LR-N25-0005, 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

ML25031A371
Person / Time
Site:	Salem   (DPR-070, DPR-075)
Issue date:	01/31/2025
From:	Sharbaugh D Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N25-0005, LAR S24-03
Download: ML25031A371 (1)

Text

David Sharbaugh Site Vice President - Salem Generating Station - PSEG Nuclear PO Box 236 Hancocks Bridge, New Jersey 08038-0221 david.sharbaugh@pseg.com LR-N25-0005 10 CFR 50.90 LAR S24-03 10 CFR 50.69 January 31, 2025 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311

Subject:

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 10 CFR 50.69 and 10 CFR 50.90, PSEG Nuclear LLC (PSEG) is requesting an amendment to the license of Salem Generating Station (Salem),

Units 1 and 2.

The proposed amendment would modify the Salem licensing basis, by the addition of a License Condition, to allow for the implementation of the provisions of 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." 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 Salem Operating License. The categorization process being implemented through this change is consistent with NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision 0, dated July 2005, which was endorsed by the 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. 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.

LR-N25-0005 10 CFR 50.90 Page 2 10 CFR 50.69 The PRA models described within this license amendment request (LAR) are the same as those described within the PSEG submittal of the LAR dated January 31, 2025, "License Amendment Request - Revise Salem Generating StationTechnical Specifications to Adopt Risk Informed Completion Times TSTF505-A, Revision 2, 'Provide RiskInformed Extended Completion Times - RITSTF Initiative 4b' and TSTF-591-A, Revision 0, 'Revise the Risk Informed Completion Time (RICT) Program'." PSEG requests that the NRC conduct their review of the PRA technical adequacy details for this application in coordination with the review of the application currently in-process. This would reduce the amount of PSEG and NRC resources necessary to complete the review of the applications. These requests should not be considered linked licensing actions, as the details of the PRA models in each LAR are complete which will allow the NRC staff to independently review and approve each LAR on their own merits without regard to the results from the review of the other.

PSEG requests approval of the proposed license amendment by February 28, 2026, with the amendment being implemented within 180 days following NRC approval.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the State of New Jersey.

This letter contains no new or revised regulatory commitments.

Should you have any questions concerning this submittal, please contact Shane Jurek at Shane.Jurek@pseg.com.

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

Executed on January 31, 2025.

Respectfully, David Sharbaugh Site Vice President - Salem Generating Station PSEG Nuclear Digitally signed by Sharbaugh, David L Date: 2025.01.31 12:22:34 -05'00'

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LR-N25-0005 10 CFR 50.90 Page 3 10 CFR 50.69

Enclosure:

1. Evaluation of the Proposed Change Attachments:

1. List of Categorization Prerequisites

2. Description of PRA Models Used in Categorization

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

4. Other External Hazards Disposition

5. Progressive Screening Approach for Addressing External Hazards

6. Disposition of Key Assumptions/Sources of Uncertainty cc:

Administrator, Region I, NRC NRC Project Manager, Salem NRC Senior Resident Inspector, Salem Manager, NJBNE

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

Enclosure LR-N25-0005 LAR S24-03 1

Evaluation of the Proposed Change Table of Contents 1

SUMMARY

DESCRIPTION.................................................................................................. 3 2

DETAILED DESCRIPTION................................................................................................... 3 2.1 CURRENT REGULATORY REQUIREMENTS............................................................ 3 2.2 REASON FOR PROPOSED CHANGE........................................................................ 3

2.3 DESCRIPTION

OF THE PROPOSED CHANGE......................................................... 4 3

TECHNICAL EVALUATION.................................................................................................. 5 3.1 CATEGORIZATION PROCESS DESCRIPTION (10 CFR 50.69(b)(2)(i)).................... 6 3.1.1 Overall Categorization Process...................................................................... 6 3.1.2 Passive Categorization Process................................................................... 11 3.2 TECHNICAL ADEQUACY EVALUATION (10 CFR 50.69(b)(2)(ii))............................ 12 3.2.1 Internal Events and Internal Flooding........................................................... 12 3.2.2 Fire Hazards................................................................................................. 13 3.2.3 Seismic Hazards.......................................................................................... 13 3.2.4 Other External Hazards................................................................................ 21 3.2.5 Low Power & Shutdown............................................................................... 21 3.2.6 PRA Maintenance and Updates................................................................... 21 3.2.7 PRA Uncertainty Evaluations....................................................................... 21 3.3 PRA REVIEW PROCESS RESULTS (10 CFR 50.69(b)(2)(iii))................................. 22 3.4 RISK EVALUATIONS (10 CFR 50.69(b)(2)(iv)).......................................................... 24 3.5 FEEDBACK AND ADJUSTMENT PROCESS............................................................ 24 4

REGULATORY EVALUATION........................................................................................... 25 4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA.................................... 25 4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS................................... 26

4.3 CONCLUSION

S......................................................................................................... 27 5

ENVIRONMENTAL CONSIDERATION.............................................................................. 27 6

REFERENCES................................................................................................................... 28

Enclosure LR-N25-0005 LAR S24-03 2

LIST OF ATTACHMENTS

List of Categorization Prerequisites...................................................................... 1 : Description of PRA Models Used in Categorization.............................................. 1 : Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items................................................................................................................................... 1 : Other External Hazards Disposition..................................................................... 17 : Progressive Screening Approach for Addressing External Hazards...................... 1 : Disposition of Key Assumptions/Sources of Uncertainty...................................... 8

Enclosure LR-N25-0005 LAR S24-03 3

1

SUMMARY

DESCRIPTION The proposed amendment modifies the licensing basis to allow for the implementation of the provisions of 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." 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 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 DBEs 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 for particular 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 50, is not explicitly defined.

2.2 REASON FOR 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 potential for common cause failures. The probabilistic approach to regulation is an

Enclosure LR-N25-0005 LAR S24-03 4

extension and enhancement of traditional regulation by considering risk in a comprehensive manner.

To take advantage of the safety enhancements available through the use of 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 controls (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) Topical Report NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline" (Reference [2]),

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 maintain functionality and reliability and is a function of the SSC categorization results and associated bases. Finally, periodic assessment activities are conducted to make adjustments to the categorization and/or treatment processes as needed so that SSCs continue to meet all applicable 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, though reduced, level of confidence that these SSCs will satisfy functional requirements.

Implementation of 10 CFR 50.69 will allow PSEG Nuclear LLC (PSEG) to improve focus on equipment that has safety significance resulting in improved plant safety.

2.3 DESCRIPTION

OF THE PROPOSED CHANGE PSEG proposes the addition of the following condition to the renewed operating license of Salem Generating Station, Units 1 and 2 (Salem), to document the NRC's approval of the use 10 CFR 50.69.

PSEG is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 SSCs using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive

Enclosure LR-N25-0005 LAR S24-03 5

component risk for Class 2 and Class 3 SSCs and their associated supports; the results of the non-PRA 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 as described in the PSEG submittal letter dated [DATE], and all its subsequent associated supplements 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 (e.g., change from a seismic margins approach to a seismic PRA approach).

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 PSEG submittal of the LAR to adopt Technical Specifications Task Force (TSTF) Traveler TSTF-505 (Reference [1]).

PSEG requests that the NRC conduct their review of the PRA technical adequacy details for this application in coordination with the review of the application currently in-process. This would reduce the amount of PSEG and NRC resources necessary to complete the review of the applications. These requests should not be considered linked licensing actions as the

Enclosure LR-N25-0005 LAR S24-03 6

details of the PRA models in each LAR are complete which will allow the NRC staff to independently review and approve each LAR on their own merits without regard to the results from the review of the other.

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

3.1.1 Overall Categorization Process PSEG 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 [3]). 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 evaluation of impact of the seismic hazard, which will use the Electric Power Research Institute (EPRI) Technical Report 3002017583 (Reference [4]) approach for seismic Tier 2 sites, which includes Salem, to assess seismic hazard risk for 10 CFR 50.69. Additional process steps discussed below address seismic considerations to ensure that reasonable confidence in the evaluations 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 or order for each of the elements to be completed.

Therefore, the order in which each of the elements of the categorization process (listed below) is completed is flexible and as long as they are all complete; they may even be performed in parallel. Note that NEI 00-04 only requires Item 3 to be completed for components/functions categorized as LSS by all other elements. Similarly, NEI 00-04 only requires Item 4 to be completed for safety-related active components/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 Figure 3-1 is an example of the major steps of the categorization process described in NEI 00-04; two steps (represented by four blocks on the figure) have been included to highlight review of seismic insights as pertains to this application, as explained further in Section 3.2.3:

Enclosure LR-N25-0005 LAR S24-03 7

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: the term "preliminary HSS or LSS" 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 below. The safety significance determination of each element, identified above, is independent of each other 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.

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 these IDP limitations in NEI 00-04. The steps of the process are performed at either the function level, component level, or both. This is also summarized in the Table 3-1. A component is assigned its final RISC category upon approval by the IDP.

Figure 3-1: Categorization Process Overview

Enclosure LR-N25-0005 LAR S24-03 8

Element Categorization Step - NEI 00-04 Section Evaluation Level IDP Change HSS to LSS Drives Associated Functions Risk (PRA Modeled)

Internal Events Base Case Section 5.1 Component Not Allowed Yes Fire, Seismic and Other External Events Base Case Allowed No PRA Sensitivity Studies Allowed No Integral PRA Assessment Section 5.6 Not Allowed Yes Risk (Non-modeled)

Fire, seismic and Other External Hazards Component Not Allowed No Seismic Function/Component Allowed2 No Shutdown Section 5.5 Function/Component Not Allowed No Defense-in-Depth Core Damage Section 6.1 Function/Component Not Allowed Yes Containment Section 6.2 Component Not Allowed Yes Qualitative Criteria Considerations Section 9.2 Function Allowed1 N/A Passive Passive Section 4 Segment/Component Not Allowed No Notes:

1 The assessments of the qualitative considerations are agreed upon by the IDP in accordance with Section 9.2. In some cases, a 10 CFR 50.69 categorization team may provide preliminary assessments of the seven considerations for the IDPs consideration, however, the final assessments of the seven considerations are the direct responsibility of the IDP.

The seven considerations are addressed preliminarily by the 10 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 Table 3-1: Categorization Evaluation Summary

Enclosure LR-N25-0005 LAR S24-03 9

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 IDP as preliminary HSS.

Conversely, if all the seven considerations are confirmed, then the function is presented to the IDP as preliminary LSS.

The System Categorization Document (SCD), including the justifications provided for the qualitative considerations, is reviewed by the IDP. The IDP is responsible for reviewing the preliminary assessment to the same level of detail as the 10 CFR 50.69 team (i.e. all considerations for all functions are reviewed). The IDP may confirm the preliminary function risk and associated justification or may direct that it be changed based upon their expert knowledge. Because the Qualitative Criteria are the direct responsibility of the IDP, changes may be made from preliminary HSS to LSS or from preliminary LSS to HSS at the discretion of the IDP. If the IDP determines any of the seven considerations cannot be confirmed (false response) for a function, then the final categorization of that function is HSS.

2 IDP consideration of seismic insights can also result in an LSS to HSS determination.

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; and 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 a HSS function but which 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 IDPs 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 above or may remain LSS. For the seismic hazard, given that Salem is a seismic Tier 2 (moderate seismic hazard) plant as defined in Reference [4], 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 are clarifications to be 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 IDP who has a minimum of three years of experience in the modeling and updating of the plant-specific PRA.

Enclosure LR-N25-0005 LAR S24-03 10 The IDP will be trained in the specific technical aspects and requirements related to the categorization process. Training will address at a minimum the purpose of the categorization; present treatment requirements for SSCs including requirements for DBEs; PRA fundamentals; details of the plant specific PRA including the modeling, scope, and assumptions, the interpretation of risk importance measures, and the role of sensitivity studies and the change-in-risk evaluations; and the DID philosophy and requirements to maintain this philosophy.

The decision criteria for the IDP for categorizing SSCs as HSS or LSS pursuant to 10 CFR 50.69(f)(1) will be documented in PSEG procedures.

Decisions of the IDP 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.

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 (SE) (Reference [5]) 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 defense-in-depth 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 considers whether the active function is called out or relied upon in the plant Emergency/Abnormal Operating Procedures, PSEG will not take credit for alternate means unless the alternate means are proceduralized and included in Licensed Operator training.

PSEG proposes to apply an alternative seismic approach to those listed in NEI 00-04, Sections 1.5 and 5.3. This approach is specified in EPRI 3002017583 (Reference [4]) for Tier 2 plants and is discussed in Section 3.2.3.

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

Internal Event Risks: Full Power Internal Events (FPIE), including internal flooding PRA, as submitted to the NRC for TSTF-505 (Reference [1]) (Refer to Attachment 2).

Enclosure LR-N25-0005 LAR S24-03 11 Fire Risks: Fire PRA (FPRA) model, as submitted to the NRC for TSTF-505 (Reference [1])

(Refer to Attachment 2).

Seismic Risks: EPRI Alternative Approach in EPRI 3002017583 for Tier 2 plants with the markups provided in Attachment 2 of References [6] and [7] and additional considerations discussed in Section 3.2.3 of this LAR.

Other External Risks (e.g., tornados, external floods): Using the Individual Plant Examination of External Events (IPEEE) screening process as approved on May 21, 1999 (ADAMS Accession No. ML18107A306). The other external hazards were determined to be insignificant contributors to plant risk.

Low Power and Shutdown Risks: Qualitative DID shutdown model for shutdown Configuration Risk Management based on the framework for DID provided in NUMARC 91-06, "Guidance for Industry Actions to Assess Shutdown Management" (Reference [8]), 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 a seismic margins approach to a seismic PRA (SPRA) approach) will not be used 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

Enclosure LR-N25-0005 LAR S24-03 12 Repair/Replacement Activities (RI-RRA) methodology contained in Reference [9] consistent with the related 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 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 was previously approved to be used for a 10 CFR 50.69 application by NRC in the SE for Vogtle (Reference [5]). The RI-RRA method as approved for use at Vogtle for 10 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 as this approach will not allow the categorization of SSCs to be affected by any 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 the ANO2-R&R-004 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 American Society of Mechanical Engineers (ASME) Code Cases N-660 and N-662 as published in RG 1.147, 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 RISC and cannot be changed by the IDP.

Therefore, this methodology and scope for passive categorization is acceptable and appropriate for use at Salem 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 described within this LAR are the same as those described within the PSEG submittal to adopt TSTF-505 (Reference [1]).

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

Enclosure LR-N25-0005 LAR S24-03 13 3.2.2 Fire Hazards The Salem categorization process for fire hazards will use a peer reviewed plant-specific FPRA model. The FPRA model was developed consistent with NUREG/CR-6850 and only utilizes methods previously accepted by the NRC. The PSEG risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for Salem. at the end of this enclosure identifies the applicable FPRA 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 [2])

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 Salem seismic hazard assessment, PSEG proposes to use a risk informed graded approach that meets the requirements of 10 CFR 50.69(b)(2) as an alternative to those listed in NEI 00-04, Sections 1.5 and 5.3. This approach is specified in EPRI 3002017583, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," (Reference [4]), and includes additional qualitative considerations that are discussed in this section1.

Note: The discussion below pertaining to Reference [4] includes the markups provided in Attachment 2 of References [6] and [7].

The proposed categorization approach for Salem is a risk-informed graded approach that is demonstrated to produce categorization insights equivalent to an SPRA. This approach relies on the insights gained from the SPRAs examined in Reference [4] and plant specific insights considering seismic correlation effects and seismic interactions. Following the criteria in Reference [4], the Salem site 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 NRC required the plant to perform an SPRA to respond to Recommendation 2.1 of the Near Term Task Force 50.54(f) letter (Reference [10]).

Reference [4] 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.

1 EPRI 3002017583 is an update to EPRI 3002012988, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," July 2018 (Reference [79] which was referenced in the NRC issued amendment and SE for Hope Creek (Reference [81] and Calvert Cliffs Nuclear Power Plant, Units 1 and 2, to implement 10 CFR 50.69 as noted below:

(1) Calvert Cliffs Nuclear Power Plant, Units 1 and 2, "Issuance of Amendment Nos. 332 and 310 Re:

Risk-Informed Categorization and Treatment of Systems, Structures, and Components,"

February 28, 2020. (ADAMS Accession No. ML19330D909) (Reference [80]).

(2) This LAR incorporates by Reference the Clinton Power Station, Unit 1 response to request for additional information letter of November 24, 2020 (ML20329A433) (Reference [78], in particular, the response to the question regarding the differences between the initial EPRI report 3002012988 and the current EPRI report 3002017583.

Enclosure LR-N25-0005 LAR S24-03 14 The trial studies in Reference [4], as amended by their RAI responses and amendments (References [11], [12], [13], [14], [15], [16], [17], [18], and [19]) show 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 Reference [4]:

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 FPIE PRA can identify the appropriate seismic insights to be considered with the other categorization insights by the Integrated Decision-making Panel (IDP) for the final HSS determinations.

At sites with moderate seismic demands (i.e., Tier 2 range) such as Salem, there is no need to perform more detailed evaluations to demonstrate the inherent seismic capacities documented in industry sources such as Reference [20]. 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 Salem.

Test cases described in Section 3 of Reference [4], as amended by their RAI responses and amendments (References [11], [12], [13], [14], [15], [16], [17], [18], and [19]) showed that there are very few, if any, SSCs that would be designated HSS for seismic unique reasons. The test cases identified that the unique seismic insights were typically associated with seismically correlated failures and led to unique HSS SSCs. While it would be unusual even for moderate hazard plants to exhibit any unique seismic insights, it is prudent and recommended by Reference [4] 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 Reference [4] uses common cause failures, similar to the approach taken in a FPIE PRA, and can identify the appropriate seismic insights to be considered with the other categorization insights by the IDP for the final HSS determinations.

PSEG is using test case information from Reference [4], developed by other licensees. The test case information is being incorporated by reference into this application, specifically Case Study A (Reference [21]), Case Study C (Reference [22]), and Case Study D (Reference [23]), as well as RAI responses and amendments (References [11], [12], [13], [14], [15]), [16], [17], [18], and

[19]) clarifying aspects of these case studies.

Basis for Salem being Tier 2 As defined in Reference [4], Salem 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.

Enclosure LR-N25-0005 LAR S24-03 15 Note: Reference [4] 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 50.54(f) letter (Reference [10]).

As shown in Figure A4-1, comparing the Salem GMRS (derived from the seismic hazard) to the SSE (i.e., seismic design basis capability), the GMRS is below the SSE up to approximately 5 Hz and then exceeds the SSE and then drops back below at approximately 50 Hz (Reference

[24]). The NRC screened out Salem from performing an SPRA in response to the Near-Term Task Force (NTTF) Recommendation 2.1 50.54(f) letter (Reference [25]). As such, it is appropriate that Salem is considered a Tier 2 plant. The basis for Salem being Tier 2 will be documented and presented to the IDP for each system categorized.

The following paragraphs provide 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 Reference [4] 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 (Reference [4]) 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 [20])

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.

Enclosure LR-N25-0005 LAR S24-03 16 In applying the Reference [4] process for Tier 2 sites to the Salem 10 CFR 50.69 categorization process, the IDP will be provided with the rationale for applying the Reference [4] guidance and informed of plant SSC-specific seismic insights that the IDP may choose to consider in their HSS/LSS deliberations. As part of the categorization team's preparation of the SCD that is presented to the IDP, a section will be included that provides identified plant seismic insights as well as the basis for applicability of the Reference [4] study and the bases for Salem 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 Salem is a Tier 2 plant At several steps of the categorization process, (e.g., as noted in Figure 3-1 and Table 3-1), 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 modeled hazards (i.e., internal events, including internal flooding, and internal fire for Salem) 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.

Seismic hazard risk impacts for SSC categorization will consider both qualitative and quantitative seismic risk insights. For SSCs not uniquely identified as HSS by the Salem FPIE PRA model or the PRA integral assessment but having design-basis functions during seismic events or functions credited for mitigation and prevention of severe accidents caused by seismic events, these will be addressed using non-PRA based qualitative assessments in conjunction with any seismic insights provided by the PRA. This is described in the ensuing paragraphs.

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

Impact of relay chatter Implications related to potential seismic interactions such as with block walls Seismic failures of passive SSCs such as tanks and heat exchangers Any known structural or anchorage issues with a particular SSC Components implicitly part of PRA-modeled functions (including relays)

Seismic insights identified from the review of Salem past seismic studies will then be considered along with a quantitative seismic risk sensitivity study for each system being categorized. For each system categorized, the categorization team will evaluate with a quantitative sensitivity study (using a seismically-biased PRA model based on the Salem FPIE PRA) seismically correlated failures and seismic interactions between SSCs. This quantitative seismic correlated failure sensitivity study process is detailed in Reference [4], Section 2.3.1 and is summarized below in Figure 3-2.

Enclosure LR-N25-0005 LAR S24-03 17 2 Reproduced from Reference [4] Figure 2-3 including the markups provided in Attachment 2 of References [6] and [7].

Figure 3-2: Seismic Correlation Failure Assessment for Tier 2 Plants2

Enclosure LR-N25-0005 LAR S24-03 18 The seismic correlated failure sensitivity study will make use of the FPIE PRA model adjusted to reflect seismic impacts and 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:

Identify SSCs within the system to be categorized Group SSCs within the system into seismic fragility classes of equipment and distributed systems used for SPRAs Refine the list and screen out the following SSCs from consideration of functional correlated seismic failures:

o Seismically inherently rugged components o Components not used in safety functions that support mitigation of core damage or containment performance o Components already identified as HSS components from the FPIE PRA or Integrated assessment Perform a seismic walkdown of the system SSCs:

o For SSCs screened IN look for seismic correlation o For SSCs screened IN or OUT look for spatial interaction configurations that could fail multiple components in the system, or could fail a single component in the system due to either seismic interaction or direct component failure modes, that result in total loss of a multi-train system and where there is not another system that independently provides the same function Based on results of the seismic walkdown, screen out from further evaluation:

o SSCs not subject to seismic correlation failure AND not subject to seismic interaction failure o SSCs with sufficiently high seismic capacity (i.e., such that they would be non-significant risk contributors to seismic risk) AND not subject to seismic interaction failure For the remaining unscreened SSCs, add seismic surrogate basic events to the FPIE model that simulate seismic interaction failures or seismic correlation failures (for the system being categorized) - set the probability of failure for each seismic surrogate basic event to 1E-04.

Quantify the FPIE model (for the system being categorized) for loss of offsite power (LOOP) and Small Loss of Coolant Accident (SLOCA) initiated accident sequences including the following seismically-biased adjustments: (1) set LOOP initiating event frequency to 1.0/year; (2) set SLOCA initiating event frequency to 1E-02/year; (3) set initiating event frequency for all other initiators to 0 (zero); and (4) remove probabilistic

Enclosure LR-N25-0005 LAR S24-03 19 credit for restoration of offsite power as well as for other functional recoveries modeled in LOOP and SLOCA initiated sequences.

Utilize the Importance Measures from this sensitivity study to identify appropriate SSCs (in the system being categorized) that should be HSS due to seismic correlation or seismic interactions 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 (e.g., Figure 3-1). The IDP cannot challenge any candidate HSS recommendation for any SSC from a seismic perspective if they believe there is a basis, except for certain conditions identified in Step 10 of Section 2.3.1 of Reference [4]. 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 FPRA as candidate HSS, which are not HSS from the FPIE 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 Salem seismic hazard changes from medium risk (i.e., Tier 2) at some future time, prior NRC approval, under 10 CFR 50.90, will be requested if Salem's feedback process determines that a process different from the proposed alternative seismic approach is warranted for seismic risk consideration in categorization under 10 CFR 50.69. After receiving NRC approval, PSEG will follow its categorization review and adjustment process to review the changes to the plant and update, as appropriate, the SSC categorization in accordance with 10 CFR 50.69(e) and the EPRI 3002017583 SSC categorization criteria for the updated Tier. This includes use of the PSEG corrective action program (CAP).

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

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

b) Seismic considerations for subsequent system categorization activities will be performed in accordance with the guidance in EPRI 3002017583 Section 2.2, "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, PSEG will implement the following process following completion of the SPRA, including adequate closure of Peer Review Findings and Observations (F&Os).

a) For previously completed system categorizations, PSEG will review the categorization results using the SPRA insights as prescribed in NEI 00-04 Section 5.3 and Section 5.6.

If changes are warranted, they will be implemented through the PSEG design control program and CAP and NEI 00-04 Section 12.

Enclosure LR-N25-0005 LAR S24-03 20 b) Seismic considerations for subsequent system categorization activities will follow the guidance in NEI 00-04, as recommended in EPRI 3002017583 Section 2.4, "High Seismic Hazard / Low Seismic Margin Sites".

Historical Seismic References for Salem The Salem GMRS and SSE curves from the seismic hazard and screening response are shown in Section 2.4 and 3.1, respectively, in the seismic hazard and screening report of Reference [26]. The Salem SSE and GMRS curves from Reference [26] are shown in Figure A4-1. The NRC's Staff assessment of the Salem seismic hazard and screening response is documented in Reference [25]. In the Staff Confirmatory Analysis (Section 3.3.3 of Reference

[25]), the NRC concluded that the methodology used by PSEG in determining the GMRS was acceptable and that the GMRS determined by PSEG adequately characterizes the reevaluated hazard for the Salem site.

Section 1.1.3 of Reference [4] cites various post-Fukushima seismic reviews performed for the U.S. fleet of nuclear power plants. For Salem, the specific seismic reviews prepared by the licensee and the NRCs staff assessments are provided here. These licensee documents were submitted under oath and affirmation to the NRC.

1. NTTF Recommendation 2.1 seismic hazard screening (References [26], [25]).

2. NTTF Recommendation 2.3 seismic walkdowns (References [27], [28]).

3. NTTF Recommendation 4.2 seismic mitigation strategy assessment (References

[29], [30]).

The following additional post-Fukushima seismic reviews were performed for Salem.

4. NTTF Recommendation 2.1 seismic high frequency evaluation (Reference [31]).

5. NTTF Recommendation 2.1 seismic spent fuel pool evaluation (References [32],

[33])

Summary Based on the above, the Summary from Section 2.3.3 of Reference [4] applies to Salem; namely, Salem 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. References [6], [7], and [34]3 are incorporated into this LAR as they provide additional supporting bases for Tier 2 plants. In addition, References [35],

[36], and [37] are incorporated into this LAR as they provide additional supporting bases for Tier 1 plants that is also used for Tier 2 plants. 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 EPRI approach outlined in Reference [4] 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).

3 Excludes RAI APLC 50.69-RAI No. 12 that addresses a non-seismic topic (external events).

Enclosure LR-N25-0005 LAR S24-03 21 3.2.4 Other External Hazards All external hazards, except for seismic, were screened for applicability to Salem per a plant-specific evaluation in accordance with Generic Letter 88-20 (Reference [38]) and updated to use the criteria in ASME PRA Standard RA-Sa-2009. Attachment 4 provides a summary of the external hazards screening results. Attachment 5 provides a summary of the progressive screening approach for external hazards.

3.2.5 Low Power & Shutdown Consistent with NEI 00-04, the Salem categorization process will use the shutdown safety management plan described in NUMARC 91-06 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 NUMARC 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.

3.2.6 PRA Maintenance and Updates The PSEG risk management process ensures that the applicable PRA models used in this application continue to reflect the as-built and as-operated plant for Salem. 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, PSEG will implement a process that addresses the requirements in NEI 00-04, Section 11. 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.

Enclosure LR-N25-0005 LAR S24-03 22 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, PSEG will utilize a factor of 3 to increase the unavailability or unreliability of LSS components consistent with that approved for Vogtle in Reference [5].

Consistent with the NEI 00-04 guidance, PSEG 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.

The detailed process of identifying, characterizing, and qualitative screening of model uncertainties is found in Section 5.2.1 of NUREG-1855 and Section 3.1.1 of EPRI TR-1016737 (Reference [39]). The process in these References was mostly developed to evaluate the uncertainties associated with the FPIE PRA model; however, the approach can be applied to other types of hazard groups.

Each PRA element notebook was reviewed for assumptions and sources of uncertainties. The characterization of assumptions and sources of uncertainties are based on whether the assumption and/or source of uncertainty is key to the 10 CFR 50.69 application in accordance with RG 1.200, Revision 2.

Key Salem PRA model specific assumptions and sources of uncertainty for this application were identified and dispositioned in Attachment 6. The conclusion of this review is that no additional sensitivity analyses are required to address Salem PRA model specific assumptions or sources of uncertainty.

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

The PRA models described in Section 3.2 have been assessed against RG 1.200, Revision 2 (Reference [40]), consistent with NRC Regulatory Issue Summary 2007-06.

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, Close-out of Facts and Observations (F&Os) (Reference [41]) as accepted by NRC in the letter dated May 3, 2017 (Reference [42]). The results of this review have been documented and are available for NRC audit.

Scope and Technical Acceptability of Salem Internal Events and Internal Flooding PRA Model A full-scope peer review of the Salem Units 1 and 2 FPIE PRA model was conducted in November 2008 (Reference [43]). In 2018 an assessment was performed which documented the resolution of F&Os that were generated during the 2008 peer review (Reference [44]). The assessment also documented gaps that existed between the 2005 version of the ASME PRA

Enclosure LR-N25-0005 LAR S24-03 23 Standard (Reference [45]) and NRC RG 1.200, Revision 1 (Reference [46]) that were used during the 2008 peer review.

The 2018 assessment was performed using the 2009 version of the ASME PRA Standard (Reference [47]) and RG 1.200 Revision 2 (Reference [40]). The F&O information and corresponding resolutions followed the guidance given in Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13 (Reference [41]). The resolutions were all deemed to be associated with update and maintenance of the PRA model.

The resolution of internal flood Findings was performed in 2011 (Reference [48] to address the peer review comments and Findings from the November 2008 peer review (Reference [43]). No revisions or corrections were required for the internal flood model and no outstanding issues remain against the internal flood model.

In November 2018, an F&O Independent Assessment and Focused-Scope Peer Review (Reference [49]) was performed in accordance with Appendix X of NEI 05-04 and NEI 07-12 to review closeout of Finding-level F&Os from the prior reviews and to perform a focused-scope peer review for selected Technical Elements, including those affected by an upgrade to the FPIE PRA model. In particular, a new uncertainty assessment was performed, including full propagation of probability distributions for basic events. The expanded uncertainty assessment also addressed sources of modeling uncertainty related to internal flooding.

In conclusion, for the FPIE PRA model, all Finding-Level F&Os are closed and all applicable supporting requirements from the ASME/ANS PRA Standard (ASME/ANS RA-Sa-2009)

(Reference [47]) are met with a Capability Category II or greater. PSEG performed a self-assessment as to whether the resolution of each Finding constituted maintenance or upgrade of the PRA, as defined in the ASME/ANS PRA standard. Ultimately, the F&O Independent Assessment Team concurred with the PSEG assessment that there were no PRA upgrades associated with the resolution of findings nor were there any open Findings associated with newly developed methods.

Scope and Technical Acceptability of Salem Fire PRA Model A full scope peer review of the Salem FPRA model was performed from October 2022 to January 2023 (Reference [50]). The scope reviewed the Salem Units 1 and 2 FPRA against Capability Category II of the Technical Elements in Part 4 of the ASME/ANS PRA Standard (Reference [47]), considering NRC clarifications and qualifications in RG 1.200, Revision 3 (Reference [51]). This peer review was performed using the process defined in NEI 17-07 (Reference [52]). The cable selection task was incomplete at the time of the review for both units, as well as the development of the PRA logic model and quantification for Unit 2. Both of these items have since been addressed and were closed out during a Focused-Scope Peer Review later in 2023 (Reference [53]). During this same effort, all Findings were closed and all supporting requirements were found to meet at least Capability Category II. The process and conclusions are documented in PWROG-23035-P, "Independent Assessment of Facts &

Observations Closure and Focused Scope Peer Review of the Salem Units 1 and 2 Fire Probabilistic Risk Assessment" (Reference [53]).

In conclusion, for the FPRA model, all Finding-Level F&Os are closed and all applicable supporting requirements from the ASME/ANS PRA Standard (ASME/ANS RA-Sa-2009)

(Reference [47]) are met with a Capability Category II or greater. PSEG performed a self-assessment as to whether the resolution of each Finding constituted maintenance or upgrade of

Enclosure LR-N25-0005 LAR S24-03 24 the PRA, as defined in the ASME/ANS PRA standard. Ultimately, the F&O Independent Assessment Team concurred with the PSEG assessment that there were no PRA upgrades associated with the resolution of findings nor were there any open Findings associated with newly developed methods.

This demonstrates that the PRA models are of sufficient quality and level of detail to support the categorization process and have been subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC as required by 10 CFR 50.69(c)(1)(i).

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

The Salem 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 10 CFR 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, an immediate evaluation and review will be performed prior to the normally scheduled periodic review. Otherwise, the assessment of potential equipment performance changes and new technical information will be performed during the normally scheduled periodic review cycle.

To more specifically address the feedback and adjustment (i.e., performance monitoring) process as it pertains to the proposed Salem Tier 2 approach discussed in Section 3.2.3, implementation of the PSEG design control program and CAP will ensure the inputs for the qualitative determinations for seismic continue to remain valid to maintain compliance with the requirements of 10 CFR 50.69(e).

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

The Salem 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.

Enclosure LR-N25-0005 LAR S24-03 25 Salem has a comprehensive problem identification and CAP that ensures that issues are identified and resolved. Any issue that may impact the 10 CFR 50.69 categorization process will be identified and addressed through the problem identification and CAP, including seismic-related issues.

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

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

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.

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

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

NRC RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, April 2015.

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

Enclosure LR-N25-0005 LAR S24-03 26 NRC RG 1.200, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 3, December 2020.

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

4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS PSEG Nuclear LLC (PSEG) proposes to modify the licensing basis of the Salem Generating Station, Units 1 and 2 (Salem) to allow for the voluntary implementation of the provisions of 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." 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.

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

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

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of Structures, Systems and Components (SSCs) subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The process used to evaluate SSCs for changes to NRC special treatment requirements and the use of alternative requirements 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.

Enclosure LR-N25-0005 LAR S24-03 27 The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not change the functional requirements, configuration, or method of operation of any SSC.

Under the proposed change, no additional plant equipment will be installed.

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

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

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not affect any Safety Limits or operating parameters used to establish the safety margin.

The safety margins included in analyses of accidents are not affected by the proposed change. The regulation requires that there be no significant effect on plant risk due to any change to the special treatment requirements for SSCs and that the SSCs continue to be capable of performing their design basis functions, as well as to perform any beyond design basis functions consistent with the categorization process and results.

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

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

4.3 CONCLUSION

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

5 ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Enclosure LR-N25-0005 LAR S24-03 28 6

REFERENCES

[1] PSEG Letter to NRC, "License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b'," dated January 31, 2025, (ADAMS Accession No. ML25031A359).

[2] NEI Topical Report NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision 0, dated July 2005 (ADAMS Accession No. ML052910035).

[3] NRC 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 (ADAMS Accession No. ML061090627).

[4] EPRI Technical Report 3002017583, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," dated September 2021 (ADAMS Accession No. ML21082A168).

[5] NRC letter to Southern Nuclear, "Vogtle Electric Generating Plant, Units 1 and 2 -Issuance of Amendments Re: Use of 10 CFR 50.69," dated December 17, 2014 (ADAMS Accession No. ML14237A034).

[6] Exelon 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," dated October 16, 2020 (ADAMS Accession No. ML20290A791).

[7] Exelon letter to NRC, "Response to Request for Additional Information Regarding the License Amendment Request to Adopt 10 CFR 50.69," dated January 22, 2021 (ADAMS Accession No. ML21022A130).

[8] NEI Topical Report NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown Management," dated December 1991 (ADAMS Accession No. ML14365A203).

[9] NRC letter to Entergy, "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,"

dated April 22, 2009 (ADAMS Accession No. ML090930246).

[10] 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," dated March 12, 2012 (ADAMS Accession No. ML12053A340).

Enclosure LR-N25-0005 LAR S24-03 29

[11] Exelon letter to NRC, "Seismic Probabilistic Risk Assessment Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near Term Task Force Review of Insights from the Fukushima Dai-ichi Accident,"

dated August 28, 2018 (ADAMS Accession No. ML18240A065).

[12] NRC letter to Exelon, "Staff Review of Seismic Probabilistic Risk Assessment Associated with Reevaluated Seismic Hazard Implementation of the Near-Term Task Force Recommendation 2.1: Seismic," dated June 10, 2019 (ADAMS Accession No. ML19053A469).

[13] NRC letter to Exelon, "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," dated October 8, 2019 (ADAMS Accession No. ML19248C756).

[14] Southern Nuclear letter to NRC, "License Amendment Request to Modify Approved 10 CFR 50.69 Categorization Process," dated June 22, 2017 (ADAMS Accession No. ML17173A875).

[15] NRC letter to Southern Nuclear, "Issuance of Amendments Regarding Application of Seismic Probabilistic Risk Assessment Into the Previously Approved 10 CFR 50.69 Categorization Process," dated August 10, 2018 (ADAMS Accession No. ML18180A062).

[16] TVA Letter to NRC, "Response to NRC Request for Information Pursuant to Title 10 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," dated June 30, 2017 (ADAMS Accession No. ML17181A485).

[17] TVA letter to NRC, "Seismic Probabilistic Risk Assessment Supplemental Information,"

dated April 10, 2018 (ADAMS Accession No. ML18100A966).

[18] NRC letter to TVA, "Staff Review of Seismic Probabilistic Risk Assessment Associated With Reevaluated Seismic Hazard Implementation of the Near-Term Task Force Recommendation 2.1: Seismic," dated July 10, 2018 (ADAMS Accession No. ML18115A138).

[19] NRC letter to TVA, "Issuance of Amendment Nos. 134 And 38 Regarding Adoption of Title 10 of the Code of Federal Regulations Section 50.69, 'Risk-Informed Categorization and Treatment Of Structures, Systems, and Components For Nuclear Power Plants,'"

dated April 30, 2020 (ADAMS Accession No. ML20076A194).

[20] EPRI Technical Report NP-6041-SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," Revision 1, dated August 1991.

[21] Exelon 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,'" dated June 6, 2018 (ADAMS Accession No. ML18157A260).

Enclosure LR-N25-0005 LAR S24-03 30

[22] Southern Nuclear letter to NRC, "License Amendment Request to Incorporate Seismic Probabilistic Risk Assessment into 10 CFR 50.69 Categorization Process Response to Request for Additional Information (RAIs 4-11)," dated February 21, 2018 (ADAMS Accession No. ML18052B342).

[23] TVA letter to NRC, "Application to Adopt 10 CFR 50.69, 'Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors,'" dated November 29, 2018 (ADAMS Accession No. ML18334A363).

[24] NRC Internal Memorandum, "Support Document for Screening and Prioritization Results Regarding Seismic Hazard Re-Evaluations for Operating Reactors in the Central and Eastern United States," dated May 21, 2014 (ADAMS Accession No. ML14136A126).

[25] NRC letter to PSEG, "Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations, 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," dated February 18, 2016 (ADAMS Accession No. ML16041A033).

[26] PSEG letter to NRC, "Seismic Hazard and Screening Report (CEUS Sites) 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," dated March 28, 2014 (ADAMS Accession No. ML14090A043).

[27] PSEG letter to NRC, "Salem Generating Station Response to Recommendation 2.3:

Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," dated November 26, 2012 (ADAMS Accession No. ML12339A127).

[28] NRC letter to PSEG, "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," dated May 14, 2014 (ADAMS Accession No. ML14113A236).

[29] PSEG letter to NRC, "NEI 12-06, Appendix H, Revision 2, H.4.3 Path 3: GMRS > SSE but

< IHS, Mitigating Strategies Assessment (MSA) report for the New Seismic Hazard Information," dated December 30, 2016 (ADAMS Accession No. ML16365A152).

[30] NRC letter to PSEG, "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," dated April 18, 2017 (ADAMS Accession No. ML17101A604).

[31] NRC letter to Specified Power Reactor Licensees, "Staff Review of High Frequency Confirmation Associated with Reevaluated Seismic Hazard in Response to March 12, 2012, 50.54(f) Request for Information," dated February 18, 2016 (ADAMS Accession No. ML15364A544).

Enclosure LR-N25-0005 LAR S24-03 31

[32] PSEG letter to NRC, "Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident,"

dated December 6, 2016 (ADAMS Accession No. ML16342C496).

[33] NRC letter to PSEG, "Staff Review of Spent Fuel Pool Evaluation Associated With Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1 and Staff Closure of Activities Associated with Recommendation 2.1," dated January 19, 2017 (ADAMS Accession No. ML16351A231).

[34] Exelon 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," dated October 1, 2020 (ADAMS Accession No. ML20275A292).

[35] Exelon letter to NRC, "Response to Request for Additional Information Regarding the Application to Adopt 10 CFR 50.69, 'Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors,'" dated July 1, 2019 (ADAMS Accession No. ML19183A012).

[36] Exelon letter to NRC, "Response to Request for Additional Information Regarding the Application to Adopt 10 CFR 50.69, 'Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors,'" dated July 19, 2019 (ADAMS Accession No. ML19200A216).

[37] Exelon letter to NRC, "Revised Response to Request for Additional Information Regarding the Application to Adopt 10 CFR 50.69, 'Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors" letter dated July 19, 2019," dated August 5, 2019 (ADAMS Accession No. ML19217A143).

[38] Generic Letter 88-20, Supplement 4, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f)," dated June 28, 1991 (ADAMS Accession No. ML031150485).

[39] EPRI Technical Report -1016737, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments," dated December 2008.

[40] NRC Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, dated March 2009 (ADAMS Accession No. ML090410014).

[41] NEI letter to NRC, "Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os)," dated February 21, 2017 (ADAMS Accession No. ML17086A431).

Enclosure LR-N25-0005 LAR S24-03 32

[42] 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)," dated May 3, 2017 (ADAMS Accession No. ML17079A427).

[43] Westinghouse Report, "RG 1.200 PRA Peer Review Against the ASME PRA Standard Requirements for the Salem Generating Station, Units 1 & 2 Probabilistic Risk Assessment," LTR-RAM-II-09-001, Westinghouse, June 2009.

[44] PSEG Calculation SA-MISC-024, "2018 Salem FPIE F&O Closure Review," Revision 0, dated October 26, 2018.

[45] ASME/ANS RA-Sb-2005, "Addenda to ASME RA-S-2002 Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," dated December 2005.

[46] NRC Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 1, dated January 2007 (ADAMS Accession No. ML070240001).

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

[48] PSEG Calculation SA-MISC-005, "Resolution of Internal Flood Peer Review Comments,"

Revision 0, dated October 28, 2011.

[49] PSEG Calculation SA-MISC-025, "PRA Finding-Level Fact and Observation Independent Assessment & Focused Scope Peer Review," Revision 1, dated March 26, 2024.

[50] Westinghouse Report PWROG-22025-P, "Peer Review of the Salem Units 1 and 2 Fire Probabilistic Risk Assessment," Revision 0, dated January 2023.

[51] NRC Regulatory Guide 1.200, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 3, dated December 2020 (ADAMS Accession No. ML20238B871).

[52] NEI Topical Report NEI 17-07, "Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard," Revision 2, dated August 2019 (ADAMS Accession No. ML19228A242).

[53] Westinghouse Report PWROG-23035-P, "Independent Assessment of Facts &

Observations Closure and Focused Scope Peer Review of the Salem Units 1 and 2 Fire Probabilistic Risk Assessment," Revision 0-A, dated December 2023.

[54] PSEG Report, "Individual Plant Examination for External Events," dated January 1996.

[55] PSEG Early Site Permit Site Safety Analysis Report, Chapter 3, "Design of Structures, Components, Equipment, and Systems," Revision 4, dated June 5, 2015 (ADAMS Accession No. ML15169A283).

Enclosure LR-N25-0005 LAR S24-03 33

[56] Salem Updated Final Safety Analysis Report, Revision 34, dated May 9, 2024 (ADAMS Accession No. ML24130A164).

[57] NRC Generic Letter 89-13, Supplement 1, "Service Water System Problems Affecting Safety-Related Equipment," dated April 4, 1990 (ADAMS Accession No. ML031140185).

[58] PSEG Procedure OP-AA-108-111-1001, "Severe Weather and Natural Disaster Guidelines".

[59] PSEG Procedure ER-AA-310-101, "Condition Monitoring of Structures".

[60] Salem Nuclear Generating Station, Units 1 and 2, "External Hazards Assessment for the Salem Nuclear Generating Station," Revision 0, dated October 2024.

[61] ASCE Standard 7-22, "Minimum Design Loads and Associated Criteria for Buildings and Other Structures," 2022.

[62] ASCE Hazard Tool, "https://ascehazardtool.org".

[63] NRC NUREG/CR-4461, "Tornado Climatology of the Contiguous United States," Revision 2, dated February 2007 (ADAMS Accession No. ML070810400).

[64] PSEG Procedure SC.OP-PT.ZZ-0002(Q), "Station Preparations For Seasonal Conditions".

[65] PSEG letter to NRC, "Response to Request for Information Regarding Flooding Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident - Salem Generating Station Flood Hazard Reevaluation," dated March 11, 2014 (ADAMS Accession No. ML14071A401).

[66] PSEG Procedure SC.OP-AB.ZZ-0001(Q), "Adverse Environmental Conditions".

[67] PSEG letter to NRC, "Focused Evaluation of External Flooding for Salem Generating Station, Units 1 and 2," dated June 30, 2017 (ADAMS Accession No. ML17181A221).

[68] NRC Regulatory Guide 1.78, "Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release,"

Revision 0, dated June 1974 (ADAMS Accession No. ML003740298).

[69] NRC letter to PSEG, "Salem Nuclear Generating Station, Units 1 and 2, - Staff Assessment of Response to 10 CFR 50.54(f) Information Request - Flood-Causing Mechanism Reevaluation," dated October 7, 2016 (ADAMS Accession No. ML16265A085).

[70] Siemens Westinghouse letter to NRC, "Missile Analysis Methodology for General Electric (GE) Nuclear Steam rotors by Siemens Westinghouse Power Corporation (SWPC)," dated May 16, 2002 (ADAMS Accession No. ML021430030).

Enclosure LR-N25-0005 LAR S24-03 34

[71] Siemens Technical Report CT-27336, "Missile Probability Analysis PSEG Nuclear LLC Salem Unit 1," Revision 1, dated November 5, 2003.

[72] NRC Regulatory Guide 1.115, "Protection Against Low-Trajectory Turbine Missiles,"

Revision 1, dated July 1977 (ADAMS Accession No. ML003739456).

[73] Salem Unit 1 and 2, "Assessment of Turbine Valve Testing Interval Extension," Final v2.0.

dated October 20, 2022.

[74] 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,"

dated December 2012.

[75] NRC NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," Revision 1, dated March 2017 (ADAMS Accession No. ML17062A466).

[76] PSEG Calculation SA-PRA-018, "Uncertainty Notebook," Revision 2, dated October 2022.

[77] NRC 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, dated January 2018 (ADAMS Accession No. ML17317A256).

[78] Exelon letter to NRC, "Response to Request for Additional Information Regarding License Amendment Requests to Adopt TSTF-505, Revision 2, and 10 CFR 50.69," dated November 24, 2020 (ADAMS Accession No. ML20329A433).

[79] EPRI Technical Report 3002012988, "Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization," dated July 2018.

[80] NRC letter to Exelon, "Issuance of Amendment Nos. 332 and 310 Re: Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors," dated February 28, 2020 (ADAMS Accession No. ML19330D909).

[81] NRC letter to PSEG, "Issuance of Amendment No. 224 Regarding Adoption of 10 CFR 50.69, 'Risk-Informed Categorization and Treatment of Structures, Systems and Components of Nuclear Power Reactors,'" dated September 29, 2020 (ADAMS Accession No. ML20231A632).

LR-N25-0005 LAR S24-03 1

PSEG has established procedure(s) prior to the use of the categorization process on a plant system. The procedure(s) contain the elements/steps listed below.

IDP member qualification requirements.

Qualitative assessment of system functions. System functions are qualitatively categorized as preliminary HSS or LSS based on the seven criteria in Section 9 of NEI 00-04 (see Section 3.2 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 PRA and non-PRA methods, covering all hazards. Safety significance of passive components is assessed using a methodology for passive components.

Assessment of 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 lDP for review and approval. The lDP 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 CDF and LERF and meets the acceptance guidelines of RG 1.174.

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. : List of Categorization Prerequisites LR-N25-0005 LAR S24-03 1

Units Model Baseline CDF Baseline LERF Comments FPIE & Internal Flooding PRA Model 1 and 2 SA121A-ASM-001 (Unit 1)

SA121A-ASM-001 (Unit 2) 3.0E-6/year (Unit 1) 3.0E-6/year (Unit 2) 1.3E-7/year (Unit 1) 1.3E-7/year (Unit 2) 2024 FPIE Application Specific Model (ASM)

Fire Model 1 and 2 SA121C-F-ASM-001 (Unit 1)

SA121C-F-ASM-001 (Unit 2) 8.3E-5/year (Unit 1) 7.7E-5/year (Unit 2) 6.7E-6/year (Unit 1) 5.2E-6/year (Unit 2) 2024 FPRA ASM

Description of PRA Models Used in Categorization LR-N25-0005 LAR S24-03 1

There are no Partially Resolved or Open Peer Review Findings or Self-Assessment Open Items for the Salem Internal Events and Internal Flooding and Fire PRA models.

Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items LR-N25-0005 LAR S24-03 1

Hazard Definition Screening Criteria Salem Disposition for 10 CFR 50.69 Aircraft Impact An aircraft (either a portion of (e.g., missile) or the entire aircraft) that collides either directly or indirectly (i.e.,

skidding impact with one or more SSCs at or in the plants analyzed area causing functional failure.

Secondary hazards resulting from an aircraft impact include, but are not necessarily limited to, fire.

PS4 Aircraft impacts are discussed in Section 5.6.1 of the IPEEE (Reference [54]). The IPEEE discussed an evaluation performed in 1994 that assessed and compared the then current risk to the prior 10-years of aircraft activity. The 1994 evaluation concluded that there was no significant change in risk since the prior 10-year period and that the risk remained negligible.

An updated evaluation of aircraft hazards is discussed in Section 3.5.1.6 of the Early Site Permit Site Safety Analysis Report (SSAR) (Reference [55]). The SSAR evaluates the suitability of an adjacent site for future construction and operation of a nuclear power plant and is appropriate to assess the hazard impact to Salem.

There are no airports within five miles of the PSEG Site.

Table 3.5-1 of (Reference [55]) lists six airports about ten to thirty miles from the plant along with the projected number of annual operations for the year 2025, where available. The hazard probability for these airports is considered acceptable if the projected annual number of operations is less than 1000D2. The screening limits are listed in Table 3.5-1 of (Reference [55]). None of these airports require additional hazard probability evaluations, as the projected number of operations for each airport does not exceed the respective screening limit.

Based on this review, the aircraft impact hazard is considered to be negligible.

4 The list of hazards and their potential impacts considered those items listed in Tables D-1 and D-2 in Appendix D of RG 1.200, Rev. 3 (Reference [51]. : Other External Hazards Disposition4 LR-N25-0005 LAR S24-03 2

Hazard Definition Screening Criteria Salem Disposition for 10 CFR 50.69 Avalanche Rapid flow of a large mass of accumulated frozen precipitation and other debris down a sloped surface resulting in dynamic loading of SSCs at or in the plants analyzed area causing functional failure or adverse impact on natural water supplies used for heat rejection.

C3 Per the IPEEE (Reference [54]). the Salem site is on an artificial island in the Delaware River far from this hazard.

Based on this review, the Avalanche hazard can be considered to be negligible.

Biological Events Accumulation or deposition of vegetation or organisms (e.g.,

zebra mussels, clams, fish, algae) on an intake structure or internal to a system that uses raw cooling water from a source of surface water, causing its functional failure.

C1 Detritus impacts are discussed in Section 5.8 of the IPEEE (Reference [54]). Salem protects the Circulating Water intake against detritus effects through:

Installing blowdown fittings on screen wash headers.

Purchasing new screen wash pumps capable of digesting detritus. Replacement of stilling tubes and base plates.

Upgrading screen wash pump motors and cables.

Upgrading of corroded portions of screen wash piping.

Refurbishing screen wash control panels to allow automatic screen wash operation.

Also, per UFSAR Section A.2.1.11 (Reference [56]) - Appendix B, License Renewal Final Safety Analysis Report Supplement - the Salem Open-Cycle Cooling Water System aging management program primarily consists of station GL 89-13 (Reference [57])

activities that include sodium hypochlorite injection, system testing, periodic inspections and non-destructive examination. The program includes surveillance and control techniques to manage aging effects caused by bio-fouling, corrosion, erosion, protective coating failures, and silting in the Service Water (SW) System components and on the SSCs supported by the SW System. Other activities LR-N25-0005 LAR S24-03 3

Hazard Definition Screening Criteria Salem Disposition for 10 CFR 50.69 include station maintenance inspections, component preventive maintenance, surveillance testing, and inspections. These activities provide for management of loss of material (without credit for protective coatings) and heat transfer reduction (including fouling from biological, corrosion product, and external sources) aging effects where applicable in system components exposed to a raw water environment.

Based on this review, the Biological Events hazard can be considered to be negligible.

Coastal Erosion Removal of material from a shoreline of a body of water (e.g., river, lake, ocean) due to surface processes (e.g., wave action, tidal currents, wave currents, drainage, or winds and including river bed scouring) that results in damage to the foundation of SSCs at or in the plants analyzed area, causing functional failure.

C1 C5 Coastal erosion is a slowly developing event and could be mitigated or adequately responded to (C5).

Per UFSAR Section 3.4.3 (Reference [56]), the existing earthen dike was replaced by a protective rockfill dike along the portion of the Delaware estuary subjected to maximum wind wave forces, to protect the safety related structures and equipment. (C1)

The shoreline protection and dike system are inspected by station operating personnel prior to storms and hurricanes and following the passage of such storms and hurricanes (Reference [58].

Additionally, periodic inspections of the shoreline dike are conducted per the "condition monitoring of structures" procedure (Reference [59]) The station security forces also make regular patrols of these areas as part of their surveillance duties and are instructed to report any abnormalities observed in the structure.

(C1)

Based on this review, the Coastal Erosion hazard can be considered to be negligible.

Drought A shortage of surface water supplies due to a period of below-average precipitation in a given region, thereby depleting the water supply needed for the C1 Per the IPEEE (Reference [54]), drought-induced reduction of the heat sink is not applicable to Salem.

Based on this review, the Drought hazard can be considered to be negligible.

LR-N25-0005 LAR S24-03 4

Hazard Definition Screening Criteria Salem Disposition for 10 CFR 50.69 various water-cooling functions at the facility.

External Flood An excess of water outside the plant boundary that causes functional failure to plant SSCs.

External flood causes include, but may not be limited to, flooding due to dam failure, high tide, hurricane (tropical cyclone), ice cover, local intense precipitation (LIP), river diversion, river and stream overflow, seiche, storm surge, and tsunami.

C1 C3 C4 C5 Per the evaluation documented in Section 2.2 of the Salem External Hazards Assessment (Reference [60]), there is only one flood causing mechanism that was not bounded by the Salem current design basis (CDB): LIP. The remaining flood causing mechanisms were all bound by the design of the plant and did not receive further analysis in a focused evaluation.

For LIP, PSEG credits a list of doors shown in Attachment 1 of operating procedure SC.OP-AB.ZZ-0001 to screen this flood mechanism. Accordingly, these doors shall not be categorized lower than HSS, as they are credited for screening this hazard in accordance with Figure 5-6 of NEI 00-04.

Extreme Winds and Tornadoes Strong winds resulting in dynamic loading or missile impacts on SCCs causing functional failure.

Hazards that could potentially result in high wind include the following:

• hurricane - severe winds developed from a tropical depression resulting in missiles or dynamic loading on SSCs. Secondary hazards resulting from a hurricane, include, but are not necessarily limited to tornado

• straight wind - a strong wind resulting in missiles or dynamic loading on SSCs PS4 Tornado wind speed hazard curve information for Salem is based on the most recent tornado data from American Society of Civil Engineers (ASCE) Standard ASCE 7-22 and the ASCE Hazard Tool (References [61] and [62] respectively). Based on the Salem site specific tornado hazard data from the ASCE Hazard Tool, the tornado wind speed for the 1,000,000 year Mean Recurrence Interval for the Salem site is 195 mph. The comparable wind speed for the 1E-6 annual exceedance probability data in Table 6-1 of NUREG/CR-4461, Revision 2 (Reference [63]) shows wind speed for Class I structures is much less than 1E-6/year. Thus, wind pressure effects can be screened using Criterion PS4.

For tornado missiles, an updated conservative tornado/tornado missile risk model was developed for both units showing tornado missile risk to be less than 1E-6/year and LERF much less than 1E-7/year. Therefore, the tornado missile hazard can also be screened using Criterion PS4.

LR-N25-0005 LAR S24-03 5

Hazard Definition Screening Criteria Salem Disposition for 10 CFR 50.69 that is not associated with either hurricanes or tornadoes

• tornado - a strong whirlwind that results in missiles or dynamic loading on SSCs Fog Low-lying water vapor in the form of a cloud or obscuring haze of atmospheric dust or smoke resulting in impeded visibility that could result in, for example, a transportation accident.

Fog Low-lying water vapor in the form of a cloud or obscuring haze of atmospheric dust or smoke resulting in impeded visibility that could result in, for example, a transportation accident.

C4 The principal effects of such events (such as freezing fog) would be to cause a LOOP, which is addressed in weather-related LOOP scenarios in the FPIE PRA model for Salem.

Based on this review, the Fog hazard can be considered to be negligible.

Forest Fire Direct (e.g., thermal effects) and indirect effects (e.g., generation of combustion products, transport of firebrand) of a forest fire outside the plant boundary that causes functional failure of plant SSCs.

Hazards that could cause or be caused by a forest fire include, but may not be limited to, wildfires and grass fires.

C1 C4 Per the IPEEE (Reference [54]), offsite fires have little or no effect on site because of site clearing during construction. (C1)

In addition, forest fires originating from outside the plant boundary may cause a LOOP, which is addressed for grid related LOOP scenarios in the FPIE PRA model (C4).

Based on this review, the Forest Fire hazard can be considered to be negligible.

LR-N25-0005 LAR S24-03 6

Hazard Definition Screening Criteria Salem Disposition for 10 CFR 50.69 Frost A thin layer of ice crystals that form on the ground or the surface of an earthbound object when the temperature of the ground or surface of the object falls below freezing.

C4 The principal effects of such events would be to cause a LOOP, which is addressed for weather-related LOOP scenarios in the FPIE PRA model for Salem.

Based on this review, the Frost hazard can be considered to be negligible.

Hail A shower of ice or hard snow that could result in transportation accidents or directly causes dynamic loading or freezing conditions as a result of ice coverage.

C4 The principal effects of such events would be to cause a LOOP, which is addressed for weather-related LOOP scenarios in the FPIE PRA model for Salem.

Based on this review, the Hail hazard can be considered to be negligible.

High Summer Temperature Effects on SSC operation due to abnormally high ambient temperatures resulting from weather phenomena.

Secondary hazards resulting from high ambient temperatures, include, but are not necessarily limited, to low lake or river water levels.

C1 C4 C5 Per UFSAR Section 3.1.2 (Reference [56]), the systems and components designated Class I are designed to withstand the most severe environmental hazards discussed and analyzed in Sections 2 (Site Characteristics) and 3 (Design of Structures, Components, Equipment and Systems. (C1)

High summer temperature effects would take place slowly, allowing time for orderly plant reductions, including shutdowns. PSEG procedure SC.OP-PT.ZZ-0002(Q), "Station Preparations For Seasonal Conditions" (Reference [64]), includes instructions for summer inspections and extreme heat conditions to mitigate the effects of high summer temperature. (C5)

Also, plant trips due to this hazard are covered in the definition of another event in the PRA model (e.g., transients, loss of condenser) (C4).

Based on this review, the High Summer Temperature hazard can be considered to be negligible.

High Tide The periodic maximum rise of sea level resulting from the combined effects of the tidal gravitational forces exerted by C1 Per the Salem Flood Hazard Reevaluation Report (FHRR)

(Reference [65]), the high tide hazard is bounded by the design of the plant and was not required to be included in the flood Focused Evaluation for additional analysis.

LR-N25-0005 LAR S24-03 7

Hazard Definition Screening Criteria Salem Disposition for 10 CFR 50.69 the Moon and Sun and the rotation of the Earth.

Based on this review, the High Tide hazard can be considered to be negligible.

Hurricane (Tropical Cyclone)

Flooding that results from the intense rain fall from a hurricane (tropical cyclone).

Secondary hazards resulting from a hurricane include, but are not necessarily limited to, dam failure, high tide, river and stream overflow, seiche, storm surge, and waves.

C1 C5 PS4 The screening assessment for the extreme high winds and external flood hazards discussed in this Attachment bound the risk due to hurricanes PSEG procedure SC.OP-AB.ZZ-0001 directs operators to close all watertight doors listed in Attachment 1 of SC.OP-AB.ZZ-0001 (Reference [66]) prior to the arrival of hurricane-driven flood waters. Clear procedural guidance is provided to accomplish these actions well in advance of the storm affecting the site (C5).

These flood protection features shall not be categorized lower than HSS, as they are credited with screening this hazard per NEI 00-04, Figure 5-6.

With the doors in their closed position, the Hurricane (Tropical Cyclone) hazard can be screened from further consideration in the 10 CFR 50.69 Program based on plant design (C1).

See also Extreme High Winds / Tornadoes and External Flooding and Intense Precipitation.

Ice Cover Flooding due to downstream blockages of ice on a river.

Secondary hazards resulting from an ice blockage include, but are not necessarily limited to, river and stream overflow.

C1 Per the IPEEE (Reference [54]), ice blockage-induced reduction in heat sink is not applicable to Salem.

In addition, per UFSAR Section 3.1.2 (Reference [56]), the systems and components designated Class I are designed to withstand the most severe environmental hazards discussed and analyzed in Sections 2 (Site Characteristics) and 3 (Design of Structures, Components, Equipment and Systems.

Based on this review, the Ice Cover hazard can be considered to be negligible.

LR-N25-0005 LAR S24-03 8

Hazard Definition Screening Criteria Salem Disposition for 10 CFR 50.69 Industrial or Military Facility Accident An accident at an offsite industrial or military facility that results in a release of toxic gases, a release of combustion products, a release of radioactivity, an explosion, or the generation of missiles.

C1 Per UFSAR Section 2.2 (Reference [56]), the Salem site is located in a rural area consisting of marshes, abandoned meadowland, and some farmland. There are no major manufacturing or chemical plants within 5 miles of the site. All such facilities are beyond 8 miles and would not interfere with the normal operation of Salem.

In addition, per UFSAR Section 2.2.2.1, there are no military bases or missile sites within 10 miles of the site.

Based on this review, the industrial or military facility accident hazard can be considered negligible.

Internal Flood N/A The Salem FPIE PRA includes evaluation of risk from internal flooding events.

Internal Fire N/A The Salem FPRA model addresses risk from internal fires Landslide Dynamic loading of SSCs or impacts on natural water supplies used for heat rejection due to the movement of rock, soil, and mud down a sloped surface (does not include frozen precipitation).

C3 Per the IPEEE (Reference [54]), the Salem site is on an artificial island in the Delaware River far from this hazard.

Based on this review, the Landslide hazard can be considered to be negligible.

Lightning Effects on SSCs due to a sudden electrical discharge from a cloud to the ground or Earth-bound object.

C1 C4 Per UFSAR Section 3.8.1.1 (Reference [56]), a lightning protection system is installed on the containment dome to protect against electrical storm damage. (C1).

Lightning strikes that result in reactor trips are addressed in the Salem PRA models. (C4).

Based on this review, the Lightning hazard can be considered to be negligible.

Low Lake or River Water Level A decrease in the water level of the lake or river used for power generation.

C1 This hazard is of negligible impact on the plant due to the large volume and tidal characteristics of the Delaware River. The plant location in the Delaware Estuary system precludes impact on the plant due to this hazard.

LR-N25-0005 LAR S24-03 9

Hazard Definition Screening Criteria Salem Disposition for 10 CFR 50.69 Per UFSAR Section 2.4.11.6 (Reference [56]), the Water Intake System is designed to operate at the lowest postulated water level in the Delaware River Estuary (Elevation -13.1 feet MSL).

Based on this review, the Low Lake or River Water Level hazard can be considered to be negligible.

Low Winter Temperature Effects on SSC operation due to abnormally low ambient temperatures resulting from weather phenomena.

Secondary hazards resulting from low ambient temperatures include, but are not necessarily limited to, frost, ice cover, and snow.

C1 C5 Per UFSAR Section 3.1.2 (Reference [56]), the systems and components designated Class I are designed to withstand the most severe environmental hazards discussed and analyzed in Sections 2 (Site Characteristics) and 3 (Design of Structures, Components, Equipment and Systems. (C1)

PSEG procedure SC.OP-PT.ZZ-0002(Q), "Station Preparations For Seasonal Conditions" (Reference [64]), includes station instructions for winter inspections and extreme cold conditions. (C5)

Based on this review, the Low Winter Temperature hazard can be considered to be negligible.

Meteorite/Satellite Strikes A release of energy due to the impact of a space object such as a meteoroid, comet, or human-caused satellite falling within the Earths atmosphere, a direct impact with the Earths surface, or a combination of these effects.

This hazard is analyzed with respect to direct impacts of an SSC and indirect impact effects such as thermal effects (e.g.,

radiative heat transfer),

overpressure effects, seismic PS4 The likelihood of a large meteorite or satellite, large enough to cause significant plant damage, is judged to be very low.

Based on this review, the Meteorite or Satellite hazard can be considered to be negligible.

LR-N25-0005 LAR S24-03 10 Hazard Definition Screening Criteria Salem Disposition for 10 CFR 50.69 effects, and the effects of ejecta resulting from a ground strike.

Pipeline Accident A release of hazardous material, a release of combustion products, an explosion, or the generation of missiles due to an accident involving the rupture of a pipeline carrying hazardous materials.

C1 Per UFSAR Section 2.2 (Reference [56]), there are no major manufacturing or chemical plants within 5 miles of the site. All such facilities are beyond 8 miles and would not interfere with the normal operation of Salem. There are no pipelines within 5 miles of the site.

Based on this review, the Pipeline Accident hazard can be considered to be negligible.

Precipitation, Intense Flooding that results from LIP.

Secondary hazards resulting from LIP, include, but are not necessarily limited to, dam failure and river and stream overflow.

C1 The flood Focused Evaluation (Reference [67]) affirmed that Salem has effective flood protection from the LIP mechanism and will not require additional safety enhancements since watertight doors are provided around the power block and for buildings that house SSCs important to safety.

The watertight doors listed in Attachment 1 of PSEG Procedure SC.OP-AB.ZZ-0001 (Reference [66]) shall not be categorized lower than HSS as they are credited with screening this hazard per NEI 00-04, Figure 5-6.

Based on this review, the risk from the Intense Precipitation hazard can be considered negligible.

Release of Chemicals from Onsite Storage A release of hazardous material including, but not limited to liquids, combustion products, or radioactivity.

Such releases may be concurrent with or induce an explosion or the generation of missiles.

In this context, an onsite release of radioactivity is assumed to be C1 Per Section 5.7 of the IPEEE (Reference [54]), accidents involving release of on-site chemical storage do not pose a vulnerability at Salem owing to conformance to RG 1.78 (Reference [68]).

Per UFSAR Section 2.2.3.3 (Reference [56]), a release of ammonium hydroxide directly from a delivery tanker while onsite may exceed the toxicity limit contained in RG 1.78; however, administrative controls are in place to prevent the control rooms from the limit. Also, the station control rooms have separate and ventilation air which are automatically isolable. Calculations indicated that the toxicity limit found in RG 1.78 will not be LR-N25-0005 LAR S24-03 11 Hazard Definition Screening Criteria Salem Disposition for 10 CFR 50.69 associated with low-level radioactive waste.

exceeded in the control rooms during a postulated release at any of the sources.

See also Toxic Gas.

Based on this review, the Release of Chemicals from Onsite Storage hazard can be considered to be negligible.

River Diversion The redirection of all or a portion of river flow by natural causes (e.g., a riverine embankment landslide) or intentionally (e.g.,

power production, irrigation).

C1 UFSAR Sections 2.4.8 and 2.4.9 (Reference [56]) state that the Delaware Estuary is the cooling water reservoir for the plant. As the source of cooling water is the Delaware Estuary, no channel diversions need be considered.

Based on this review, the River Diversion hazard can be considered to be negligible.

Sandstorm Persistent heavy winds transporting sand or dust that infiltrate SSCs at or in the plants analyzed area causing functional failure.

C3 Salem is not located near sand dunes or other large sources of small airborne particles. More common wind-borne dirt can occur but poses no significant risk given the robust structures and protective features of the plant.

Based on this review, the Sandstorm hazard can be considered to be negligible.

Seiche Flooding from water displaced by an oscillation of the surface of a landlocked body of water, such as a lake, that can vary in period from minutes to several hours.

C3 Salem is an inland location and does not connect directly with any bodies of water capable of producing seiche.

Based on this review, the Seiche hazard can be considered to be negligible.

Seismic Activity Sudden ground motion or vibration of the Earth as produced by a rapid release of stored-up energy along an active fault.

Secondary hazards resulting from seismic activity include, but N/A The Salem 10 CFR 50.69 Program will use the NRC-accepted EPRI Tier 2 alternative seismic approach for categorization relative to the seismic hazard.

See Section 3.2.3 of the Enclosure to this application.

LR-N25-0005 LAR S24-03 12 Hazard Definition Screening Criteria Salem Disposition for 10 CFR 50.69 are not necessarily limited to, avalanche (both rock and snow), dam failure, industrial accidents, landslide, seiche, tsunami, and vehicle accidents.

Snow The accumulation of snow could result in transportation accidents or directly cause dynamic loading or freezing conditions as a result of snow cover.

C4 C5 This hazard is slow to develop and can be identified via monitoring and managed via normal plant processes (C5). PSEG procedure OP-AA-108-111-1001, "Severe Weather and Natural Disaster Guidelines" (Reference [58]) includes planning guidance and instruction for snow removal during winter storms.

Potential flooding impacts are accounted for under external flooding screening (C4).

Based on this review, the Snow hazard can be considered to be negligible.

Soil Shrink-Swell Dynamic forces on structures foundations due to the expansion (swelling) and contraction (shrinking) of soil resulting from changes in the soil moisture content.

C1 Per the IPEEE (Reference [54]), soil subsidence is part of the design basis. The IPEEE refers to the UFSAR stating that soil subsidence is not an issue.

Per the UFSAR Section 2.5.5 (Reference [56]) (slope stability), at the completion of construction, the only slope of significance across the site is at the sea wall. As discussed in Section 2.4.5.7, Protective Structures, the sea wall was investigated by conventional engineering procedures and designed to withstand the site maximum environmental loadings.

Based on this review, the Soil Shrink-Swell Consolidation impact hazard can be considered to be negligible.

Storm Surge Flooding that results from an abnormal rise in sea level due to atmospheric pressure changes and strong wind generally accompanied by an intense storm.

C1 C5 Per the Salem FHRR (Reference [65]), the storm surge flood mechanism is bounded by the CDB of the plant.

The risk significance from the storm surge hazard is screened utilizing flood protection features installed at the plant. These features are listed in Attachment 1 of PSEG procedure SC.OP-LR-N25-0005 LAR S24-03 13 Hazard Definition Screening Criteria Salem Disposition for 10 CFR 50.69 Secondary hazards resulting from a storm surge include, but are not necessarily limited to, high tide, river and stream overflow, and waves.

AB.ZZ-0001 (Reference [66]). These flood protection features shall not be categorized lower than HSS, as they are credited with screening this hazard per NEI 00-04, Figure 5-6.

Based on this review, the risk from the storm surge hazard can be considered negligible.

Toxic Gas A release of hazardous toxic or asphyxiant gases.

Such releases may be concurrent with or induce an explosion or the generation of missiles.

In this context, an onsite release of radioactivity is assumed to be associated with low-level radioactive waste.

C1 Per UFSAR Section 2.2 (Reference [56]), there are no major manufacturing or chemical plants within 5 miles of the site. All such facilities are beyond 8 miles and would not interfere with the normal operation of Salem. Also, Salem uses a sodium hypochlorite biocide system, thus eliminating an onsite chlorine hazard.

See also Release of Chemicals from Onsite Storage.

Based on this review, the Toxic Gas hazard can be considered to be negligible.

Transportation Accidents Accidents involving transportation resulting in collision with SSCs, a release of hazardous materials or combustion products, an explosion, or a generation of missiles causing functional failure of SSCs.

Hazards that could potentially result in transportation accidents include, for example, a vehicle, railcar or ship (boat) accident that involves a collision or derailment, potentially resulting in fire, explosions, toxic C1 Per UFSAR Section 2.2 (Reference [56]), the Delaware River, a major transportation route, represents the only possible hazard to Salem due to the Intracoastal Waterway which passes through the River 1.5 miles west of the site. There are no major harbors, railway yards, or airports within 10 miles of the site.

Per UFSAR Section 6.4.4.3, hazardous chemicals shipped past the Salem site occur infrequently. The frequencies of the deliveries are listed in Table 2.2-4. RG 1.78 requires a control room habitability evaluation for shipments of hazardous chemicals that are considered "frequent" shipments. The frequent criteria for river barges are 50 per year. None of the hazardous chemicals shipped past the site exceed these criteria, therefore, a control room habitability evaluation is not required.

Based on this review, the Transportation Accidents hazard can be considered to be negligible.

LR-N25-0005 LAR S24-03 14 Hazard Definition Screening Criteria Salem Disposition for 10 CFR 50.69 releases, missiles, or other hazardous conditions.

Tsunami Flooding that results from a series of long-period sea waves that displaces massive amounts of water as a result of an impulsive disturbance, such as a major submarine slides or landslide.

Secondary hazards resulting from a tsunami include, but are not necessarily limited to, river and stream overflow.

C1 Salem is an inland location and does not connect directly with any bodies of water capable of producing tsunamis.

Per the Salem FHRR(Reference [65], the Tsunami flood mechanism is bounded by the current design basis. The NRC accepted this in its review of the Salem FHRR (Reference [69]).

Based on this review, the Tsunami hazard can be considered to be negligible.

Turbine-Generated Missiles Damage to safety-related SSCs from a missile generated internal or external to the plant PRA boundary from rotating turbines or other external sources (e.g., high-pressure gas cylinders).

Damage may result from a falling missile or a missile ejected directly toward safety-related SSCs (i.e., low-trajectory missiles).

PS4 Per the IPEEE (Reference [54]), the turbine trip capability was upgraded following a 1991 overspeed event. Based on regular inspection of low pressure turbine discs and the overspeed protection system, the probability of turbine failure leading to missiles is considered acceptably small.

UFSAR Section 3.5.4 (Reference [56]) further discusses the turbine missile hazard, stating that Siemens Westinghouse Power Corporation (SWPC), prepared missile probabilities using a methodology approved by the NRC (Reference [70]). The results of SWPC missile analysis are documented in Reference [71]. The UFSAR refers to Reference [71] concluding that the missile probabilities are within the limits required by RG 1.115 (Reference [72]) of 1E-5 per year for an unfavorably oriented low pressure turbine.

The UFSAR states that the Unit 1 low pressure turbines will be inspected on a 100,000 equivalent operating hours frequency to ensure that the probability of turbine generated missile remains within the requirements of RG 1.115. The Unit 2 low pressure turbines are equipped with fully integral rotors which will be LR-N25-0005 LAR S24-03 15 Hazard Definition Screening Criteria Salem Disposition for 10 CFR 50.69 inspected at suitable intervals to ensure the probability of rotor burst is acceptably low. A 2022 assessment of turbine valve testing for an interval extension provides a recent confirmatory assessment (Reference [73]).

Based on this review, the Turbine Missile hazard can be considered to be negligible.

Volcanic Activity Opening of Earths crust resulting in tephra (i.e., rock fragments and particles ejected by volcanic eruption), lava flows, lahars (i.e., mud flows down volcano slopes), volcanic gases, pyroclastic flows (i.e., fast-moving flow of hot gas and volcanic matter moving down and away from a volcano), and landslides.

Indirect impacts include distant ash fallout (e.g., tens to potentially thousands of miles away).

Secondary hazards resulting from volcanic activity, include, but are not necessarily limited to, seismic activity and fire.

C3 Per the IPEEE (Reference [54]), the Salem site is on an artificial island in the Delaware River far from this hazard.

Based on this review, the Volcanic Activity hazard can be considered to be negligible.

Waves An area of moving water that is raised above the main surface of a body of water as a result of the wind blowing over an area of fluid surface.

C1 Per the Salem FHRR (Reference [65]), wind-generated waves and run-up effects (included in the combined effects flood mechanism) are bounded by the CDB of the plant.

See also Storm Surge.

LR-N25-0005 LAR S24-03 16 Hazard Definition Screening Criteria Salem Disposition for 10 CFR 50.69 Based on this review, the risk from waves coincident with flooding can be considered negligible.

LR-N25-0005 LAR S24-03 17 Salem Response Spectra Figure A4-1: GMRS and SSE Response Spectra for Salem From Reference [26], Table 2-4 (GMRS) and Table 3-1(SSE)

LR-N25-0005 LAR S24-03 1

Event Analysis Criterion Source Comments Initial Preliminary Screening C1. Event damage potential is < events for which plant is designed.

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

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

NUREG/CR-2300 and ASME/ANS Standard RA-Sa-2009 C4. Event is included in the definition of another event.

NUREG/CR-2300 and ASME/ANS Standard RA-Sa-2009 Not used to screen. Used only to include within another event.

C5. Event develops slowly, allowing adequate time to eliminate or mitigate the threat.

ASME/ANS Standard RA-Sa-2009 Progressive Screening PS1. Design basis hazard cannot cause a core damage accident.

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

NUREG-1407 and ASME/ANS Standard RA-Sa-2009 PS3. Design basis event mean frequency is < 1E-5/y and the mean conditional core damage probability is <

0.1.

NUREG-1407 as modified in ASME/ANS Standard RA-Sa-2009 PS4. Bounding mean CDF is < 1E-6/y.

NUREG-1407 and ASME/ANS Standard RA-Sa-2009 Detailed PRA Screening not successful.

PRA needs to meet requirements in the ASME/ANS PRA Standard.

NUREG-1407 and ASME/ANS Standard RA-Sa-2009

Progressive Screening Approach for Addressing External Hazards LR-N25-0005 LAR S24-03 1

The Salem FPIE and FPRA models and documentation were reviewed for plant-specific modeling assumptions and related sources of uncertainty, and the applicable lists of EPRI-identified generic sources of uncertainty per EPRI 1016737 (Reference [39]) and EPRI 1026511 (Reference [74]) were also reviewed.

Each PRA model includes an evaluation of the potential sources of uncertainty for the base case models using the approach that is consistent with the ASME/ANS RA-Sa-2009 (Reference [47]) 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 [75]).

These evaluations are documented in the internal events and internal flooding uncertainty notebook (Reference [76]). 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.

Additionally, an evaluation of Level 2 Internal Events PRA model uncertainty was performed, based on the guidance in NUREG-1855 and EPRI report 1026511 (Reference [74]). The potential sources of model uncertainty in the Salem PRA models were evaluated for the 32 Level 2 PRA topics outlined in EPRI 1026511 (Reference [74]).

For the 10 CFR 50.69 Program, the guidance in NEI 00-04 (Reference [2]) 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 1.174, Revision 3 (Reference [77]) 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 10 CFR 50.69 application in Attachment 6 to determine if additional sensitivity evaluations are needed.

Note: As part of the required 10 CFR 50.69 PRA categorization sensitivity cases directed by NEI 00-04, internal events / internal flood and FPRA models human error and common cause basic events are increased to their 95th percentile and also decreased to their 5th percentile values. 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 10 CFR 50.69 application.

Disposition of Key Assumptions/Sources of Uncertainty LR-N25-0005 LAR S24-03 2

The table below describes the internal events / internal flooding (IE / IF) PRA sources of model uncertainty and their impact.

IE / IF PRA Sources of Assumption/

Uncertainty IE / IF PRA 10 CFR 50.69 Impact IE / IF PRA Model Sensitivity and Disposition SW HVAC: Loss of SW HVAC (not explicitly modeled) would be detectable and could be compensated for prior to significant consequences.

Lacking a detailed analysis this could be a source of uncertainty for model applications.

Components or functions related to loss of SW Modeling is judged to be realistic.

Sensitivity studies indicated that the base model results are not significantly impacted by the current modeling assumptions.

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

Battery Life: Credit for battery life out to four hours is identified as a candidate source of model uncertainty.

Components or functions related to loss of offsite and onsite AC sources Modeling is judged to be realistic.

Sensitivity studies indicated that the base model results are not significantly impacted by the current modeling assumptions.

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

Sump blockage: The incorporation of a single value for unrecoverable failure due to sump strainer plugging for all sequences is a potential source of uncertainty.

Components or functions for which emergency core cooling systems are credited Modeling is judged to be reasonably conservative. Sensitivity studies indicated that the base model results are not significantly impacted by the current modeling assumptions.

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

Induced Tube Rupture Probability: Salem Analysis uses PWROG-21024-P as the basis for induced rupture probabilities.

Although this has been peer reviewed for Salem, this method is not yet an accepted peer reviewed Newly Developed Method.

Components or functions for which LERF may be impacted.

Modeling is judged to be realistic.

Sensitivity studies indicated that the base model results are not significantly impacted by the current modeling assumptions.

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

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IE / IF PRA Sources of Assumption/

Uncertainty IE / IF PRA 10 CFR 50.69 Impact IE / IF PRA Model Sensitivity and Disposition Inter-System Loss of Coolant Accident (ISLOCA): The approach for the ISLOCA frequency determination may not take into account the latest research, data and practices that have been developed.

ISLOCA components or functions ISLOCA impacts are also required for containment DID assessments for 10 CFR 50.69 applications.

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

FLEX Human Reliability Analysis (HRA): HRA in general is recognized as a potential source of uncertainty. This is especially true for treatment and credit for FLEX mitigating strategies using portable equipment.

Components or functions related to loss of offsite or onsite AC sources.

The Salem model HRA is based on industry consensus modelling 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.

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The table below describes the FPRA sources of model uncertainty and their impact.

FPRA Description FPRA Sources of Uncertainty FPRA Disposition Analysis boundary and partitioning This task establishes the overall spatial scope of the analysis and provides a framework for organizing the data for the analysis. The partitioning features credited are required to satisfy established industry standards.

Based on a review of the assumptions and potential sources of uncertainly associated with this element, it is concluded that the methodology for the Analysis Boundary and Partitioning task does not introduce any epistemic uncertainties that would affect the 10 CFR 50.69 calculations.

Component Selection This task involves the selection of components to be treated in the analysis in the context of initiating events and mitigation. The potential sources of uncertainty include those inherent in the internal events PRA model as that model provides the foundation for the FPRA.

In the context of the FPRA, one of the uncertainty issues that is unique to the analysis is related to initiating event identification. However, that impact is minimized through use of the Generic Multiple Spurious Operation (MSO) list and the process used to identify and assess potential MSOs.

Based on the discussion of sources of uncertainty and the discussion above, it is concluded that the methodology for the Component Selection task does not introduce any epistemic uncertainties that would require sensitivity treatment.

Cable Selection The selection of cables to be considered in the analysis is identified using industry guidance documents.

The overall process is essentially the same as that used to perform the analyses to demonstrate compliance with 10 CFR 50.48.

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

Qualitative Screening 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.

Qualitative screening was performed on the Global Analysis Boundary. The only qualitative screening criterion subject to uncertainty is the potential for plant trip.

Based on a review of the assumptions and potential sources of uncertainty related to this element, 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.

Fire-Induced Risk Model The internal events PRA model was updated to add fire specific initiating event structure as well as additional system logic. The methodology used is consistent with that used for the internal The identified source of uncertainty could result in the over-estimation of fire risk. In general, the FPRA development process would have reviewed significant fire initiating LR-N25-0005 LAR S24-03 5

FPRA Description FPRA Sources of Uncertainty FPRA Disposition events PRA model development and was subjected to industry Peer Review.

The developed model is applied in such a fashion that all postulated fires are assumed to generate a plant trip. This represents a source of uncertainty, as it is not necessarily clear that fires would result in a trip. In the event the fire results in damage to cables and/or equipment identified in Task 2, the PRA model includes structure to translate them into the appropriate induced initiator.

events and performed supplemental assessments to address this possible source of uncertainty.

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 Fire-Induced Risk Model task does not introduce any epistemic uncertainties that would affect the 10 CFR 50.69 calculations.

Fire Ignition Frequency Ignition source counting is an area with inherent uncertainty. However, the results are not particularly sensitive to changes in ignition frequency. The primary source of uncertainty for this task is associated with the generic ignition frequency values provided in NUREG-2178 (for Bin 4), NUREG-2230 (for Bin 15), NUREG--2262 (for High Energy Arc Fault 16.* Bins), and NUREG-2169 (for all other Bins), which were Bayesian updated as part of the Fire Ignition Frequency Notebook, SA-PRA-110.

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.

Quantitative Screening Quantitative screening was not performed for the Salem FPRA. For further details, see the Quantitative Screening Notebook, SA-PRA-115.

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.

Scoping Fire Modeling This task was accomplished with the use of fire modeling treatments in lieu of a conservative scoping analysis technique. The primary conservatism introduced by this task is associated with the heat release rates specified in NUREG/CR-6850 for non-electrical enclosure ignition sources.

The employment of fire modeling solutions (as opposed to more generic scoping fire modeling) has the potential to reduce conservatisms.

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

Detailed Circuit Failure Analysis Uncertainty considerations for the circuit failure analysis task are Circuit analysis was performed as part of the deterministic post fire safe LR-N25-0005 LAR S24-03 6

FPRA Description FPRA Sources of Uncertainty FPRA Disposition addressed via the use of circuit failure mode probability factors in Task 10 (Circuit Failure Mode Likelihood Analysis (CFMLA)). No specific uncertainty is associated with the performance of the circuit analysis.

No specific uncertainty is associated with the performance of the circuit analysis, though the CFMLA Notebook, SA-PRA-113 identifies a number of assumptions in Section 2.2.

shutdown analysis. Refinements in the application of the circuit analysis results to the FPRA were performed on a case-by-case basis where the scenario risk quantification was large enough to warrant further detailed analysis. Hot short probabilities and hot short duration probabilities as defined in NUREG-7150, Volume 2, based on actual fire test data, were used in the Salem FPRA. The uncertainty (conservatism) which may remain in the FPRA is associated with scenarios that do not contribute significantly to the overall fire risk.

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.

Circuit Failure Mode Likelihood Analysis The uncertainty associated with the applied conditional failure probabilities poses competing considerations, which is primarily due to the assumption that all spurious operations occur at the same time. The hot short probabilities and the hot short duration factors defined in NUREG/CR-7150 are considered the best available data.

CFMLA was generally limited to those components where spurious operation was expected to be a large contributor to total risk.

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 CFMLA task does not introduce any epistemic uncertainties that would affect the 10 CFR 50.69 calculations.

Detailed Fire Modeling The primary uncertainty in this task is in the area of target failure probabilities.

Conservative heat release rates may result in additional target damage. Non-conservative heat release rates would have an opposite effect. Credit for fire brigade response and detection are limited to the Multi-Compartment Analysis and the Hot Gas Layer evaluation.

Consensus modeling approach is used for the Detailed Fire Modelling.

The methodology for the Detailed Fire Modelling task does not introduce any epistemic uncertainties that affect the 10 CFR 50.69 calculations LR-N25-0005 LAR S24-03 7

FPRA Description FPRA Sources of Uncertainty FPRA Disposition Fire modeling was used to evaluate the time to abandonment for control room fire scenarios for a range of fire heat release rates. For further details, see the FPRA Control Room Fire Modeling Analysis Notebook, SA-PRA-106.

Post-Fire Human Reliability Analysis The HEPs used in the FPRA were adjusted to consider the additional challenges that may be present given a fire. The HEPs were obtained using the EPRI HRAC and included the consideration of degradation or loss of necessary cues due to fire. Given the methodology used, the impact of any remaining uncertainties is expected to be small.

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 Salem 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.

Seismic-Fire Interactions Assessment Since this is a qualitative evaluation, there is no quantitative impact with respect to the uncertainty of this task.

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.

Fire Risk Quantification As the culmination of other tasks, most of the uncertainty associated with quantification has already been addressed. The other source of uncertainty is the selection of the truncation limit. However, the selected truncation was confirmed to be consistent with the requirements of the PRA Standard (Reference [47]).

The selected truncation was confirmed to be consistent with the requirements of the PRA Standard.

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 Fire Risk Quantification task does not introduce any epistemic uncertainties that would affect the 10 CFR 50.69 calculations.

Uncertainty and Sensitivity Analyses 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.

The Uncertainty and Sensitivity Analyses task does not introduce any epistemic uncertainties that would affect the 10 CFR 50.69 calculations.

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FPRA Description FPRA Sources of Uncertainty FPRA Disposition FPRA Documentation This task does not introduce any new uncertainties to the fire risk.

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

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