L-2025-041, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

ML25080A172
Person / Time
Site:	Seabrook
Issue date:	03/21/2025
From:	Mack K NextEra Energy Seabrook
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
L-2025-041
Download: ML25080A172 (1)

Text

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

Seabrook Station Docket No. 50-443 Renewed Facility Operating License No. NPF-86 March 21, 2025 L-2025-041 10 CFR 50.90 10 CFR 50.69 Application to Adopt 10 CFR 50.69, "Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors" In accordance with the provisions of Sections 50.90 and 50.69 of Title 10 of the Code of Federal Regulations (10 CFR), NextEra Energy Seabrook, LLC (NextEra) is requesting an amendment to the Renewed Facility Operating License (RFOL) of Seabrook Station, Unit 1 (Seabrook).

The proposed amendment would modify the Seabrook licensing basis, by the addition of a license condition, to allow for the implementation of the provisions of 10 CFR 50.69, "Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors." The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

The enclosure to this letter provides the basis for the proposed change to the Seabrook RFOL.

The categorization process being implemented through this change is consistent with Nuclear Energy Institute (NEI) 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision 0, dated July 2005, which was endorsed by the Nuclear Regulatory Commission (NRC) in Regulatory Guide 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," Revision 1, dated May 2006. Attachment 1 of the enclosure provides a list of categorization prerequisites. Use of the categorization process on a plant system will only occur after these prerequisites are met.

The probabilistic risk assessment (PRA) models described within this license amendment request (LAR) are the same as those described within the NextEra submittal of the Seabrook LAR dated February 3, 2025, for the adoption of Technical Specifications Task Force (TSTF)

Traveler 505, Revision 2 (Letter No. L-2025-015). NextEra requests that the NRC conduct their review of the PRA technical adequacy details for this application in coordination with the review of the TSTF-505 application. This would reduce the NextEra and NRC resources necessary to complete the review of the applications. This request should not be considered a linked NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874

Seabrook Station Docket No. 50-443 L-2025-041 Page 2 of 2 requested licensing action (RLA), as the details of the PRA models in each LAR are complete.

NextEra requests approval of the proposed license amendment within 12 months following acceptance, with an implementation period of 60 days.

There are no regulatory commitments made in this submittal.

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

In accordance with 10 CFR 50.91(b)(1), a copy of this application is being forwarded to the designee for the State of New Hampshire.

Should you have any questions regarding this submittal, 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~ day of March 2025.

Sincerely, Kenneth A. Mack Director, Licensing and Regulatory Compliance

Enclosure:

Description and Assessment of the Proposed Change Attachments (6) cc:

USNRC Region I Administrator USNRC Project Manager USNRC Senior Resident Inspector Director Homeland Security and Emergency Management New Hampshire Department of Safety Division of Homeland Security and Emergency Management Bureau of Emergency Management 33 Hazen Drive Concord, NH 03305 Kimberly Castle, Technological Hazards Unit Supervisor The Commonwealth of Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399

Seabrook Station Docket No. 50-443 Enclosure Description and Assessment of the Proposed Change L-2025-041 Enclosure Page 1 of 34

Subject:

Application to Adopt 10 CFR 50.69, "Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors"

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION 2.1 Current Regulatory Requirements 2.2 Reason for the Proposed Change 2.3 Description of the Proposed Change

3. TECHNICAL EVALUATION 3.1 Categorization Process Description (1 O CFR 50.69(b)(2)(i))

3.1.1 Overall Categorization Process 3.1.2 Passive Categorization Process 3.2 Technical Adequacy Evaluation (1 O CFR 50.69(b)(2)(ii))

3.2.1 Internal Events and Internal Flooding 3.2.2 Fire Hazards 3.2.3 Seismic Hazards 3.2.4 Other External Hazards 3.2.5 Low Power and Shutdown 3.2.6 PRA Maintenance and Updates 3.2.7 PRA Uncertainty Evaluations 3.3 PRA Review Process Results (1 O CFR 50.69(b)(2)(iii))

3.4 Risk Evaluations (1 O CFR 50.69(b)(2)(iv))

3.5 Feedback and Adjustment Process

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 No Significant Hazards Consideration Determination Analysis 4.3 Conclusions

5. ENVIRONMENTAL CONSIDERATION

6. REFERENCES

Seabrook Station Docket No. 50-443 ATTACHMENTS:

1. List of Categorization Prerequisites

2. Description of PRA Models Used in Categorization L-2025-041 Enclosure Page 2 of 34

3. Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items

4. External Hazards Screening

5. Progressive Screening Approach for Addressing External Hazards

6. Disposition of Key Assumptions/Sources of Uncertainty

Seabrook Station Docket No. 50-443

1.

SUMMARY

DESCRIPTION L-2025-041 Enclosure Page 3 of 34 The proposed amendment modifies the licensing basis to allow for the implementation of the provisions of Title 1 O of the Code of Federal Regulations (1 O CFR), Section 50.69, "Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors."

The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance (LSS), alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance (HSS), requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

2.

DETAILED DESCRIPTION 2.1 Current Regulatory Requirements The Nuclear Regulatory Commission (NRC) has established a set of regulatory requirements for commercial nuclear reactors to ensure that a reactor facility does not impose an undue risk to the health and safety of the public, thereby providing reasonable assurance of adequate protection to public health and safety. The current body of NRC regulations and their implementation are largely based on a "deterministic" approach.

This deterministic approach establishes requirements for engineering margin and quality assurance in design, manufacture, and construction. In addition, it assumes that adverse conditions can exist (e.g., equipment failures and human errors) and establishes a specific set of design basis events (DBEs). The deterministic approach then requires that the facility include safety systems capable of preventing or mitigating the consequences of those DB Es to protect public health and safety. The structures, systems and components (SSCs) necessary to defend against the DBEs are defined as "safety-related," and these SSCs are the subject of many regulatory requirements, herein referred to as "special treatments," designed to ensure that they are of high quality and high reliability, and have the capability to perform during postulated design basis conditions. Treatment includes, but is not limited to, quality assurance, testing, inspection, condition monitoring, assessment, evaluation, and resolution of deviations. The distinction between "treatment" and "special treatment" is the degree of NRC specification as to what must be implemented for particular SSCs or conditions. Typically, the regulations establish the scope of SSCs that receive special treatment using one of three different terms: "safety-related," "important to safety," or "basic component." The terms "safety-related" and "basic component" are defined in the regulations, while "important to safety," used principally in the general design criteria (GDC) of Appendix A to 10 CFR Part 50, is not explicitly defined.

2.2 Reason for the Proposed Change A probabilistic approach to regulation enhances and extends the traditional deterministic approach by allowing consideration of a broader set of potential challenges to safety, providing a logical means for prioritizing these challenges based on safety significance, and allowing consideration of a broader set of resources to defend against these challenges. In contrast to the deterministic approach, probabilistic risk assessments (PRAs) address credible initiating events by assessing the event frequency. Mitigating system reliability is then assessed, including the

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 4 of 34 potential for common cause failures. The probabilistic approach to regulation is an extension and enhancement of traditional regulation by considering risk in a comprehensive manner.

To take advantage of the safety enhancements available using PRA, in 2004 the NRC published a new regulation, 10 CFR 50.69. The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of LSS, alternative treatment requirements can be implemented in accordance with the regulation. For equipment determined to be of HSS, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

The rule contains requirements on 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. The determination of safety significance is performed by an integrated decision-making process, as described by Nuclear Energy Institute (NEI) 00-04, "10 CFR 50.69 SSC Categorization Guideline" (Reference 1 ), which uses both risk insights and traditional engineering insights. The safety functions include the design basis functions, as well as functions credited for severe accidents (including external events). Special or alternative treatment for the SSCs is applied as necessary to assure functionality and reliability are maintained and is a function of the SSC categorization results and associated bases. Finally, periodic assessment activities are conducted to adjust the categorization and/or treatment processes as needed so that SSCs continue to meet all applicable requirements.

The rule 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, the rule 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 enhanced. Conversely, for SSCs that do not significantly contribute to plant safety on an individual basis, the rule allows an alternative risk-informed approach to treatment that provides reasonable confidence (which is a reduced level compared to the reasonable assurance criteria used for many special treatments) that these SSCs will satisfy functional requirements.

Implementation of 1 O CFR 50.69 will allow NextEra Energy Seabrook, LLC (NextEra) to improve focus on equipment that has safety significance resulting in improved plant safety.

2.3 Description of the Proposed Change NextEra proposes the addition of the following condition to the Renewed Facility Operating License (RFOL) of Seabrook Station, Unit 1 (Seabrook) to document the NRC's approval of the use of 1 O CFR 50.69.

NextEra Energy Seabrook, LLC, is approved to implement 1 O 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, 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, Class 3, and non-class SSCs and their associated supports; the results of non-PRA

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 5 of 34 evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach described in NextEra's submittal letter L-2025-041 dated March 21, 2025; as specified in License Amendment No. [XXX] dated [DATE].

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.

3.

TECHNICAL EVALUATION 10 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).

Each of these submittal requirements are addressed in the following sections.

The PRA models described within this license amendment request (LAR) are the same as those described within the NextEra submittal of the Seabrook LAR for the adoption of Technical Specifications Task Force (TSTF) Traveler 505, Revision 2. NextEra requests that the NRC conduct their review of the PRA technical adequacy details for this application in coordination with the review of the TSTF-505 application. This would reduce the NextEra and NRC resources necessary to complete the review of the applications. This request should not be considered a linked requested licensing action (RLA), as the details of the PRA models in each LAR are complete.

Seabrook Station Docket No. 50-443 3.1 Categorization Process Description (10 CFR 50.69(b)(2)(i))

3.1.1 Overall Categorization Process L-2025-041 Enclosure Page 6 of 34 NextEra will implement the risk categorization process in accordance with NEI 00-04, Revision 0, as endorsed by Regulatory Guide (RG) 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance" (Reference 2).

NEI 00-04 Section 1.5 states "Due to the varying levels of uncertainty and degrees of conservatism in the spectrum of risk contributors, the risk significance of SSCs is assessed separately from each of five risk perspectives and used to identify SSCs that are potentially safety-significant." A separate evaluation is appropriate to avoid reliance on a combined result that may mask the results of individual risk contributors.

The process to categorize each system will be consistent with the guidance in NEI 00-04, as endorsed by RG 1.201, with the exception of the seismic hazard impact evaluation, which will use the Electric Power Research Institute (EPRI) 3002017583 (Reference 3) approach for seismic Tier 2 sites. The inclusion of additional process steps to address seismic considerations will ensure that the reasonable confidence required by 10 CFR 50.69(c)(1)(iv) is achieved. RG 1.201 states that "the implementation of all processes described in NEI 00-04 (i.e., Sections 2 through 12) is integral to providing reasonable confidence" and that "all aspects of NEI 00-04 must be followed to achieve reasonable confidence in the evaluations required by§ 50.69(c)(1)(iv)." However, neither RG 1.201 nor NEI 00-04 prescribe a particular sequence for completing the categorization elements. Therefore, the order in which each element of the categorization process (listed below) is completed is flexible as long as all elements are completed; they may even be performed in parallel. Note that NEI 00-04 only requires Item 3 (seven qualitative criteria) to be completed for components or functions categorized as LSS by all other elements. Similarly, Item 4 (DID assessment) is only required for safety-related active components or functions categorized as LSS by all other elements.

1. PRA-based evaluations (e.g., the internal events, internal flooding, and fire PRAs)

2. Non-PRA approaches (e.g., other external events screening, and shutdown assessment)

3. Seven qualitative criteria in Section 9.2 of NEI 00-04

4. Defense-in-Depth (DID) assessment

5. Passive categorization methodology Categorization of SSCs will be completed per the NEI 00-04 process, as endorsed by RG 1.201, which includes the determination of safety significance through the various elements identified above. The results of these elements are used as inputs to arrive at a preliminary component categorization (i.e., HSS or LSS) that is presented to the Integrated Decision-Making Panel (IDP).

Note that the term "preliminary HSS or LSS" used in this application is synonymous with the NEI 00-04 term "candidate HSS or LSS." A component or function is preliminarily categorized as HSS if any element of the process results in a preliminary HSS determination in accordance with Table 3-1. The safety significance determination of each element is independent and therefore the sequence of the elements does not impact the resulting preliminary categorization of each component or function. Consistent with NEI 00-04, the categorization of a component or function will only be "preliminary" until it has been confirmed by the IDP. Once the IDP confirms that the categorization process was followed appropriately, the final RISC category can be assigned.

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 7 of 34 The IDP may direct and approve detailed categorization of components in accordance with NEI 00-04, Section 10.2. The IDP may always elect to change a preliminary LSS component or function to HSS, however the ability to change component categorization from preliminary HSS to LSS is limited. This ability is only available to the IDP for select process steps as described in NEI 00-04 and endorsed by RG 1.201. Table 3-1 summarizes the NEI 00-04 IDP limitations. The steps of the process are performed at either the function level, component level, or both. This is also summarized in Table 3-1. A component is assigned its final RISC category upon approval by the IDP.

Table 3-1: Categorization Evaluation Summary Categorization Step -

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

Not Allowed Yes Section 5.1 Fire, Seismic and Other Allowable No Risk (PRA External Events Base Case Component Modeled)

PRA Sensitivity Studies Allowable No Integral PRA Assessment -

Not Allowed Yes Section 5.6 Fire and Other External Component Not Allowed No Hazards Risk (Non-Seismic - Alternative Tier 2 modeled)

Approach Function/Component Allowed 1 No Shutdown - Section 5.5 Function/Component Not Allowed No Defense-Core Damage - Section 6.1 Function/Component Not Allowed Yes in-Depth Containment - Section 6.2 Component Not Allowed Yes Qualitative Considerations - Section 9.2 Function Allowable 2 N/A Criteria Passive Passive - Section 4 Segment/Component Not Allowed No Notes:

1 /DP consideration of seismic insights can also result in an LSS to HSS determination.

2 The assessments of the qualitative considerations are agreed upon by the !DP in accordance with NE/ 00-04, Section 9.2. In some cases, a 10 CFR 50.69 categorization team may provide preliminary assessments of the seven considerations for the IDP's consideration, however the final assessments of the seven considerations are the direct responsibility of the /DP

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 8 of 34 The seven considerations are addressed preliminarily by the 1 O CFR 50. 69 categorization team for at least the system functions that are not found to be HSS due to any other categorization step.

Each of the seven considerations requires a supporting Justification for confirming (true response) or not confirming (false response) that consideration. If the 10 CFR 50 69 categorization team determines that one or more of the seven considerations cannot be confirmed, then that function is presented to the !DP as preliminary HSS Conversely, if all the seven considerations are confirmed, then the function is presented to the /DP as preliminary LSS The System Categorization Document, including the justifications provided for the qualitative considerations, is reviewed by the !DP. The !DP is responsible for reviewing the preliminary assessment to the same level of detail as the 1 O CFR 50 69 team (i.e., all considerations for all functions are reviewed). The /DP may confirm the preliminary function risk and associated justification or may direct that tl be changed based upon their expert knowledge. Because the Qualitative Criteria are the direct responsibility of the /DP, changes may be made from preliminary HSS to LSS or from preliminary LSS to HSS at the discretion of the !DP. If the !DP determines any of the seven considerations cannot be confirmed (false response) for a function, then the final categorization of that function is HSS The mapping of components to system functions is used in some categorization process steps to facilitate preliminary categorization of components. Specifically, functions with mapped components that are determined to be HSS by the PRA-based assessment (i.e., internal events PRA or integral PRA assessment) or DID evaluation will be initially treated as HSS. However, NEI 00-04, Section 10.2, allows detailed categorization which can result in some components mapped to HSS functions being treated as LSS. Also, Section 4.0 discusses additional functions that may be identified (e.g., fill and drain) to group and consider potentially LSS components that may have been initially associated with an HSS function, but do not support the critical attributes of that HSS function. Note that certain steps of the categorization process are performed at a component level (e.g., Passive, Non-PRA-modeled hazards - see Table 3-1). Except for seismic, these components from the component level assessments will remain HSS (IDP cannot override) regardless of the significance of the functions to which they are mapped. Components having seismic functions may be HSS or LSS based on the IDP's consideration of the seismic insights applicable to the system being categorized. Therefore, if an HSS component is mapped to an LSS function, that component will remain HSS. If an LSS component is mapped to an HSS function, that component may be driven HSS based on Table 3-1 or may remain LSS. Given that Seabrook is a seismic Tier 2 (moderate seismic hazard) plant as defined in Reference 3, seismic considerations are not required to drive an HSS determination at the component level, but the IDP will consider available seismic information pertinent to the components being categorized and can, at its discretion, determine that a component should be HSS based on that information.

The following clarifications are applied to the NEI 00-04 categorization process:

The IDP will be composed of a group of at least five experts who collectively have expertise in plant operation, design (mechanical and electrical) engineering, system engineering, safety analysis, and PRA. At least three members of the IDP will have a minimum of five years of experience at the plant, and there will be at least one member of the I DP who has a minimum of three years of experience in the modeling and updating of the plant-specific PRA.

The IDP will be trained in the specific technical aspects and requirements related to the categorization process. Training will address, at a minimum, (1) the purpose of the categorization; (2) present treatment requirements for SSCs including requirements for

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 9 of 34 DBEs; (3) PRA fundamentals; (4) details of the plant-specific PRA including the modeling, scope, and assumptions, the interpretation of risk importance measures, and the role of sensitivity studies and the change-in-risk evaluations; and (5) the DID philosophy and requirements to maintain this philosophy.

The decision criteria for the IDP to categorize SSCs as HSS or LSS in accordance with

§ 50.69(f)(1) will be documented in NextEra procedures.

Decisions of the I DP will be arrived at by consensus.

Differing opinions will be documented and resolved, if possible. However, a simple majority of the panel is sufficient for final decisions regarding HSS and LSS.

Passive characterization will be performed using the processes described in Section 3.1.2 of this enclosure. Consistent with NEI 00-04, an HSS determination by the passive categorization process cannot be changed by the IDP.

An unreliability factor of 3 will be used for the sensitivity studies described in Section 8 of NEI 00-04. The factor of 3 was chosen as it is representative of the typical error factor of basic events used in the PRA model.

NEI 00-04, Section 7, requires assigning the safety significance of functions to be preliminary HSS if it is supported by an SSC determined to be HSS from the PRA-based assessment in Section 5, but does not require this for SSCs determined to be HSS from non-PRA-based, deterministic assessments in Section 5. This requirement is further clarified in the Vogtle Safety Evaluation (Reference 4) which states"... if any SSC is identified as HSS from either the integrated PRA component safety significance assessment (Section 5 of NEI 00-04) or the [DID] assessment (Section 6 of NEI 00-04),

the associated system function(s) would be identified as HSS."

Once a system function is identified as HSS, then all the components that support that function are preliminary HSS. The IDP must intervene to assign any of these HSS function components to LSS.

With regard to the criteria that consider whether the active function is called out or relied upon in the plant Emergency/Abnormal Operating Procedures, NextEra will not take credit for alternate means unless the alternate means are proceduralized and included in Licensed Operator training.

NextEra proposes to apply an alternative seismic approach to those listed in NEI 00-04 Sections 1.5 and 5.3, as discussed in Section 3.2.3 of this enclosure.

The risk analysis to be implemented for each hazard is described below:

Internal Event Risks: Internal events and internal flood PRA, as submitted to the NRC for the adoption of TSTF-505 (Reference 5).

Fire Risks: Fire PRA model, as submitted to the NRC for the adoption of TSTF-505 (Reference 5).

Seismic Risks: Alternative approach in EPRI 3002017583 (Reference 3) for Tier 2 plants and additional considerations discussed in Section 3.2.3 of this enclosure.

Other External Risks (e.g., tornados, external floods): Using the IPEEE screening process approved by NRC on May 2, 2001 (Reference 51). The other external hazards were determined to be insignificant contributors to plant risk.

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 10 of 34 Low Power and Shutdown Risks: Qualitative DID shutdown model for shutdown configuration risk management (CRM) based on the framework for DID provided in NUMARC 91-06, "Guidance for Industry Actions to Assess Shutdown Management" (Reference 6), which provides guidance for assessing and enhancing safety during shutdown operations.

A change to the categorization process that is outside the bounds specified above (e.g., change from the Tier 2 alternative seismic approach to a seismic PRA approach) will not be made without prior NRC approval. The SSC categorization process documentation will include the following elements:

1. Program procedures used in the categorization

2. System functions, identified and categorized with the associated bases

3. Mapping of components to support function(s)

4. PRA model results, including sensitivity studies

5. Hazards analyses, as applicable

6. Passive categorization results and bases

7. Categorization results including all associated bases and RISC classifications

8. Component critical attributes for HSS SSCs

9. Results of periodic reviews and SSC performance evaluations

10. IDP meeting minutes and qualification/training records for the IDP members 3.1.2 Passive Categorization Process For the purposes of 10 CFR 50.69 categorization, passive components are those components that have a pressure retaining function. Passive components and the passive function of active components will be evaluated using the Arkansas Nuclear One (ANO) Risk-Informed Repair/Replacement Activities (RI-RRA) methodology contained in Reference 7 consistent with the related Safety Evaluation (SE) issued by the NRC.

The RI-RRA methodology is a risk-informed safety classification and treatment program for repair/replacement activities for pressure retaining items and their associated supports. In this method, the component failure is assumed with a probability of 1.0 and only the consequence evaluation is performed. It additionally applies deterministic considerations (e.g., DID, safety margins) in determining safety significance. Component supports, if categorized, are assigned the same safety significance as the highest passively ranked component within the bounds of the associated analytical pipe stress model. Consistent with NEI 00-04, an HSS determination by the passive categorization process cannot be changed by the IDP.

The use of this method for a 1 O CFR 50.69 application was previously approved by the NRC as documented in the final SE for Vogtle, dated December 17, 2014 (Reference 4). The RI-RRA method as approved for use at Vogtle for 1 O CFR 50.69 does not have any plant specific aspects and is generic. It relies on the conditional core damage and large early release probabilities associated with postulated ruptures. Safety significance is generally measured by the frequency and the consequence of the event. However, this RI-RRA process categorizes components solely based on consequence, which measures the safety significance of the passive component given that it ruptures. This approach is conservative compared to including the rupture frequency in the categorization since this approach will not allow the categorization of SSCs to be affected by any

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 11 of 34 changes in frequency due to changes in treatment. The passive categorization process is intended to apply the same risk-informed process accepted by the NRC in Reference 7 for the passive categorization of Class 2, 3, and non-class components. This is the same passive SSC scope the NRC has conditionally endorsed in ASME Code Cases N-660 and N-662 as published in RG 1.147, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 15.

Both code cases employ a similar risk-informed safety classification of SSCs in order to change the repair/replacement requirements of the affected LSS components. All ASME Code Class 1 SSCs with a pressure retaining function, as well as supports, will be assigned HSS for passive categorization which will result in HSS for its risk-informed safety classification and cannot be changed by the IDP. Therefore, this methodology and scope for passive categorization is acceptable and appropriate for use at Seabrook for 10 CFR 50.69 SSC categorization.

3.2 Technical Adequacy Evaluation (10 CFR 50.69(b)(2)(ii))

The following sections demonstrate that the quality and level of detail of the processes used in categorization of SSCs are adequate. The PRA models described below have been peer reviewed and there are no PRA upgrades that have not been peer reviewed. The PRA models credited in this request are the same PRA models credited in the Seabrook application to adopt TSTF-505 (Reference 5).

3.2.1 Internal Events and Internal Flooding The Seabrook categorization process for the internal events and flooding hazard will use the plant-specific PRA model. The NextEra risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for Seabrook Unit 1. Attachment 2 of this enclosure identifies the applicable internal events and internal flooding PRA models.

3.2.2 Fire Hazards The Seabrook categorization process for fire hazards will use a peer reviewed plant-specific fire PRA model. The internal Fire PRA model was developed consistent with NUREG/CR-6850 and only utilizes methods previously accepted by the NRC. The NextEra risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for Seabrook. Attachment 2 of this enclosure identifies the applicable fire PRA model.

3.2.3 Seismic Hazards 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 1) summarizes, the use of other methods for determining SSC functional importance in the absence of a quantifiable PRA (such as Seismic Margin Analysis or IPEEE Screening) as part of an integrated, systematic process. For the Seabrook 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. The approach is specified in EPRI 3002017583, and includes the markups provided in Attachment 2 of References 10 and 11, as well as the considerations provided in this section.

Note: The discussion below pertaining to EPRI 3002017583 includes the markups provided in of References 1 O and 11.

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 12 of 34 EPRI 3002017583 (Reference 3) is an update to EPRI 3002012988, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," July 2018 (Reference 13) which was referenced in the NRG-issued amendment and SE for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, to implement 10 CFR 50.69, as noted in Reference 14.

The technical criteria in EPRI 3002017583 are unchanged from its predecessor report EPRI 3002012988.

This LAR incorporates by reference the Clinton Power Station, Unit 1 response to request for additional information (RAI) 'DRA/APLC RAI 03 - Alternate Seismic Approach' included in the letter dated November 24, 2020 (Reference 15), in particular, the response to the question regarding the differences between the initial EPRI report (3002012988) and EPRI 3002017583.

The proposed categorization approach for Seabrook 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, Seabrook is considered a Tier 2 site because the site Ground Motion Response Spectrum (GMRS) to Safe Shutdown Earthquake (SSE) comparison is above the Tier 1 threshold, but not high enough that the NRG required the plant to perform an SPRA to respond to Recommendation 2.1 of the Near-Term Task Force (NTTF) 50.54(f) letter (Reference 16). 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, including their RAI responses and amendments (References 17, 18, 19, 20, 21, 22, 23, 24, and 25) 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] PRA can identify the appropriate seismic insights to be considered with the other categorization insights by the [IDP] for the final HSS determinations.

At sites with moderate seismic demands (i.e., Tier 2 range) such as Seabrook, there is no need to perform more detailed evaluations to demonstrate the inherent seismic capacities documented in industry sources such as Reference 26. 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 Seabrook.

Test cases described in Section 3 of EPRI 3002017583, including their RAI responses and amendments (References 17, 18, 19, 20, 21, 22, 23, 24, and 25) demonstrated that there are very few, if any, SSCs that would be designated HSS for seismic unique reasons. The test cases

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 13 of 34 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 10 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. The IDP considers these seismic insights along with the other categorization insights when making the final HSS determinations.

The test case information from EPRI 3002017583, developed by other licensees, including Case Study A (Reference 27), Case Study C (Reference 28), and Case Study D (Reference 29), as well as the associated RAI responses and amendments (References 17, 18, 19, 20, 21, 22, 23, 24, and 25) that clarify aspects of the case studies, 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 Seabrook being a Tier 2 Plant As defined in EPRI 3002017583, Seabrook 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 10 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 1 O CFR 50.54(f) letter (Reference 16).

The NRC issued its final determination of licensee SPRAs in a letter dated October 27, 2015 (Reference 30). The letter informed power reactor licensees of the remaining seismic evaluations to be performed and specifically informed those licensees that would perform an SPRA. The letter stated:

If the seismic hazard exceedance, peak of the spectral acceleration, and the general estimation of the SCDF [seismic core damage frequency] 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.

Seabrook was identified as a site where an SPRA or Seismic Margin Analysis were no longer expected (Note 3 of Table 1 a in Reference 30).

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 14 of 34 As illustrated in Figure 1, the Seabrook GMRS, derived from the seismic hazard, exceeds the SSE (seismic design basis capability) between 1.0 Hz and 10 Hz. Note that Figure 1 presents the GMRS as determined by the NRC, when the staff reviewed information related to the reevaluated seismic hazard for Seabrook (References 31 and 32). NextEra's GMRS curve for Seabrook, labeled as "Licensee GMRS" in Figure 1, closely resembles the NRC's determination, referred to as "NRC GMRS." Both GMRS curves exceed the SSE in the upper portion of the response spectrum between 1.0 and 10 Hz.

Therefore, the Seabrook seismic hazard meets the criteria for Tier 2 from EPRI 3002017583. The basis for Seabrook 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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Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 15 of 34 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 10 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 26) 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 Seabrook 1 O 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. The System Categorization Document (SCD) that is presented to the IDP will include the identified plant seismic insights. It will also outline the applicability of the EPRI 3002017583 study as well as the basis for Seabrook 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 Seabrook 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., internal events, internal flooding, and internal fire for Seabrook) 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 Seabrook PRA models but

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 16 of 34 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 described in the SCD.

The categorization team will review available Seabrook plant-specific seismic insights and other resources such as those identified above. The objective 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 10 and 11, and as described in this request. The process is summarized in Figure 2.

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Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 18 of 34 Determination of seismic insights will make use of the FPI E 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 5b 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 5a/5b 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 (including internal flooding) PRA and fire PRA. Those remaining components proceed forward for inclusion of associated seismic surrogate events in the Tier 2 Adjusted PRA Model (Steps 5c/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 functional recoveries (e.g., Emergency Diesel Generator (EOG) and DC power recovery).

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 19 of 34 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 1 O 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 1 O of Section 2.3.1 of EPRI 3002017583 (including markups in References 10 and 11). 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 a means to consider potential impacts of seismic events in the categorization process.

If the Seabrook seismic hazard changes from medium risk (i.e., Tier 2) at some future time and the feedback process determines that a change from the proposed alternative seismic approach is warranted, 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 impact of the changes 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:

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 changes are warranted, they will be implemented through the NextEra corrective action program and NEI 00-04 Section 12.

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 20 of 34 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 Seabrook The Seabrook GMRS and SSE curves from the seismic hazard and screening response are shown in Figure 1 of this enclosure and are replicated from the seismic hazard and screening report (Reference 31). The NRC staff assessment of the Seabrook seismic hazard and screening response is documented in Reference 32. In Section 4.0 of Reference 32, the NRC concluded that the methodology used by NextEra adequately characterized the seismic hazard for the Seabrook 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 Seabrook, the specific seismic reviews prepared by NextEra and the NRC's assessments of those reviews are provided in the following licensing documents.

1. NTTF Recommendation 2.1 Seismic Hazard Screening (References 31 and 32)

2. NTTF Recommendation 2.1 Spent Fuel Pool assessment (References 33 and 34)

3. NTTF Recommendation 2.3 Seismic Walkdowns (References 35, 36, and 37)

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

(References 38 and 39)

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

5. NTTF Recommendation 2.1 Expedited Seismic Evaluation Process (ESEP) (References 40 and 41)

6. NTTF Recommendation 2.1 Seismic High Frequency Evaluation (References 42 and 43)

Technical Information Precedent By letter dated January 31, 2020, Exelon Generation Company, LLC (EGC) submitted a LAR (Reference 8) to allow for the implementation of the provisions of 1 O CFR 50.69 for LaSalle County Station (LaSalle), Units 1 and 2. Following the criteria in EPRI report 3002012988 (Reference 13), based on the GMRS-to-SSE comparison, the LaSalle site is considered a Tier 2 site, similar to the Seabrook 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 dated October 1, 2020, October 16, 2020, and January 22, 2021 (References 9, 10, and 11, respectively). Based on information provided in Reference 8, as modified by the EGC RAI responses in References 9, 10, and 11, the NRC issued the license amendments, approving the EGC request, on May 27, 2021 (Reference 12).

NextEra will follow the same alternative seismic approach in the 1 O CFR 50.69 categorization process for Seabrook as that which was approved by the NRC staff for LaSalle (Reference 12),

except for the site-specific LaSalle information (e.g., seismic capacity discussions, etc.). Seabrook site-specific seismic capacity information is described above herein. References 9, 10, and 11 are incorporated by reference into this amendment request as they provide additional supporting

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 21 of 34 bases for Tier 2 plants, such as Seabrook, to adopt the alternative seismic methodology for use in the 10 CFR 50.69 categorization process. Note that Reference 9 included a response to RAI

'APLC 50.69-RAI No. 12', which is not relevant to the alternative seismic approach.

In addition, References 14, 44, 45, and 46 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 Seabrook 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 10 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).

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 22 of 34 3.2.4 Other External Hazards External hazards were screened for applicability to Seabrook per a plant-specific evaluation in accordance with Generic Letter (GL) 88-20 (Reference 47) and updated to use the criteria in ASME/ANS PRA Standard RA-Sa-2009 (Reference 48). Attachment 4 of this enclosure provides a summary of the other external hazards screening results. Attachment 5 provides a summary of the progressive screening approach for external hazards. All other external hazards, except for seismic, were screened from applicability.

As part of the categorization assessment of other external hazards, an evaluation is performed to determine if there are components being categorized that participate in screened scenarios and whose failure would result in an unscreened scenario. Consistent with the flow chart in Figure 5-6 in Section 5.4 of NEI 00-04, these components would be considered HSS and cannot be overridden by the IDP. Table 3-2 provides a summary of the Seabrook SSCs (doors) credited in the external hazards screening. See Attachment 4 for additional information.

Table 3-2: Summary of Components Credited in External Hazards Screening Hazard (Att. 4 Summary of Seabrook Doors Credited in External Hazards Screening Reference)

External Flooding The Mitigating Strategies Assessment Report (Reference 54), Attachment A, (LIP) provides a table with the critical plant areas impacted by Local Intense Precipitation (LIP) or Probable Maximum Storm Surge (PMSS) flooding and their Hurricane associated flood doors. These doors will be reviewed as part of categorization if (PMSS) their associated systems are categorized.

There are over 30 doors that are verified closed, per plant procedures Extreme Wind or (References 55 and 56), in the event of a tornado watch or warning (e.g., the Tornado control room door is to be verified closed during a tornado warning). These doors will be reviewed as part of categorization if their associated systems are categorized.

3.2.5 Low Power and Shutdown Consistent with NEI 00-04, the NextEra categorization process will use the shutdown safety management plan described in NUMARC 91-06 (Reference 6) for evaluation of safety significance related to low power and shutdown conditions. The overall process for addressing shutdown risk is illustrated in Figure 5-7 of NEI 00-04.

NUMARC 91-06 specifies that a DID approach should be used with respect to each defined shutdown key safety function. The key safety functions defined in NU MARC 91-06 are evaluated for categorization of SSCs.

SSCs that meet either of the two criteria (i.e., considered part of a "primary shutdown safety system" or a failure would initiate an event during shutdown conditions) described in Section 5.5 of NEI 00-04 will be considered preliminary HSS.

Seabrook Station Docket No. 50-443 3.2.6 PRA Maintenance and Updates L-2025-041 Enclosure Page 23 of 34 The NextEra risk management process ensures that the applicable PRA models used in this application continue to reflect the as-built and as-operated plant. The process delineates the responsibilities and guidelines for updating the PRA models and includes criteria for both regularly scheduled and interim PRA model updates. The process includes provisions for monitoring potential areas affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, and industry operational experience) for assessing the risk impact of unincorporated changes, and for controlling the model and associated computer files. The process will assess the impact of these changes on the plant PRA model in a timely manner but no longer than once every two refueling outages. If there is a significant impact on the PRA model, the SSC categorization will be re-evaluated.

In addition, NextEra will implement a process that addresses the requirements in NEI 00-04, Section 11, "Program Documentation and Change Control." The process will review the results of periodic and interim updates of the plant PRA that may affect the results of the categorization process. If the results are affected, adjustments will be made as necessary to the categorization or treatment processes to maintain the validity of the processes. In addition, any PRA model upgrades will be peer reviewed prior to implementing those changes in the PRA model used for categorization.

3.2.7 PRA Uncertainty Evaluations Uncertainty evaluations associated with any applicable baseline PRA model(s) used in this application were evaluated during the assessment of PRA technical adequacy and confirmed through the self-assessment and peer review processes as discussed in Section 3.3 of this enclosure.

Uncertainty evaluations associated with the risk categorization process are addressed using the processes discussed in Section 8 of NEI 00-04 and in the prescribed sensitivity studies discussed in Section 5.

In the overall risk sensitivity studies, NextEra will utilize a factor of 3 to increase the unavailability or unreliability of LSS components consistent with that approved for Vogtle in Reference 4.

Consistent with the NEI 00-04 guidance, NextEra will perform both an initial sensitivity study and a cumulative sensitivity study. The initial sensitivity study applies to the system that is being categorized. In the cumulative sensitivity study, the failure probabilities (unreliability and unavailability, as appropriate) of all LSS components modeled in all identified PRA models for all systems that have been categorized are increased by a factor of 3. This sensitivity study together with the periodic review process assures that the potential cumulative risk increase from the categorization is maintained acceptably low. The performance monitoring process monitors the component performance to ensure that potential increases in failure rates of categorized components are detected and addressed before reaching the rate assumed in the sensitivity study.

Additional information is provided in Attachment 6, including a description of the Seabrook PRA model-specific assumptions and sources of uncertainty and their impact on 10 CFR 50.69 categorization.

Seabrook Station Docket No. 50-443 3.3 PRA Review Process Results (1 O CFR 50.69(b)(2)(iii))

L-2025-041 Enclosure Page 24 of 34 The PRA models described in Section 3.2 have been assessed against RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2 (Reference 49) consistent with NRC RIS 2007-06 (Reference 50).

Facts and observations (F&O) closure reviews were conducted on the PRA models discussed in this section. Closed findings were reviewed and closed using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13 (Reference 52) as accepted by the NRC in a letter dated May 3, 2017 (Reference 53). The results of this review have been documented and are available for NRC audit.

Internal Events and Internal Flood PRA Models The Seabrook internal events and internal flood PRA models were peer reviewed in August 2005 (full scope), in January 2010 (internal flood focused scope), and in June 2012 (LERF focused scope) applying NEI 05-04, the ASME/ANS PRA Standard and RG 1.200, Revision 2. The purpose of these reviews was to provide a method for establishing the technical acceptability of the PRA for the spectrum of potential risk-informed plant licensing applications for which the PRA may be used.

A closure review of open findings was completed in October 2017; a total of 30 findings were reviewed and 26 were clos*ed.

A focused-scope review was conducted in April 2020 in accordance with the process documented in Appendix X to NEI 05-04, 07-12, and 12-13 as well as the requirements published in the ASME/ANS PRA Standard and RG 1.200, Revision 2. For each closed F&O, the resolution was assessed to determine if the PRA met the Capability Category II requirements of the ASME PRA Standard's SRs that were referenced in the F&O. A specific evaluation was also provided for each closed F&O to document whether the review team considered the F&O resolution a "PRA update" or a "PRA upgrade".

Following the independent assessment, one finding (LE-D6-01) remained open. Attachment 3 discusses the open finding with its disposition. The "Open" peer review finding, related to thermally-induced steam generator tube rupture, will be addressed as an implementation item to meet the requirements for PRA technical acceptability for this application.

Fire PRA Model The Seabrook internal fire PRA model was peer reviewed in September 2023 (full scope),

applying the AS ME/ANS PRA Standard and RG 1.200, Revision 2. The purpose of this review was to provide a method for establishing the technical acceptability of the PRA for the spectrum of potential risk-informed plant licensing applications for which the PRA may be used.

An independent assessment was conducted in August 2024 in accordance with the process documented in NEI 17-07 as well as the requirements published in the ASME/ANS PRA Standard and RG 1.200, Revision 2. For each closed F&O, the resolution was assessed to determine if the PRA met the Capability Category II requirements of the ASME PRA Standard's SRs that were referenced in the F&O. A specific evaluation was also provided for each closed F&O to document whether the review team considered the F&O resolution a "PRA update" or a "PRA upgrade".

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 25 of 34 Following the independent assessment team review, five (5) internal fire findings remained Not Assessed. discusses the six (6) findings designated as "Not Assessed" or "Partially Closed" with their dispositions. The "Partially Closed" peer review finding, related to instrumentation power supplies and interlocks, will be addressed as an implementation item to meet the requirements for PRA technical acceptability for this application.

3.4 Risk Evaluations (10 CFR 50.69(b)(2)(iv))

The NextEra 10 CFR 50.69 categorization process will implement the guidance in NEI 00-04. The overall risk evaluation process described in the NEI guidance addresses both known degradation mechanisms and common cause interactions and meets the requirements of§ 50.69(b)(2)(iv).

Sensitivity studies described in NEI 00-04, Section 8, will be used to confirm that the categorization process results in acceptably small increases to core damage frequency (CDF) and large early release frequency (LERF). The failure rates for equipment and initiating event frequencies used in the PRA include the quantifiable impacts from known degradation mechanisms, as well as other mechanisms (e.g., design errors, manufacturing deficiencies, and human errors). Subsequent performance monitoring and PRA updates required by the rule will continue to capture 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 timely 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.

The performance monitoring process will be described in NextEra's 1 O 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, PRA group, 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.

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 Section 3.2.3 of this enclosure, implementation of NextEra's design control and corrective action programs provide assurance that the inputs for qualitative determinations regarding seismic factors remain valid, thereby maintaining compliance with the requirements of 10 CFR 50.69(e).

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

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 26 of 34 systems that have been categorized in accordance with 10 CFR 50.69 to ensure that physical changes to the plant or changes to plant documents are evaluated.

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 10 CFR 50.69 categorization process will be identified and addressed through the problem identification and corrective action program, including seismic-related issues.

The NextEra 1 O 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 are completed at least once every two refueling cycles and 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 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.

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 27 of 34 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 includes 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 also monitors 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 in Title 10 of the Code of Federal Regulations (10 CFR) Part 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.

The proposed change is consistent with the applicable regulations and regulatory guidance.

4.2 No Significant Hazards Consideration Analysis NextEra Energy Seabrook, LLC (NextEra) proposes to modify the licensing basis for Seabrook Station, Unit 1, to allow for the voluntary implementation of the provisions of Title 1 O of the Code of Federal Regulations (1 O CFR), Part 50.69, "Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors." The provisions of 1 O CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

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:

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 28 of 34

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

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of structures, systems and components (SSCs) subject to NRG special treatment requirements and to implement alternative treatments per the regulations. The process used to evaluate SSCs for changes to NRG special treatment requirements and the use of alternative requirements ensures the ability of the SSCs to perform their design function. The potential change to special treatment requirements does not change the design and operation of the SSCs. As a result, the proposed change does not significantly affect any initiators to accidents previously evaluated 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 change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRG 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 change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRG 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.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 29 of 34 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 Conclusions 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.

6.

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 1 O 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.

Seabrook Station Docket No. 50-443

6.

REFERENCES L-2025-041 Enclosure Page 30 of 34

1.

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

Revision 0, July 2005 (ML052910035)

2.

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)

3.

Electric Power Research Institute (EPRI) 3002017583, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," Technical Update, February 2020 (ML21082A170)

4.

NRC letter to Southern Nuclear, "Vogtle Electric Generating Plant, Units 1 and 2 -

Issuance of Amendments Re: Use of 10 CFR 50.69," December 17, 2014 (ML14237A034)

5.

NextEra (Seabrook) letter to NRC, "License Amendment Request 25-01, Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, 'Provide Risk Informed Extended Completion Times - RITSTF Initiative 4b'," February 3, 2025 (ML25034A143)

6.

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

December 1991 (ML14365A203)

7.

NRC letter to Entergy Operations, Inc., "Arkansas Nuclear One, Unit 2 - Approval of Request for Alternative AN02-R&R-004, Revision 1, Request to Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Systems," April 22, 2009 (ML090930246)

8.

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)

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 1, 2020 (ML20275A292)

10. 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 10 CFR 50.69, 'Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors',"

October 16, 2020 (ML20290A791)

11. 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)

12. 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)

13. Electric Power Research Institute (EPRI) 3002012988, "Alternative Approaches for Addressing Seismic Risk in 1 O CFR 50.69 Risk-Informed Categorization," July 2018

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 31 of 34

14. 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)

15. 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)

16. 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)

17. Exelon Generation Company (Peach Bottom) Letter to NRC, "Seismic Probabilistic Risk Assessment Report, Response to 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," August 28, 2018 (ML18240A065)

18. 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)

19. 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)

20. 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)

21. 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)

22. 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 1 O of the Code of Federal Regulations 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," June 30, 2017 (ML17181A485)

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

April 10, 2018 (ML18100A966)

24. 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)

25. 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 1 O of the Code of Federal Regulations Section 50.69, 'Risk-Informed Categorization and Treatment of

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 32 of 34 Structures, Systems, and Components for Nuclear Power Plants'," April 30, 2020 (ML20076A194)

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

27. 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)

28. 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 10 CFR 50.69 Categorization Process Response to Request for Additional Information (RAls 4-11)," February 21, 2018 (ML18052B342)

29. Tennessee Valley Authority Letter to NRC, "Watts Bar Nuclear Plant, Units 1 and 2, Application to Adopt 10 CFR 50.69, 'Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors'," November 29, 2018 (ML18334A363)

30. 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)

31. NextEra Letter SBK-L-14052 to NRC, "NextEra Energy Seabrook, LLC Seismic Hazard and Screening 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," March 27, 2014 (ML14092A413)

32. NRC Letter to NextEra, "Seabrook Station, Unit 1 - Staff Assessment of Information Provided Pursuant to Title 1 O 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 12, 2015 (ML15208A049)

33. NextEra (Seabrook) Letter to NRC, "Spent Fuel Pool Integrity Evaluation Summary Supplemental Report in Response to 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, November 2, 2017 (ML17307A016)

34. NRC Letter to NextEra, "Seabrook Station, Unit 1 - Staff Review of Spent Fuel Pool Evaluation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1: Seismic," April 30, 2018 (ML18115A509)

35. NextEra (Seabrook) Letter SBK-L-12242, "Response to 10 CFR 50.54(f) Request for Information Regarding Near-Term Task Force Recommendation 2.3, Seismic,"

November 26, 2012 (ML12340A486 / ML12340A487)

36. NextEra (Seabrook) Letter SBK-L-13159 to NRC, "Supplement to NRC 10 CFR 50.54(f)

Request for Information Regarding Near-Term Task Force Recommendation 2.3, Seismic," September 30, 2013 (ML13275A202)

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 33 of 34

37. NRC Letter to NextEra, "Seabrook Station, Unit 1 - U.S. Nuclear Regulatory Commission 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," January 24, 2014 (ML13364A311)

38. NextEra (Seabrook) Letter SBK-L-17135 to NRC, "Seismic Mitigating Strategies Assessment (MSA) Report for the Reevaluated Seismic Hazard Information - NEI 12-06, Appendix H, Revision 4, H.4.4 Path 4: GMRS < 2xSSE," August 28, 2017 (ML17241A140)

39. NRC Letter to NextEra, "Seabrook, Unit 1 - Staff Review of Mitigation Strategies Assessment Report of the Impact of the Reevaluated Seismic Hazard Developed in Response to the March 12, 2012, 50.54(f) Letter," August 29, 2018 (ML18236A191)

40. NextEra (Seabrook) Letter SBK-L-14229 to NRC, "NextEra Energy Seabrook, LLC Expedited Seismic Evaluation Process Report (CEUS Sites) Related to the 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," December 19, 2014 (ML14360A016)

41. NRC Letter to NextEra, "Seabrook Station, Unit 1 - Staff Review of Interim Evaluation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1," October 19, 2015 (ML15282A019)

42. NextEra (Seabrook) Letter SBK-L-17136 to NRC, "NextEra Energy Seabrook, LLC's Seismic High Frequency Confirmation Report for the Reevaluated Seismic Hazard Information, August 28, 2017 (ML17241A150)

43. NRC Letter to NextEra, "Seabrook, Unit 1 - Staff Review of High Frequency Confirmation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1," June 5, 2018 (ML18138A451)

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

45. Exelon Generation Company (Calvert Cliffs) Letter to NRC, "Response to Request for Additional Information Regarding the Application to Adopt 1 O CFR 50.69, 'Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors'," July 19, 2019 (ML19200A216)

46. Exelon Generation Company (Calvert Cliffs) Letter to NRC, "Revised Response to Request for Additional Information Regarding the Application to Adopt 1 O CFR 50.69,

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

47. Response to Generic Letter 88-20, Supplement 4, "Individual Plant Examination External Events Report for Seabrook Station," September 1992 (ML20118C090)

48. ASME/ANS RA-Sa-2009, Standard for Level I/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RA-S-2008, February 2009

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

Seabrook Station Docket No. 50-443 L-2025-041 Enclosure Page 34 of 34

50. NRC Regulatory Issue Summary 2007-06, "Regulatory Guide 1.200 Implementation,"

March 22, 2007 (ML070650428)

51. NRC letter, "Seabrook Station, Unit No. 1 - Individual Plant Examination of External Events (IPEEE)," Staff Evaluation Report (SER) dated May 2, 2001

52. NEI Letter to NRC, "Final Revision of Appendix X to NEI 05-04/07-12/12-16 [sic], Close-out of Facts and Obsetvations (F&Os)," February 21, 2017 (ML17086A431)

53. NRC letter to NEI, "U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 7-12, and 12-13, Close Out of Facts and Observations (F&Os)," May 3, 2017 (ML17079A427)

54. NextEra (Seabrook) Letter SBK-L-17084 to NRC, "Mitigating Strategies Assessment (MSA) Report for Impact of New Flooding Hazard Information on FLEX Strategies," June 14, 2017 (ML17166A001)

55. Seabrook Procedure OS1200.03, "Severe Weather Conditions," Revision 32

56. Seabrook Procedure ON1090.13, "Response to Natural Phenomena Affecting Plant Operations," Revision 29

Seabrook Station Docket No. 50-443 : List of Categorization Prerequisites L-2025-041 NextEra has established fleet procedures that outline the process for performing 10 CFR 50.69 categorization. These fleet procedures will be implemented at Seabrook prior to performing system categorizations on Seabrook systems. The fleet procedures to be implemented at Seabrook contain the elements/steps listed below.

Integrated Decision-Making Panel (IDP) member qualification requirements.

Qualitative assessment of system functions. System functions are qualitatively categorized as preliminary High Safety Significant (HSS) or Low Safety Significant (LSS) based on the seven criteria in Section 9 of NEI 00-04 (see Section 3.1 of the enclosure). Any component supporting an HSS function is categorized as preliminary HSS. Components supporting an LSS function are categorized as preliminary LSS.

Component safety significance assessment. Safety significance of active components is assessed through a combination of Probabilistic Risk Assessment (PRA) and non-PRA methods, covering all hazards. Safety significance of passive components is assessed using a methodology for passive components.

Assessment of defense-in-depth (DID) and safety margin. Safety-related components that are categorized as preliminary LSS are evaluated for their role in providing DID and safety margin and, if appropriate, upgraded to HSS.

Review by the IDP. The categorization results are presented to the IDP for review and approval. The IDP reviews the categorization results and makes the final determination on the safety significance of system functions and components.

Risk sensitivity study. For PRA-modeled components, an overall risk sensitivity study is used to confirm that the population of preliminary LSS components results in acceptably small increases to core damage frequency (CDF) and large early release frequency (LERF) and meets the acceptance guidelines of Regulatory Guide 1.17 4.

Periodic reviews are performed to ensure continued categorization validity and acceptable performance for those SSCs that have been categorized.

Documentation requirements per Section 3.1.1 of the enclosure.

Seabrook Station Docket No. 50-443 L-2025-041 : Description of PRA Models Used in Categorization Model Baseline CDF Baseline LERF Comments

(/yr)

(/yr)

The PRA models used to Internal Events 2.5E-06 1.1 E-07 support categorization under

§ 50.69 are the same models Internal that are being reviewed by the Flooding 1.8E-06 7.7E-10 NRC as part of adopting Technical Specifications Task Force Traveler 505, Rev. 2.

Internal Fire 3.0E-05 6.2E-07

Seabrook Station Docket No. 50-443 SR Category LE-D6 CC-1,11 & Ill Finding LE-D6-01 FQ-F1 CC-I, II & Ill Finding 04-012 SF-A2 CC-I, II & Ill Finding 06-003 : Disposition and Resolution of Open Peer Review Findings and Self-Assessments Open Items Other Affected Peer Review Findings Resolution SRs N/A Peer Review Suggestion:

Peer Review Plant Response:

The analysis did not consider an Based on a review of EPRI TR-107623 increased probability of a thermally-Rev. 1 and NUREG 1570, a sensitivity induced steam generator tube analysis was performed with the value rupture due to depressurized steam ofXSGTI increased from 0.001 to 0.1.

generators that may occur due to Based on the sensitivity analysis the secondary side conditions.

release category LE13A increased from

1. 7E-12/yr to 2.3E-10/yr, increasing total LERF by approximately 0.15 percent.

QU-F3 Peer Review Suggestion:

Peer Review Plant Response:

Seabrook uses tables that need to This is tracked in the open Item section be enhanced to provide sufficient of NAH-FIRE-TR-RQ-000001. There is basis for the relative importance of no planned update to add details to events. There is no documentation of the Importances in this supporting the various types of revision.

importances relative to operator There is no impact to 50.69 since this is action, CCF events, joint HFEs documentation only.

(combinations), components as examples.

N/A Peer Review Suggestion:

Peer Review Plant Response:

Add a section to Seabrook Fire PRA Documentation will be resolved in the documentation to address the future to resolve the finding for seismic-potential for diversion of fire interaction.

suppressants.

F&O related to Seismic Fire Interaction do not need to be resolved to support risk-informed applications given that the associated requirements are qualitative L-2025-041 Page 1 of 3 Impact on Application Minor Impact The PRA models will be revised using the value of XSGTI prior to program implementation.

No Impact Finding is Documentation Only.

No Impact Finding is associated with a qualitative assessment that will not impact 50.69 related risk analyses.

Seabrook Station Docket No. 50-443 SR Category SF-A3 CC-I, II &Ill Finding 06-004 SF-A5 CC-I, II & Ill Finding 06-004 HRA-E1 CC-I, II & Ill Finding 08-007 other Affected SRs N/A N/A N/A Peer Review Findings Peer Review Suggestion:

Provide a qualitative assessment of the potential for common-cause failure of multiple fire suppression systems due to the seismically induced failure. The documentation needs to be updated to address if the redundant equipment is in the same location or have the same orientation such that it is not subject to a common cause failure in a seismic event.

Peer Review Suggestion:

Modify Seabrook documentation to state that the fire brigade is trained on the procedures.

Peer Review Suggestion:

Update the documentation in the HRA Calculator to remove discussion of the need for new MAAP analysis as cited in report.

Resolution only and will not impact risk supporting 50.69 categorizations.

Peer Review Plant Response:

F&O related to Seismic Fire Interaction do not need to be resolved to support risk-informed applications.

Documentation will be resolved in the future to resolve the finding for seismic-fire interaction.

Peer Review Plant Response:

F&O related to Seismic Fire Interaction do not need to be resolved to support risk-informed applications.

Documentation will be resolved in the future to resolve the finding for seismic-fire interaction.

Peer Review Plant Response:

The Internal Events HRA Calculator must be revised first. Documented as Open Item in Fire HRA Notebook.

L-2025-041 Page 2 of 3 Impact on Application No Impact Finding is associated with a qualitative assessment that will not impact 50.69 related risk analyses.

No Impact Finding is associated with a qualitative assessment that will not impact 50.69 related risk analyses.

No Impact Finding is Documentation Only, associated with updates of HRAC notes and not impacting HRAC HEP calculations.

Seabrook Station Docket No. 50-443 SR Category ES-B4 NOT MET Finding 03-002 Other Affected SRs ES-A2 PRM-B9 Peer Review Findings Peer Review Suggestion:

Perform a systematic review of all power supplies and interlocks identified for each credited equipment functional state, and document where in the PRA logic the power supplies and interlocks are either already included or were added.

A similar process should be applied for instrumentation power supplies as was performed for the power supply review for other components with power dependencies. A similar review for the Fire PRA components is needed to identify whether interlocks exist and under what gate they are modeled or dispositioned as to why they are not modeled.

Resolutlon Peer Review Plant Response:

The FSS Database Equipment List was revised to add columns for whether a power supply was required, the required power supply and the power supply modeling in the PRA logic model. The power supply logic was confirmed in the model. The interlocks were explicitly identified in the cable selection table and confirmatory notes added.

L-2025-041 Page 3 of 3 Impact on Application Finding is Partially Closed This is an implementation item that will be resolved prior to implementation of the 50.69 program.

The impacted is expected to be primarily related to enhanced documentation with a limited potential impact on risk results.

Seabrook Station Docket No. 50-443 External Hazard Aircraft Impact Avalanche Biological Event (Animal Infestation)

Biological Event (Aquatic Growth)

L-2025-041 Page 1 of 10 : External Hazards Screening Screening Result Screened?

Screening Criterion Comment (Y/N)

(Note a)

The aircraft crash hazard frequency analysis determined the risk associated with the aircraft crash hazard at Seabrook. It was updated to reflect more recent (i.e., through 2023) aircraft operations at Portsmouth International Airport, to include Conditional Core Damage Probabilities (CCDPs) from the current full power internal events (FPIE) probabilistic risk assessment (PRA), and to refine some other parts of the C1 calculation.

y Based on the updated calculations, Core Damage PS4 Frequency (CDF) and Large Early Release Frequency (LERF) from aircraft crashes are conservatively estimated to be approximately 1 E-7/yr and 8E-9/yr, respectively. Although the estimate of aircraft operations considers projections of future operations, significant increases in aircraft operations beyond those projected would be unlikely to result in CDF approaching 1 E-6/yr (i.e., operations would need to increase by a factor of 10).

Excluded due to site topography that would not y

C3 support snow buildup that would lead to an avalanche.

The hazard is included implicitly in the loss of C4 offsite power (LOOP) initiator. These are slow developing events and have limited impact.

y Preventative measures are in place through C5 station housekeeping and material control procedures.

Aquatic growth is a slow developing hazard that can be detected and managed. Plant programs C1 are in place to periodically inspect and clean the Service Water (SW) screen wash system.

y C3 Plant programs are in place to periodically inspect C5 and monitor heat exchangers cooled by SW.

Chlorine is injected into the intake tunnel to prevent biological growth. The design of SW is bounding for marine growth.

Seabrook Station Docket No. 50-443 External Hazard Biological Event (Organic Material in Water)

Coastal Erosion Drought External Flooding Screened?

Screening Criterion (YIN}

(Note a}

C1 y

C5 C1 y

C3 C1 y

C5 C1 y

C5 Screening Result Comment L-2025-041 Page 2 of 10 Storm induced clogging of SW screen wash and/or SW strainers have been analyzed. Storms large enough to cause intake clogging from debris have sufficient warning time. The SW screen wash system functions to remove such debris.

Excluded based on location of intake and discharge structures on the plant site. The SW intake is -50 feet below sea level and at a distance from the plant not subject to erosion.

The intake structure and pumphouses are approximately two miles from the shoreline. If the intake structures were to be damaged, the cooling towers would provide adequate backup cooling.

Excluded since there are two diverse Ultimate Heat Sinks (UHS) - the Atlantic Ocean and the Cooling Tower Basin, neither of which are impacted by drought. Drought is also slow in developing, allowing time to mitigate potential problems.

The Local Intense Precipitation (LIP) event results in standing water for a limited period of time outside of normally closed exterior doors. Water in-leakage was evaluated and determined to cause limited internal flooding. The majority of internal flooding is restricted to non-critical areas.

A minor amount of water enters the A ESWGR Room, A RHR Vault, and West MSFW Pipe Chase. The volume of water ingress was calculated and determined to not impact any safety related equipment. No anticipatory actions are required for LIP and permanently installed passive features (doors) prevent significant water ingress directly to critical areas. The available physical margins for critical areas of the plant are included in Table 1 "LIP APM" in the Focused Flooding Evaluation Summary (ML17181A409),

documenting that there are no locations with less than two inches of margin.

The NRG staff provided the following conclusion in the Staff Assessment of the Focused Evaluation (ML18039A920): "The staff verified the elevation and location of the critical SSCs, and confirmed the values used in the APM calculation.

Also, as documented in the MSA staff

Seabrook Station Docket No. 50-443 External Hazard Extreme Wind or Tornado Fog Forest or Range Fire Screened?

(Y/N) y y

y Screening Criterion (Note a)

C1 PS4 C4 C1 C3 C4 C5 Screening Result Comment L-2025-041 Page 3 of 10 assessment, the NRC staff finds that the licensee's estimation of water ingress and accumulation is reasonable."

Therefore, the LIP mechanism screens from further consideration. The plant design does not require manual actions other than ensuring the doors are closed prior to adverse weather and all the flood protection features are permanently installed and passive, leading to a high degree of reliability to prevent excessive water ingress to the plant.

Safety-related SSCs are protected in seismic Category I structures, which are designed to protect against tornado wind speeds up to 360 mph. These structures also protect SSCs from the effects of tornado missiles; several openings were evaluated using a probabilistic missile analysis.

In response to RIS 2015-06 (ML15020A419), a review of changes to Seabrook which could potentially affect the tornado missile design was completed. It was determined that there were no non-conformances to the tornado missile design and licensing basis.

The conservative estimate of missiles potentially damaging safety related SSCs is on the order of 1 E-6/yr or lower. Since additional failures or unavailability of SSCs would be needed to result in a core damage sequence and the tornado strike frequencies used in the probabilistic missile analysis are conservative, it is estimated that the CDF associated with missiles is less than 1 E-6/yr.

Additionally, there are over 30 "Tornado Doors" that need to be verified closed, per procedure, if a tornado watch or warning is in effect.

Fog and mist may increase the frequency of accidents involving aircraft, ships, or vehicles.

This weather condition is included implicitly in the accident rate data for these Transportation Accidents.

The hazard is included implicitly in the LOOP initiator. Forest and grass are somewhat distant from the plant and would not propagate to site equipment. Offsite fires from industrial or transportation accidents and toxic gas release

Seabrook Station Docket No. 50-443 External Hazard Frost Hail High Summer Temperature (Air)

High Summer Temperature (Water)

High Tide, Lake Level, or River Stage Hurricane Screened?

Screening Criterion (Y/N)

(Note a)

C1 y

C4 C1 y

C4 C1 y

C5 y

C5 C3 y

C4 C1 y

C5 Screening Result Comment L-2025-041 Page 4 of 10 should bound the impacts of smoke on personnel or equipment.

The hazard is included implicitly in the weather-related LOOP initiator.

Building design for high wind and missiles is bounding. The hazard is included implicitly in the weather-related LOOP initiator.

The air conditioning and ventilation systems within the plant are designed for extreme heat loads. Weather forecasts would provide warning of extended high temperatures. High temperatures would impact equipment reliability only over extended times. Preparations for summer-time operations are proceduralized.

The SW Cooling Tower operational procedure provides specific actions to monitor and reduce basin temperature. There is significant time for operator response based on weather forecasts.

The circulating water (CW) tunnel water temperature is not expected to exceed 65°F.

During the summer months, extended hot weather combined with ocean current changes can result in minor ocean temperature excursions above the 65°F design temperature threshold.

The plant is designed to continue operation up to a maximum ocean temperature of 68.5°F. The increase in ocean water temperature is a slow developing hazard, and there is sufficient time to transfer cooling modes to the cooling towers if the ocean temperature were to rise above 68.5°F.

Both high tide and river stage are included in the definition of hurricane and riverine flooding, respectively. They are not evaluated separately from these hazards. Seabrook is not located near a lake; therefore, lake level is not an applicable hazard to consider.

There are several exterior doors that are affected by the Probable Maximum Storm Surge (PMSS) wave run-up and exceed plant grade. Seabrook requires temporary flood barriers be deployed to maintain key safety features (KSFs) during major (Category 3) hurricane-induced PMSS flooding.

This involves the installation of temporary

Seabrook Station Docket No. 50-443 External Hazard Ice Cover Screened?

(YIN) y Screening Criterion

{Note a)

C1 C4 C5 Screening Result Comment L-2025-041 Page 5 of 10 sandbag dikes in 12 doorways, the installation of a flood gate in one doorway, and the sealing of floor drains.

Anticipatory installation of temporary flood barriers is performed by procedure when entry conditions are met. These measures were added to procedures in response to the Fukushima Dai-ichi accident.

An evaluation was performed that validated the time required to complete installation of temporary flood barriers (7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> and 44 minutes). Based on a warning time of approximately 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the evaluation determined there is sufficient time available to take these actions, with margin.

With the deployment of the temporary flood barriers, all KSFs and SSCs important to safety will remain available during the PMSS event and there are no postulated effects to the site.

Therefore, the PMSS mechanism can be screened from further consideration. The station has ample warning time to receive a hurricane warning (Category 3 or higher) and install the temporary flood barriers as outlined in procedures. Once the temporary barriers are installed, the effects of the flood are mitigated and there are no postulated effects on key SSCs during the flood event.

Ice cover on the ocean is not seen in the area and cooling water is not affected since the intake is

-50 feet below sea level. Flooding due to ice blockage is included in external flood if applicable; loss of cooling-water flow is considered in the plant design.

Operating procedures provide winter storm preparation actions and monitoring for the accumulation of snow and ice. The site has equipment to address snow and ice buildup to maintain equipment functionality and site accessibility. Forecasting allows additional time to prepare in the event of a winter storm.

The effect of ice on offsite power lines, switchyard, or offsite power transformers is included implicitly in weather-related or other LOOP initiators in the internal events PRA.

Seabrook Station Docket No. 50-443 External Hazard Industrial or Military Facility Accident Internal Flooding Internal Fire Landslide Lightning Low Lake Level or River Stage Low Winter Temperature (Air)

Low Winter Temperature (Water)

Screened?

Screening Criterion

{Y/N}

(Note a)

C1 y

C3 N

None N

None y

C3 C1 y

C4 C1 y

C3 C1 y

cs C1 y

C3 Screening Result Comment L-2025-041 Page 6 of 10 There are no military facilities within five miles of Seabrook. Explosions at nearby industrial facilities cannot affect the site, or the probabilities of these accidents are on the order of 1 E-7 or lower.

No nearby facilities were identified to house toxic chemicals that would affect control room habitability.

Also see Release of Chemicals in Onsite Storage and Toxic Gas.

The internal flooding PRA model addresses risks from internal flood events.

The internal fire PRA model addresses risks from internal fires.

The hazard is excluded due to site topography that would not support landslide of any significance.

The hazard is included implicitly in weather-related LOOP. The transmission lines are designed to have no more than one outage per 100 miles per year due to lightning. The plant grounding system provides protection to emergency AC power to reduce the likelihood of lightning-induced failure. Physical and electrical train separation provides additional protection and ensures that other hazard events (e.g., fire) are bounding.

This hazard screens out since the UHS is the ocean and the SW intake is -50 feet below the surface; therefore, low ocean level is not a concern.

Building structures, ventilation and monitoring systems are designed for low temperatures.

The seasonal readiness process prepares the site for reliable operation during sustained cold weather periods. Procedures provide guidance for extended cold temperatures.

The hazard is screened due to the intake's location, which is -50 feet below the surface of the ocean (not susceptible to freezing). The backup standby cooling tower has temperature monitoring and de-icing capabilities.

Seabrook Station Docket No. 50-443 External Hazard Meteorite or Satellite Impact Pipeline Accident Release of Chemicals in Onsite Storage River Diversion Sand or Dust Storm Screened?

Screening Criterion (Y/N)

{Note a)

PS4 y

C2 C2 y

C3 y

C1 y

C3 C1 y

C4 Screening Result Comment L-2025-041 Page 7 1of 10 Since the intake is -50 feet below the ocean's surface, frazil ice is also not a concern.

See also Ice Cover.

A conservative, bounding assessment shows that meteorite strikes can be screened. The estimated CDF contribution for this hazard is -3.6E-7/yr, which is less than 1 E-6/yr.

It is extremely unlikely for satellite debris of any significant size to hit or damage the site. Any such strike would be localized and is not expected to cause direct core damage or containment failure.

The satellite hazard is bounded by meteorite strike and other missiles because of the limited size of a satellite, comparable to a small meteorite, which is screened out.

There are no major pipelines that pass within five miles of the site. There's also not any major natural gas, fuel or other petroleum product storage facilities.

However, there is a gas feeder pipeline to the Seabrook area that may have an increase in pressure to meet customer demand. The potential impacts on plant operations from the increase in high-pressure to 200-psi was evaluated to have no impact from accidental release. The distance between the point of a postulated gas explosion to the plant is greater than the safe separation distance estimated in accordance with Regulatory Guide 1.91. The gas concentration buildup in the control room would be below the minimum needed to affect habitability.

As part of the station's control room envelope habitability program, a survey of onsite chemicals that can affect the control room is performed once per year. Based on the latest hazardous chemical evaluation walkdown conducted in 2024, there are no chemicals stored onsite that pose a threat to control room habitability.

Excluded since the UHS (the ocean) does not depend on a river.

Plant equipment is protected from or designed to preclude foreign material.

Seabrook Station Docket No. 50-443 External Hazard Seiche Seismic Activity Snow Soil Shrink-Swell Consolidation Storm Surge Toxic Gas Transportation Accident Screened?

Screening Criterion (YIN}

(Note a}

y C1 N

None C1 y

C5 C1 y

C3 C1 y

C5 C1 y

C3 C1 C2 y

C3 C4 Screening Result Comment L-2025-041 Page 8 of 10 This hazard was found to have damage potential less than that for which the plant is designed and needs no further consideration.

Seabrook is a Tier 2 Plant as defined by EPRI 3002017583. See Section 3.2.3 of the enclosure.

Plant design includes snow loads and other bounding loads. All safety-related structures are designed to withstand the force from "unusual" loads.

Emergency diesel generator air intakes are above ground elevation (51-ft elevation}, precluding snow buildup from affecting combustion air.

Plant procedure identifies the need to monitor ventilation air intakes that might be impacted by drifting snow or ice. Significant time would be available to mitigate blockage.

The hazard is excluded based on Seabrook's location on a "rock" site where soil shrink is not a concern.

See information related to Hurricane.

There's no significant quantity of toxic gases stored at industrial facilities in the vicinity of the site. Toxic gas protection for control room operators is not required since no potential chemical hazards were identified to create a hazardous environment in the control room.

Preventive maintenance tasks include 3-year offsite chemical surveys and 1-year onsite chemical surveys.

See also Release of Chemicals in Onsite Storage.

Impurity releases from ships cannot impact the SW ocean intake as it's -50 feet below the surface.

Shipping lanes are more than five miles away, thus only the intake and discharge structures for CW and SW may be impacted. These structures reach -50 feet below the water's surface at three miles from the plant, one mile from shore.

Navigational hazards (shallow water and rocks) prevent any vessel with a draft Qreater than five

Seabrook Station Docket No. 50-443 External Hazard Tsunami Turbine-Generated Missiles Volcanic Activity Screened?

Screening Criterion (Y/N)

(Note a) y C1 C1 y

PS4 y

C3 Screening Result Comment L-2025-041 Page 9 of 10 feet from passing over the intake structures, effectively preventing impact with the structures.

The potential impact of vehicle/ship explosion causing chemical release impacting the control room or pressure wave that fails safety-related structures is enveloped by industrial hazards (Toxic Gas, Industrial/Military Facility Accidents, Pipeline Accident).

The Guilford Rail System, Portsmouth-Hampton Branch, serves the areas within five miles of the plant site. The Rail System has no plans to rehabilitate the Portsmouth to Seabrook line beyond Hampton, NH, leaving the closest point to the plant as 3.1 miles. The Northern New England Passenger Rail Authority instituted passenger service in 2002 from Portland, Maine to Boston, Massachusetts. Even in the unlikely scenario that passenger service trains haul hazardous materials, the closest stopping point will be Exeter, NH, which is beyond five miles from the plant.

Vehicle crashes around transmission lines would have limited severity due to the guard rails and other barriers that protect the SF6 bus ducts from the switchyard to the transition yard. After that point, the transmission lines are on large poles on two separate rights-of-way.

This hazard was found to have damage potential less than that for which the plant is designed and needs no further consideration.

The turbine missile analysis concluded that the plant design and layout provide adequate protection from turbine missiles. Specifically, the probability of unacceptable damage to safety-related components is less than 1 E-7 /yr. The probability of a high trajectory missile strike on any single structure is less than 1E-7/yr, and many of these missile strikes will not result in unacceptable damage.

Additionally, a separate conservative analysis determined that CDF and LERF from turbine missiles are less than 1 E-6/yr and 1 E-7 /yr, respectively.

Excluded due to distance from nearest potentially active volcano.

Seabrook Station Docket No. 50-443 External Hazard Waves Screened?

(Y/N) y Screening Criterion (Note a)

C1 C5 Screening Result Comment L-2025-041 Page 10 of 10 See information related to Hurricane.

Note a - See Attachment 5 for descriptions of the screening criteria.

Seabrook Station Docket No. 50-443 L-2025-041 : Progressive Screening Approach for Addressing External Hazards Event Criterion Source Comments Analysis C1. Event damage potential is less NUREG/CR-2300 and than events for which plant is ASME/ANS Standard designed.

RA-Sa-2009 C2. Event has lower mean frequency NUREG/CR-2300 and and no worse consequences than ASME/ANS Standard other events analyzed.

RA-Sa-2009 Initial C3. Event cannot occur close enough NUREG/CR-2300 and ASME/ANS Standard Preliminary to the plant to affect it.

RA-Sa-2009 Screening Not used to C4. Event is included in the definition NUREG/CR-2300 and screen. Used of another event.

ASME/ANS Standard only to include RA-Sa-2009 within another event.

C5. Event develops slowly, allowing adequate time to eliminate or mitigate ASME/ANS Standard the threat.

PS1. Design basis hazard cannot ASME/ANS Standard cause a core damage accident.

RA-Sa-2009 PS2. Design basis for the event NUREG-1407 and meets the criteria in the NRC 1975 ASME/ANS Standard Standard Review Plan (SRP).

RA-Sa-2009 Progressive PS3. Design basis event mean Screening NUREG-1407 as frequency is less than 1 E-5/y and the modified in ASME/ANS mean conditional core damage Standard RA-Sa-2009 probability is less than 0.1.

PS4. Bounding mean CDF is less NUREG-1407 and ASME/ANS Standard than 1 E-6/y.

RA-Sa-2009 Detailed Screening not successful. PRA needs NUREG-1407 and PRA to meet requirements in the ASME/ANS Standard ASME/ANS PRA Standard.

RA-Sa-2009

Seabrook Station Docket No. 50-443 L-2025-041 Page 1 of 8 : Disposition of Key Assumptions/Sources of Uncertainty The Seabrook FPIE/IF (Full Power Internal Events and Internal Flood) and FPRA (fire PRA) models and documentation were reviewed for plant-specific modeling assumptions and related sources of uncertainty, and the applicable lists of EPRl-identified generic sources of uncertainty per EPRI 1016737 (Reference 1) and EPRI 1026511 (Reference 2) were also reviewed.

Each PRA model includes an evaluation of the potential sources of uncertainty for the base models using the approach that is consistent with the ASME/ANS RA-Sa-2009 (Reference 3) requirements for identification and characterization of uncertainties and assumptions. This evaluation identifies those sources of uncertainty that are important to the PRA results and may be important to PRA applications. The process meets the intent of steps C-1 and E-1 of NUREG-1855 (Reference 4).

These evaluations are documented for internal events and internal flooding in Section 17.0, "Sources of Uncertainty" in the Seabrook Station Probabilistic Safety Study (Reference 5). The results of the base PRA evaluations were reviewed to determine which potential uncertainties could impact the 10 CFR 50.69 categorization process results. This evaluation meets the intent of the screening portion of steps C-2 and E-2 of NUREG-1855, Revision 1.

Additionally, an evaluation of Level 2 internal events PRA model uncertainty was performed, based on the guidance in NUREG-1855 (Reference 4) and EPRI report 1026511 (Reference 2).

The potential sources of model uncertainty in the Seabrook Fire PRA model were evaluated for the 32 Level 2 PRA topics outlined in EPRI 1026511.

For the 1 O CFR 50.69 Program, the guidance in NEI 00-04 (Reference 6) specifies sensitivity studies to be conducted for each PRA model to address key sources of uncertainty. The sensitivity studies are performed to ensure that assumptions and sources of uncertainty (e.g.,

human error, common cause failure, and maintenance probabilities) do not mask the SSC(s) importance. RG 1.17 4, Revision 3 (Reference 7) cites NUREG-1855, Revision 1, as related guidance. In Section B of RG 1.174, Revision 3, the guidance acknowledges specific revisions of NUREG-1855 to include changes associated with expanding the discussion of uncertainties.

The results of the evaluation of PRA model sources of uncertainty as described above are evaluated relative to the 50.69 application to determine if additional sensitivity evaluations are needed.

Note: As part of the required 50.69 PRA categorization sensitivity cases directed by NEI 00-04, internal events/ internal flood and fire PRA models' human error and common cause basic events are increased to their 95th percentile and also decreased to their 5th percentile values.

In addition, maintenance unavailability terms are set to 0.0. For the fire PRA model only, a sensitivity case is required to allow no credit for manual suppression. These results are capable of driving a component and respective functions HSS and therefore the uncertainty of the PRA modeled Human Error Probabilities (HEPs) and CCFs are accounted for in the 50.69 application.

Seabrook Station Docket No. 50-443 L-2025-041 Page 2 of 8 The following table describes the internal events/ internal flooding (IE/ IF) PRA sources of model uncertainty and their impact.

IE/IF PRA IE/IF PRA IE/IFPRA Sources of Assumption!

10 CFR 50.69 Impact Model Sensitivity and Disposition Uncertainty (50.69)

Sump blockage: The Components or functions Modeling is judged to be reasonably incorporation of single for which emergency core conservative. Sensitivity studies indicated values for unrecoverable cooling systems are that the base model results are not failure due to sump credited significantly impacted by the current strainer plugging for all modeling assumptions.

sequences is a potential This will not be a key source of uncertainty source of uncertainty.

for the Seabrook 10 CFR 50.69 Program.

Internal Flood Initiating Components or functions Internal flooding is not a dominant Event Frequencies:

which are credited to contributor to the Seabrook risk profile.

Newer data is available mitigate flood scenarios Sensitivity studies indicated that base GDF to potentially include into (which is the same as for and LERF results are not significantly the Seabrook PRA all other hazard groups) impacted by the current assumptions model.

utilized. Scenarios that become more important can likely be further refined as needed when new frequencies are implemented.

This will not be a key source of uncertainty for the Seabrook 10 CFR 50.69 Program.

Induced Tube Rupture Components or functions The approach utilized in the model provides Probability: Seabrook for which LERF may be a best estimate assessment for the site.

Analysis uses older impacted Sensitivity studies indicated that the base methods as the basis for model results are not significantly impacted induced rupture by the current modeling assumptions.

probabilities.

This will not be a key source of uncertainty Newer data and for the Seabrook 10 CFR 50.69 Program.

methods are available to potentially include into the Seabrook PRA models.

Floor Value for Joint Components or functions The approach utilized in the model Human Error which are credited in long provides a best estimate assessment for Probabilities: A separate term loss of decay heat the site. Sensitivity studies indicate that the floor JHEP value is removal scenarios base case GDF and LERF results are not utilized for long-term significantly impacted by the current value loss of containment heat utilized.

removal in the PRA This will not be a key source of uncertainty models for Seabrook.

for the Seabrook 10 CFR 50.69 Program.

Seabrook Station Docket No. 50-443 IE/IF PRA Sources of Assumption/

Uncertainty No Credit for FLEX:

FLEX is not currently credited in the Seabrook PRA models.

IE/IF PRA 10 CFR 50.69 Impact Components or Functions for which FLEX provides an alternate success path IE/IF PRA L-2025-041 Page 3 of 8 Model Sensitivity and Dlsposttlon (50.69)

The approach utilized provides a slightly conservative treatment for the site.

Sensitivity studies indicated that the base case CDF and LERF results are not significantly impacted by the current assumptions utilized.

This will not be a key source of uncertainty for the Seabrook 10 CFR 50.69 Program.

Seabrook Station Docket No. 50-443 L-2025-041 Page 4 of 8 The following table describes the fire PRA sources of model uncertainty and their impact.

FlrePRA Fire PRA Sources of Uncertainty Fire PRA Disposition Description Analysis The primary goal of this task is to define fire Based on a review of the boundary and compartments whose boundaries are assumptions and potential partitioning expected to substantially contain the adverse sources of uncertainly effects of fire. There is some uncertainty associated with this element it is associated with the ability of compartment concluded that the methodology boundaries to confine fire (i.e., random for the Plant Boundary and failure of fire barriers, random failure of Partitioning task does not active suppression systems credited as introduce any epistemic partitioning features, etc.); however, this is uncertainties that would affect treated in the multi-compartment analysis.

the 10 CFR 50.69 calculations.

Plant boundary definition and partitioning is otherwise a deterministic task and there are no significant sources of uncertainty.

Component The primary goal of this task is to identify all A bounding sensitivity analysis Selection initiating events, accident sequence models, was performed to measure the and mitigating components to which fire risk associated with the always impact will be modelled by the FPRA. This is failed components. This generally a deterministic task and there are concluded the impact is limited.

no significant sources of uncertainty.

Based on the discussion of A number of systems and components were sources of uncertainty and the assumed failed for the FPRA. This is a discussion above, it is conservative assumption because not all fire concluded that the methodology scenarios are capable of failing these for the Equipment Selection systems. Accuracy and completeness task does not introduce any concerns are minimized by formal epistemic uncertainties that independent technical review of the notebook would require sensitivity and its associated attachments and inputs.

treatment.

Cable Selection Cable selection identifies all cables that, if Based on a review of the damaged by fire, could induce an initiating assumptions and potential event or fail component(s) required to sources of uncertainty related to mitigate fire-induced initiating events.

this element it is concluded that Cable selection also determines the routing of the methodology for the Cable Selection task does not each identified cable. This is generally a introduce any epistemic deterministic task and there are no significant uncertainties that would affect sources of uncertainty. Accuracy and the 10 CFR 50.69 calculations.

completeness concerns are minimized by formal independent technical review of the notebook and use of quality input documents (drawings, cable routing database, etc.) that are developed under the quality assurance program.

Seabrook Station Docket No. 50-443 FlrePRA Description Qualitative Screening Fire-Induced Risk Model Fire Ignition Frequency Quantitative Screening Fire PRA Sources of Uncertainty Qualitative screening element is an optional task whose objective is to identify physical analysis units whose potential fire risk contribution can be judged negligible without quantitative analysis.

No qualitative screening was implemented in the Seabrook Fire PRA.

This task develops a logic model and quantification process for calculating fire scenario CCDP and CLERP. The fire risk model effectively integrates all model and data uncertainty sources associated with each individual task. Most of the Fire PRA model structure is consistent with the internal events model and therefore does not present any significant uncertainties beyond those present in the internal events model.

Accuracy and completeness concerns are minimized by formal independent technical review of the notebook.

This task determines ignition source frequencies, which are based on apportioning generic fire frequencies developed from industry operating experience. A review of Seabrook fire events was performed, and an outlier experience was identified with Bin 24 which required Bayesian update. There was no outlier experience identified in the remaining fire ignition source bins suggesting the generic frequencies are appropriate for Seabrook for all bins except Bin 24. There is some data uncertainty associated with the generic frequencies themselves, and they are therefore provided as distributions.

The Seabrook Station fire PRA does not include fire compartment quantitative screening. All fire scenarios are retained in the fire PRA model.

L-2025-041 Page 5 of 8 Fire PRA Disposition Based on the discussion of source of uncertainty, it is concluded that the methodology for the Qualitative Screening task does not introduce any epistemic uncertainties that would affect the 10 CFR 50.69 calculations.

Based on a review of the assumptions and potential sources of uncertainty related to this, it is concluded that the methodology for the Plant Response Model task does not introduce any epistemic uncertainties that would affect the 10 CFR 50.69 calculations.

Based on a review of the assumptions and potential sources of uncertainty related to this element it is concluded that the methodology for the Fire Ignition Frequency task does not introduce any epistemic uncertainties that would affect the 10 CFR 50.69 calculations.

Consensus approaches are employed in the model.

Based on the discussion of source of uncertainty, it is concluded that the methodology for the Quantitative Screening task does not introduce any epistemic uncertainties that would affect the 10 CFR 50.69 calculations.

Seabrook Station Docket No. 50-443 FirePRA Description Scoping Fire Modeling Detailed Circuit Failure Analysis Circuit Failure Mode Likelihood Analysis Detailed Fire Modeling Fire PRA Sources of Uncertainty This task calculates severity factors, which represent the fraction of fires not capable of damaging targets external to the ignition source. The primary source of uncertainty is the peak HRR. The peak HRRs for each ignition source type are presented as distributions.

In the cable selection task, a generally conservative set of cables associated with each basic event is identified. The detailed circuit failure analysis is a refinement of the cable selection task. This task identifies the specific cables that will induce the specific failures of concern for risk-significant components. This is a deterministic task and there are no significant sources of uncertainty. Accuracy and completeness concerns are minimized by a formal independent technical review of the notebook.

This task applies conditional probabilities that particular failure modes will occur given that particular cables are damaged. There is some data uncertainty with the conditional probabilities provided in NUREG/CR-7150.

These values are provided with beta-distribution parameters and the values of three characteristics of this probability distribution. The parametric uncertainty analysis is documented in Section 5.5.19 of the Fire PRA Summary, Quantification, and Uncertainty Notebook (Reference 8).

Accuracy and completeness concerns are minimized by formal independent technical review of the notebook.

This task implements a fire progression event tree for each modelled ignition source which models the fire progression from ignition to growth, possible non-suppression at various stages, and potential propagation into adjacent compartments. Several sources L-2025-041 Page 6 of 8 Fire PRA Disposition Based on a review of the assumptions and potential sources of uncertainty related to this element, it is concluded that the methodology for the Scoping Fire Modeling task does not introduce any epistemic uncertainties that would affect the 10 CFR 50.69 calculations.

Based on a review of the assumptions and potential sources of uncertainty related to this element and the discussion above, it is concluded that the methodology for the Detailed Circuit Failure Analysis task does not introduce any epistemic uncertainties that would affect the 10 CFR 50.69 calculations.

The use of hot short failure probability and duration probability is based on fire test data and associated consensus methodology published in NUREG/CR-7150, Volume 2.

Based on a review of the assumptions and potential sources of uncertainty related to this element and the discussion above, it is concluded that the methodology for the Circuit Failure Mode Likelihood Analysis task does not introduce any epistemic uncertainties that would affect the 10 CFR 50.69 calculations.

Consensus modeling approaches are used for the Detailed Fire Modeling.

As such, the methodologies employed for the Detailed Fire

Seabrook Station Docket No. 50-443 FlrePRA Description Post-Fire Human Reliability Analysis Seismic-Fire Interactions Assessment Fire Risk Quantification Fire PRA Sources of Uncertainty of uncertainty are qualitatively discussed in Section 5.5.11 of the Seabrook FPRA Summary, Quantification, and Uncertainty Notebook (Reference 8).

This task includes identification of relevant Human Failure Events (HFEs), assessment of scenario specific impact on performance shaping factors, and calculation of scenario specific Human Error Probabilities (HEPs).

Note that every fire HFE is developed on a detailed basis, beyond just a screening HEP level. Most uncertainty is addressed by using a conservative bias (i.e., using the CBDTM/HCR method in the HRA calculator to estimate the HEP). Accuracy and completeness concerns are minimized by formal independent technical review of the notebook.

Since this is a qualitative evaluation, there is no quantitative impact with respect to the uncertainty of this task.

Seismic fire interaction has been qualitatively evaluated and therefore has no impact on fire risk quantification. For further details, see the Seismic-Fire Interactions Notebook.

The modelling tools used to quantify the fault tree-based risk model are subject to uncertainty due to the algorithms and assumptions inherent to the tools. Limitations of the software are identified in their respective configuration control reference and are either well known or nothing is available to address the uncertainty, so no sensitivity is performed. The steps used to generate the results of sensitivities are provided with the items modified. The steps provide sufficient information to recreate the sensitivity results, and therefore, the modified L-2025-041 Page 7 of 8 Fire PRA Disposition Modeling task do not introduce any epistemic uncertainties that affect the 10 CFR 50.69 calculations.

The HEPs include the consideration of degradation or loss of necessary cues due to fire. The fire risk importance measures indicate that the results are somewhat sensitive to HRA model and parameter values.

The Seabrook FPRA model HRA is based on industry consensus modeling approaches for its HEP calculations, so this is not considered a significant source of epistemic uncertainty.

10 CFR 50.69 applications already require assessment of the impact of operator action failure likelihood by assessing the 5% and 95% percentile impact of characterization.

The qualitative assessment of seismic induced fires should not be a source of model uncertainty as it is not expected to provide changes to the quantified FPRA model.

Based on the discussion of source of uncertainty, it is concluded that the methodology for the Fire Risk Quantification task does not introduce any epistemic uncertainties that would affect the 10 CFR 50.69 calculations.

Seabrook Station Docket No. 50-443 FlrePRA Description Uncertainty and Sensitivity Analyses FPRA Documentation REFERENCES Fire PRA Sources of Uncertainty files are not attached to the notebook.

Accuracy and completeness concerns are minimized by formal independent technical review of the notebook.

This task does not introduce any new uncertainties. This task is intended to address how the fire risk assessment could be impacted by the various sources of uncertainty.

This task does not introduce any new uncertainties to the fire risk as it outlines documentation requirements.

L-2025-041 Page 8 of 8 Fire PRA Dlsposttlon The Uncertainty and Sensitivity Analyses task does not introduce any epistemic uncertainties that would affect the 10 CFR 50.69 calculations.

The methodology for the FPRA documentation task does not introduce any epistemic uncertainties that would affect the 10 CFR 50.69 calculations.

1.

EPRI Technical Report -1016737, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments," December 2008

2.

EPRI Technical Update 1026511, "Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty," December 2012

3.

ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," February 2009

4.

NRC NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," Revision 1, March 2017 (ML17062A466)

5.

SSPSS-2019, "Seabrook Station Probabilistic Safety Study," September 2019

6.

NEI Topical Report NEI 00-04, "1 O CFR 50.69 SSC Categorization Guideline," Revision 0, July 2005 (ML052910035)

7.

NRC Regulatory Guide 1.17 4, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 3, January 2018 (ML17317A256)

8.

SBK-BFJR-23-042, "Seabrook Unit 1 At-Power Fire PRA, Qualitative Screening, Quantification, and Uncertainty Analysis Notebook," Revision 2, September 2024