 (ML173108531).

43. NextEra Letter 2014-0088 to NRC, "NextEra Energy Point Beach, LLC's Expedited Seismic Evaluation Process Report (CEUS Sites), Response NRC Request for Information Pursuant to 1 O CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," December 22, 2014 (ML14356A426).

44. NRC Letter to NextEra, "Point Beach Nuclear Plant, Units 1 and 2 - Staff Review of Interim Evaluation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1," July 29, 2015 (ML15209A657).

45. NextEra Letter 2017-0037 to NRC, "NextEra Energy Point Beach, LLC, High Frequency Seismic Evaluation Confirmation Report, August 2, 2017 (ML17214A268).

46. NRC Letter to NextEra, "Point Beach Nuclear Plant, Units 1 and 2 - Staff Review of High Frequency Confirmation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1," August 21, 2017 (ML172298187).

47. Exelon Generation Company (Calvert Cliffs) Letter to NRC, "Response to Request for Additional Information Regarding the Application to Adopt 10 CFR 50.69, 'Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors'," July 1, 2019 (ML19183A012).

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Enclosure Page 29 of 29

48. Exelon Generation Company (Calvert Cliffs) Letter to NRC, "Response to Request for Additional Information Regarding the Application to Adopt 10 CFR 50.69, 'Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors'," July 19, 2019 (ML19200A216).

49. Exelon Generation Company (Calvert Cliffs) Letter to NRC, "Revised Response to Request for Additional Information Regarding the Application to Adopt 10 CFR 50.69,

'Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors'," August 5, 2019 (ML19217A143).

50. Enclosures 1 and 2 of NextEra's License Amendment Request to Adopt the RICT Program, May 20, 2022 (ML22140A142).

51. Point Beach License Amendment Request 280, Removal of Completed License Conditions and Changes to the Ventilation Filter Testing Program (ML16043A217).

52. Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Amendments - Removal of Completed License Conditions and Changes to the Ventilation Filter Testing Program (ML17039A300).

53. Point Beach License Amendment Request 297, Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk Informed Extended Completion Times - RITSTF Initiative 4b" (ML22140A131).

54. Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Amendment Nos. 271 and 273 Regarding Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk Informed Extended Completion Times - RITSTF Initiative 4b" (ML23103A133).

55. Point Beach License Amendment Request 271, Transition to 10 CFR 50.48(c) - NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition (ML13182A353 and ML13182A350).

56. Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Amendments Regarding Transition to a Risk-Informed, Performance-Based Fire Protection Program in Accordance with 10 CFR 50.48(c) (ML16196A093).

ATTACHMENT 1 Point Beach Nuclear Plant - Units 1 and 2 LAR 301 - Revise Renewed Facility Operating Licenses to Adopt an Alternative Approach for Addressing Seismic Risk in 1 O CFR 50.69 Risk-Informed Categorization Mark-ups of Point Beach Units 1 and 2 Renewed Facility Operating Licenses (4 pages follow)

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301

~.. '..

L-2024-120 Point Beach Nuclear Plant Unit 1, Renewed Facility Operating License No. DPR-24 Markup of 1 O CFR 50.69 License Condition E. Adoption of 10 CFR 50.69, "Risk-Informed categorization and treatment of structures, systems, and components for nuclear power plants" 4--:-

NextEra Energy Point Beach is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using:

Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (AN0-2) 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 License Amendment No. 262 dated November 26, 2018.

In addition, NextEra Energy Point Beach is approved to implement 1 O CFR 50.69 using the alternative seismic approach described in NextEra Energy Point Beach letter L-2024-120, dated September X, 2024, for categorization of RISC-1, RISC-2, RISC 3, and RISC-4 SSCs, as specified in License Amendment No. (XXX] dated (DATE].

2.

Prior to implementation of the provisions of 10 CFR 50.69, NextEra Energy Point Beach shall complete the items belmv:

a.

Item A in,1\\ttachment 1, List of Categorization Prerequisites, to NextEra Energy Point Beach letter NRG 2017 0043, "License Amendment Request 287, Application to Adopt 10 CFR 50.69, 'Risk Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear PmNer Plants,'

11 dated August 31, 2017; and

b. Attachment 1, Point Beach 10 CFR 50.69 PRA Implementation Items, in NextEra Energy Point Beach letter NRG 2018 0044, "Supplement to Response to Request for Additional Information Regarding License Amendment Request 287, Application to Adopt 10 CFR 50.69, 'Risk informed Categorization and Treatment of Structures, System, and Components (SSCs) for Nuclear Power Plants,'

11 dated September 28, 2018.

J,. 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).

A-1

1. Fire fighting response strategy with the following elements:

a.

Pre-defined coordinated fire response strategy and guidance

b.

Assessment of mutual aid fire fighting assets

c.

Designated staging areas for equipment and materials

d.

Command and control

e.

Training of response personnel

2. Operations to mitigate fuel damage considering the following:

a.

Protection and use of personnel assets

b.

Communications

c.

Minimizing fire spread

d.

Procedures for implementing integrated fire response strategy

e.

Identification of readily-available pre-staged equipment

f.

Training on integrated fire response strategy

g.

Spent fuel pool mitigation measures

3. Actions to minimize release to include consideration of:

a.

Water spray scrubbing

b.

Dose to onsite responders M. Additional Conditions The additional conditions contained in Appendix C, as revised through Amendment No. ~258, are hereby incorporated into this license. NextEra Energy Point Beach shall operate the facility in accordance with the additional conditions.

5.

The issuance of this renewed operating license is without prejudice to subsequent licensing action which may be taken by the Commission with regard to the ongoing rulemaking hearing on the Interim Acceptance Criteria for Emergency Core Cooling Systems (Docket No. RM 50-1).

6.

This renewed operating license is effective as of the date of issuance, and shall expire at midnight on October 5, 2030.

FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By R. W. Borchardt, Deputy Director Office of Nuclear Reactor Regulation Attachments:

1. Appendix A - Technical Specifications

2. Appendix B - Environmental Technical Specifications

3. Appendix C - Additional Conditions Date of Issuance: December 22, 2005 Renewed License No. DPR-24 Amendment No. 263

Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2024-120 Point Beach Nuclear Plant Unit 2, Renewed Facility Operating License No. DPR-27 Markup of 10 CFR 50.69 License Condition E. Adoption of 10 CFR 50.69, "Risk-Informed categorization and treatment of structures, systems, and components for nuclear power plants" 4-:-

NextEra Energy Point Beach is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using:

Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (AN0-2) 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 License Amendment No. 265 dated November 26, 2018.

In addition, NextEra Energy Point Beach is approved to implement 10 CFR 50.69 using the alternative seismic approach described in NextEra Energy Point Beach letter L-2024-120, dated September X, 2024, for categorization of RISC-1, RISC-2, RISC 3, and RISC-4 SSCs, as specified in License Amendment No. (XXX] dated (DATE].

2. Prior to implementation of the provisions of 10 CFR 50.69, NextEra Energy Point Beach shall complete the items below:

a. Item A in Attachment 1, List of Categorization Prerequisites, to NextEra Energy Point Beach letter NRG 2017 0043, "License Amendment Request 287, Application to Adopt 10 CFR 50.69, 'Risk Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Plants,'

11 dated August 31, 2017; and

b. Attachment 1, Point Beach 10 CFR 50.69 PRA Implementation Items, in NextEra Energy Point Beach letter NRG 2018 0044, "Supplement to Response to Request for Additional Information Regarding License Amendment Request 287, Application to Adopt 10 CFR 50.69, 'Risk informed Categorization and Treatment of Structures, System, and Components (SSCs) for Nuclear Power Plants,'

11 dated September 28, 2018.

a.., 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).

A-2 L. Additional Conditions The additional conditions contained in Appendix C, as revised through Amendment No. ~262, are hereby incorporated into this license. NextEra Energy Point Beach shall operate the facility in accordance with the additional conditions.

5.

The issuance of this renewed operating license is without prejudice to subsequent licensing action which may be taken by the Commission with regard to the ongoing rulemaking hearing on the Interim Acceptance Criteria for Emergency Core Cooling Systems (Docket No. RM 50-1).

6.

This renewed operating license is effective as of the date of issuance, and shall expire at midnight on March 8, 2033.

FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By R. W. Borchardt, Deputy Director Office of Nuclear Reactor Regulation Attachments:

1. Appendix A-Technical Specifications

2. Appendix B - Environmental Technical Specifications

3. Appendix C - Additional Conditions Date of Issuance: December 22, 2005 Renewed License No. DPR-27 Amendment No. 262

ATTACHMENT 2 Point Beach Nuclear Plant - Units 1 and 2 LAR 301 - Revise Renewed Facility Operating Licenses to Adopt an Alternative Approach for Addressing Seismic Risk in 1 O CFR 50.69 Risk-Informed Categorization Proposed Technical Specification Pages (Mark-Ups)

(2 pages follow)

~

RPS Instrumentation 3.3.1

I Table 3.3.1-1 (page 1 of 8)

Reacto r Protection System lnstru.mentation FUNCTION APPLICABLE REQUIRED

• . SURVEILLANCE ALLOWABLE NOMINAL TRIP MODES CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

1.

Manual 1,2 2

B SR 3.3.1.13 NA NA Reactor Trip 3(a), 4(a), 5 2

C SR 3.3. 1. 13 NA NA (a)

2.

Power Range Neutron Flux

a.

High 1,2 4

D SR 3.3.1.1

,; 108% RTP 106% RTP SR 3.3. 1.2 SR 3.3.1.?<m>

SR 3.3. 1. 11 <m>

b.

Low 1 (b),2 4

D SR3.3.1.1

,; 27% RTP 20% RTP SR 3.3. 1.a<m>

SR 3.3. 1. 11 <m>

3.

Intermediate 1 (b) ' 2(c) 2 F,G SR 3.3. 1. 1

,; 43% RTP 25% RTP Range SR 3.3. 1.a<m>

Neutron Flux SR 3.3. 1. 11 <m>

4.

Source Range 2(d) 2 H,I SR 3.3.1.1

~ 3.0 E5 cps 1.5 E5 cps Neutron Flux SR 3.3. 1.a<m>

SR 3.3.1.11<m>

3(a), 4(a), 5(a) 2 l,J SR 3.3.1.1

~ 3.0 E5 cps 1.5 E5 cps SR 3.3.1.?<m>

SR3.3.1.11<m>

5.

1,2 4

D SR 3.3.1.1 Refer to Note 1 Refer to Note 1 Overtemper SR 3.3.1.3 (Page (Page 3.3. 1-ature ~T SR 3.3.1.6 3.3. 1-48£ill 48£ill SR 3.3.1.?<m>

SR 3.3.1.1 1<m>

6.

Overpower ~ T 1,2 4

D SR 3.3.1.1 Refer to Note 2 Refer to Note 2 SR 3.3. 1.7<m>

(Page (Page SR 3.3. 1. 11<m>

3.3.1-49.f.!)

3.3.1-49.f.!)

(continued)

(a)

With Reactor Trip Breakers (RTBs) closed and Rod Control System capable of rod withdrawal.

(b)

Below the P-1 O (Power Range Neutron Flux) interlock.

(c)

Above the P-6 (Intermediate Range Neutron Flux) interlock.

(d)

Below the P-6 (Intermediate Range Neutron Flux) interlock.

(m)

Table 3.3.1-1 Notes 3 and 4 are applicable Point Beach 3.3.1-15 Unit 1 - Amendment No. 271 Unit 2 - Amendment No. 273 4 -

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.18 5.5.19 Point Beach Control Room Envelope Habitability Program (continued)

g.

h.

An adequate supply of self contained breathing apparatus (SCBA) units in the CRE to protect CRE occupants from a hazardous chemical release.

Portable smoke ejection equipment per the Fire Protection Evaluation Report and Safe Shutdown Analysis ReportProgram Design Document to address a potential smoke challenge.

Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operations are met:

a. The Surveillance Frequency Control Program shall contain a list of frequencies of those Surveillance Requirements for which the frequency is controlled by the program.

b. Changes to the frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the frequencies established in the Surveillance Frequency Control Program.

5.5-20 Unit 1 - Amendment No. 271 Unit 2 - Amendment No. 273