ML25219A763

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Seabrook Station, Unit No. 1 - Request for Additional Information LARs to Adopt TSTF-505 and Provisions of 10 CFR 50.36 (Email Dated 8-7-2025) (EPID L-2025-LLA-0025, L-2025-LLA-0058)
ML25219A763
Person / Time
Site: Seabrook 
Issue date: 08/07/2025
From: Richard Guzman
NRC/NRR/DORL/LPL1
To: Mack J
Florida Power & Light Energy Seabrook
References
EPID L-2025-LLA-0025, EPID L-2025-LLA-0058
Download: ML25219A763 (1)
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Text

{{#Wiki_filter:From: Richard Guzman To: jarrett.mack@fpl.com Cc: Hipo Gonzalez; Ricardo Lantigua; jerry.phillabaum@fpl.com

Subject:

Seabrook Station, Unit 1 - Request for Additional Information re: License Amendment Requests to Adopt TSTF-505 and Provisions of 10 CFR 50.69 (EPID L-2025-LLA-0025, L-2025-LLA-0058) Date: Thursday, August 7, 2025 6:34:07 PM

Jarrett, By letters dated February 3, 2025, and March 21, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML25034A143 and ML25080A172, respectively), NextEra Energy Seabrook, LLC submitted license amendment requests (LARs) to amend the license for Seabrook Station, Unit 1, Renewed Facility Operating License No. NPF-86. The proposed LARs would adopt Technical Specifications Task Force Traveler 505 (TSTF-505), Revision 2, Provide Risk-informed Extended Completion Times, RITSTF Initiative 4b, and the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors.

The U.S. Nuclear Regulatory Commission staff has reviewed the information provided in the LARs and has determined that additional information is needed to complete its review, as described in the request for additional information (RAI) shown below. Please respond to this RAI by September 8, 2025, which is approximately 30 days from the date of this e-mail communication. A publicly available version of this message will be placed in ADAMS as an official agency record. Please contact me if you have any questions regarding this request. Thank you, Richard Guzman Sr. PM, Division of Operating Reactor Licensing Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Phone: (301) 415-1030 Richard.Guzman@nrc.gov

=========================================================

REQUEST FOR ADDITIONAL INFORMATION

IN SUPPORT LICENSE AMENDMENT REQUESTS TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT

TSTF-505, REVISION 2, RISK-INFORMED EXTENDED COMPLETION TIMES INITIATIVE 4B AND

10 CFR 50.69, RISK-INFORMED CATEGORIZATION AND TREATMENT

OF STRUCTURES, SYSTEMS, AND COMPONENTS

NEXTERA ENERGY SEABROOK, LLC

SEABROOK STATION, UNIT 1

DOCKET NO. 50-443

By \ letter dated February 3, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25034A143), NextEra Energy Seabrook, LLC (NextEra, the licensee) submitted a license amendment request (LAR) for Seabrook Station Unit 1 (Seabrook). The proposed amendment would revise technical specification (TS) requirements to permit the use of risk-informed completion times (RICTs) for actions to be taken when limiting conditions for operation(LCOs) are not met. The proposed changes are based on Technical Specifications Task Force(TSTF) Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times -RITSTF [Risk-Informed TSTF] Initiative 4b, dated July 2, 2018 (ML18183A493). The U.S. Nuclear Regulatory Commission (NRC, the Commission) issued a final model safety evaluation (SE)approving TSTF-505, Revision 2, on November 21, 2018 (ML18269A041).

By \ letter dated March 21, 2025 (ML25080A172), NextEra Energy Seabrook, LLC (NextEra, the licensee) submitted a license amendment request (LAR) for Seabrook Station, Unit 1 (Seabrook). The proposed amendment would adopt Title 10 of the Code of Federal Regulations (10 CFR), Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, into the licensing basis for Seabrook. The provisions of 10 CFR 50.69 allow licensees to use an integrated, systematic, risk-informed process for categorizing structures, systems, and components (SSCs) according to their safety significance. A licensee that has adopted 10 CFR 50.69 may specify alternative treatments for SSCs that have low safety significance.

The NRC staff has determined that additional information is needed to support the completion of its review. The following is the NRC staffs request for additional information (RAI).

RAI-01 (APLA 505-01) - In-Scope LCOs and Corresponding PRA Modeling

The NRCs safety evaluation for NEI 06-09-A specifies that the LAR should provide a

comparison of the TS functions to the PRA modeled functions to show that the PRA modeling is consistent with the licensing basis assumptions or to provide a basis when there is a difference. Table E1-1 of LAR Enclosure 1 identifies each Limiting Condition for Operation (LCO) in the TSs proposed for inclusion in the RICT program. The table also describes whether the systems and components covered by the LCO are modeled in the PRA and, if so, presents both the design success criteria and PRA success criteria. For certain LCOs, the table explains that the associated structures, systems, and components (SSCs) are not modeled in the PRAs but will be represented using a surrogate event that fails the function performed by the SSC. For some LCOs, the LAR did not provide an adequate description for the NRC staff to conclude that the PRA modeling will be sufficient.

a. Regarding TS 3.4.4 Action d - One PORV Block Valve Inoperable, LAR Table E1-1 states: "While the Block Valves are required to successfully close to maintain RCS integrity in the DSC, the PRA model also considers scenarios where the Block Valves failing to open, in conjunction with RCS overpressure, may also jeopardize RCS integrity. Thus, the PRA success criteria considers both functions of the Block Valves."Given that the PRA model appropriately accounts for multiple failure modes of the block valves (e.g., failure to open, failure to close), how does the CRMP tool handle this distinction when TS 3.4.4 Action d is entered with one block valve inoperable? Specifically:
i. Does the CRMP tool differentiate between the various failure modes of the block valve (e.g., fail-to-open vs. fail-to-close)?

ii. How is the operator expected to input or represent this condition when calculating a RICT? iii. What is the PRA model success criterion of PORVs and associated block valves used in depressurization for feed-and-bleed? Does failure of one or both block valves to open considered as a key assumption and source of uncertainty.

b. Regarding TS 3.5.2, Action a - One ECCS Subsystem Inoperable, Table E1-2 provides a RICT estimate of 4.5 days, which is shorter than the front stop completion time. The staff notes that the Technical Specifications define each ECCS subsystem as consisting of the following components:

One OPERABLE centrifugal charging pump One OPERABLE safety injection pump One OPERABLE RHR heat exchanger One OPERABLE RHR pump An OPERABLE flow path* capable of taking suction from the refueling water storage tank on a Safety Injection signal and automatically transferring suction to the containment sump during the recirculation phase of operation

Given this definition, please describe in more detail the modeling assumptions used to derive the RICT estimate in Table E1-2. Specifically:

i. Does the PRA model treat the entire ECCS subsystem as failed, or are failures modeled at the component level (e.g., charging pump, safety injection pump, flow path)?

ii. How is the operator expected to represent an inoperable ECCS subsystem in the CRMP tool when calculating the RICT?

c. Regarding TS 3.6.2.1 - One Containment Spray System Inoperable, LAR Table E1-1 refers to see additional justification; however, no such justification was identified in the submittal. Please clarify the intended information related to the modeling of the Containment Spray System in the PRA, specifically as it pertains to the CRMP model.
d. Regarding TS 3.7.3 - One Primary Component Cooling Water (PCCW) Loop Inoperable, Table E1-2 reports an estimated RICT of 5.5 days.
i. Is this the only risk estimate applicable, or could alternative estimates apply depending on which specific portions or components of the PCCW system are inoperable?

ii. Does the PRA model assume the entire PCCW loop is inoperable, or are failures modeled at the component level (e.g., individual pumps, heat exchangers)? iii. How is the operator expected to input or represent an inoperable PCCW loop in the CRMP tool when calculating the RICT?

e. Regarding TS 3.6.1.3 Action b, TS 3.6.1.7 Action a, TS 3.6.1.7 Action b, and TS 3.6.3 Action a, LAR Table E1-1 states that a large pre-existing leak will be used as surrogate, and as a result a RICT estimate of 9.8 days was reported in Table E1-2. These TS conditions covered varied possible containment leakage scenarios. Explain what is meant by a large pre-existing leak. Is the size of the leak sufficient to capture multiple concurrent TS entries?
f. Regarding TS 3.8.1.1 Actions c.2.(a) and c.2.(b), it is noted in the Attachment 2 mark-up that entry into these action statements shall not be applied with conjunction with a risk-informed completion time. However, it is not clear to the NRC staff how failure of the Supplemental Electric Power System (SEPS) is credited in the model. It is also unclear as to its success criterion. Discuss how SEPS is credited in the model, its success criterion, and whether it should be an assumption and source of uncertainty.

RAI-02 (APLA 505-02)- Impact of Seasonal Variations

The Tier 3 requirement of Regulatory Guide (RG) 1.177, Revision 2, Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications", dated January 2021, stipulates that a licensee should develop a program that ensures that the risk impact of out-of-service equipment is appropriately evaluated prior to performing any maintenance activity.

Section 2.3.4 of NEI 06-09-A states, in part, that:

"If the PRA model is constructed using data points or basic events that change as a result of time of year or time of cycle, then the RICT calculation shalleither 1) use the more conservative assumption at all time, or 2) be adjustedappropriately to reflect the current (e.g., seasonal or time of cycle) configurationfor the feature as modeled in the PRA."

The LAR does not appear to specify the modeling adjustments needed to account for seasonal variations and what kind of adjustments will be made. Therefore, address the following to clarify the treatment of seasonal and time of cycle variations:

a. Explain how the RICT calculations address changes in PRA data points, basic events,and SSC operability constraints as a result of extreme weather conditions, seasonalvariations, or other environmental factors. Also, explain how these adjustments are made in the configuration risk management program (CRMP) model and how this approach is consistent with the guidance in NEI 06-09-A and its associated NRC final SE.
b. Describe the criteria used to determine when PRA adjustments due to extreme weatherconditions, seasonal variations, other environmental factors, or time of cycle variationsneed to be made in the CRMP model and what mechanism initiates these changes.

RAI-03 (APLA 505-03) - Determination of Key Sources of Uncertainty and Sensitivity Results

The NRC staff safety evaluation to NEI 06-09-A specifies that the LAR should identify key assumptions and sources of uncertainty and to assess and disposition each as to their impact on the RMTS application. NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, Main Report," (ADAMS Accession No. ML17062A466) presents guidance on the process of identifying, characterizing, and qualitative screening of model uncertainties. The NRC staff reviewed the uncertainty documents provided in the LAR for the internal events, internal flooding, and fire PRA and found that further clarification is necessary regarding the review of these assumptions and sources of uncertainty for this application. It is unclear what additional analysis was performed and documented to determine if any source of uncertainty could adversely impact any RICT calculation. In light of these observations, provide the following information:

a. Provide details of how the Seabrook PRA sources of uncertainty were evaluated as a potential key source of uncertainty for this application. Include in this discussion any documentation of this process.
b. Provide the results of sensitivity studies that determined the impact on risk for each associated source of uncertainty. Include in this discussion justification

that the sensitivity results demonstrate that the associated source of uncertainty does not adversely impact any RICT calculation.

RAI-04 (APLA 505-04) - Performance Monitoring

The NRC SE for NEI 06-09-A, states in part: The impact of the proposed change should be monitored using performance measurement strategies. NEI 06-09-A considers the use of NUMARC 93-01, Revision 4F, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants (ADAMS Accession No. ML18120A069), as endorsed by RG 1.160, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 4 (ADAMS Accession No. ML18220B281), for the implementation of the Maintenance Rule. NUMARC 93-01, Section 9.0, contains guidance for the establishment of performance criteria.

In addition, the NEI 06-09-A methodology satisfies the five key safety principles specified in RG 1.177, Revision 2 relative to the risk impact due to the application of a RICT. Moreover, NRC staff position C.3.2 provided in RG 1.177, Revision 2, for meeting the fifth key safety principle acknowledges the use of performance criteria to assess degradation of operational safety over a period. It is unclear how the licensees RICT program captures performance monitoring for the SSCs within the scope of the RMTS program. Therefore:

a. Confirm that the Seabrook Maintenance Rule program incorporates the use of performance criteria to evaluate SSC performance as described in NUMARC 93-01, as endorsed by RG 1.160.
b. Alternatively, describe the approach or method used by Seabrook for SSC performance monitoring, as described in NRC staff regulatory guidance C.3.2 of RG 1.177, Revision 2, for meeting the fifth key safety principle. In the description, include criteria (e.g., qualitative, or quantitative) along with the appropriate risk metrics, and explain how the approach and criteria demonstrate the intent to monitor the potential degradation of SSCs in accordance with the NRC SE for NEI 06-09-A.

RAI-05 (APLA 505-05) - PRA Update Process

Section 2.3.4 of NEI 06-09-A specifies, criteria shall exist in PRA configuration risk management to require PRA model updates concurrent with implementation of facility changes that significantly impact RICT calculations.

LAR Enclosure 7 states that if a plant change or a discovered condition is identified and can have significant impact on the RICT calculations, then an unscheduled update of the PRA models will be implemented. More specifically, the LAR states that if the plant changes meet specific criteria defined in the plant PRA and update procedures, including criteria associated with consideration of the cumulative risk impact, then the change will be incorporated into applicable PRA models without waiting for the next periodic PRA update. The LAR does not explain under what conditions an unscheduled update of the PRA model will be performed or the criteria defined in the plant procedures that will be used to initiate

the update.

Considering these observations, describe the conditions under which an unscheduled PRA update (i.e., more than once every two refueling cycles) would be performed and the criteria that would be used to require a PRA update. In the response, define what is meant by significant impact to the RICT Program calculations.

RAI-06 (APLA 505-06) - Treatment of Common Cause Failure

The NRC SE related to NEI 06-09, Revision 0 Section 2.2, states that, specific methods and guidelines acceptable to the NRC staff are [] outlined in RG 1.177 for assessing risk-informed TS changes. The NRC SE, Section 3.2, further states that consistency with the guidance of RG 1.174, Revision 1, and RG 1.177, Revision 1, is achieved by evaluation using a comprehensive risk analysis, which assesses the configuration-specific risk by including contributions from human errors and common cause failures.

The guidance in RG 1.177, Revision 2, Section 2.3.3.1, states that, CCF modeling of components is not simply dependent on the number of remaining in-service components; it is also dependent on the reason components were removed from service (i.e. whether for preventative or corrective maintenance).

The Seabrook LAR states:

For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either: 1. Numerically accounting for the increased possibility of CCF in the RICT calculation, or 2. Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

For emergent conditions/corrective maintenance, please specify whether Seabrook plans to numerically account for the increased possibility of CCF in the RICT calculation, or if they plan to implement RMAs.

RAI-07 (APLA 505-RCI-01) - Internal Events and Internal Flood PRA Peer Review and Assessment History

The NRC staff identified the need to confirm information with regard to the proposed LAR and is therefore asking the following requests for confirmation of information (RCIs) in RAI-07 and RAI-08.

Revision 2 of RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (ML090410014), and Revision 3 of RG 1.200, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities (ML20238B871) both describe acceptable approaches for determining the acceptability of a base PRA used in regulatory decision-making for commercial light-

water nuclear power plants. Both revisions of RG 1.200 endorse, with clarifications, the American Society of Mechanical Engineers (ASME) / American Nuclear Society (ANS) PRA Standard ASME/ANS RASa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications. Enclosure 2 of the LAR states that the PRA models at Seabrook Station have been assessed against RG 1.200, Revision 2.

The NRC staff needs to confirm its understanding of the internal events and internal flood PRA peer reviews, self-assessments, and independent closure reviews of peer review facts and observations (F&O) findings that were performed in accordance with RG 1.200.

Please confirm (i.e., respond with yes/no) the following summary is accurate or provide clarifications/corrections as needed:

  • In 1999, a review of all technical elements was performed for the internal events PRA using the industry PSA [probabilistic safety assessment] Certification process, the precursor to the PRA Standard.
  • In 2005, a focused peer review was performed for the internal events PRA for the elements of Accident Sequence Analysis (AS), Success Criteria (SC), and Human Reliability Analysis (HR), as well as configuration control. This assessment replaced the 1999 peer review for those elements that were in scope. This review was done using the PRA Standard that was current at the time (ASME RA-Sa-2003).
  • In 2009, a focused peer review was performed for the internal flood PRA. This review addressed all elements for internal flood PRA and was done using the currently endorsed PRA Standard (ASME/ANS RA-Sa-2009) and RG 1.200, Revision 2.
  • In 2012, a focused peer review was performed for the internal events PRA for the element of LERF Analysis (LE). This assessment replaced the 1999 peer review for that element. This review was done using the currently endorsed PRA Standard (ASME/ANS RA-Sa-2009) and RG 1.200, Revision 2.
  • In 2019, a focused peer review was performed on all elements affected by the PRA upgrade to convert from RISKMAN (support state) to CAFTA (linked fault tree). This review was done using the currently endorsed PRA Standard (ASME/ANS RA-Sa-2009) and RG 1.200, Revision 2.
  • Multiple self-assessments were performed against various versions of the PRA Standard and RG 1.200. The PRA capability assessments in 2010 and 2015 considered all internal events supporting requirements (SRs) and assessed them against ASME/ANS RA-Sa-2009 and RG 1.200, Revision 2.
  • Multiple F&O closure reviews were performed by independent assessment teams in accordance with NRC endorsed processes (i.e., NEI 17-07, Revision 2, and Appendix X to NEI0504, NEI 07-12 and NEI 12-13). The closure review completed in October 2017 assessed open F&O findings for the internal events and internal flood PRA models using the process in Appendix X to NEI 05-04. The F&O closure review performed in 2019 (completed April 2020) was performed as part of the 2019 peer review and closed additional internal flood PRA F&Os. The F&O closure performed in August 2024 assessed open internal events and fire PRA F&O findings using the process in NEI1707, Revision 2.

RAI-08 (APLA 505-RCI-02) - Closure of peer review findings

In Table E2-1 of Enclosure 2 of the LAR (Seabrook PRA Peer Review Findings), F&O findings LE-D6-01 and ES-B4 (03002) were identified as open. Based on its audit, the NRC staff understands that these findings have since been dispositioned and closed out using an independent assessment process. Please confirm (i.e., respond with yes/no) the following information is accurate or provide clarifications/corrections as needed:

An independent assessment team completed an F&O closure review in July 2025 for peer review findings LE-D6-01 and 03002, using the process in NEI1707, Revision 2, and confirmed that no PRA upgrades were performed as part of the resolution to the findings.

RAI-09 (APLA 50.69-RCI-01) - Internal Events and Internal Flood PRA Peer Review and Assessment History

Regulatory Guide (RG) 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance (ML061090627), endorses, with regulatory positions and clarifications, the Nuclear Energy Institute (NEI) guidance document NEI 00-04, Revision 0, 10 CFR 50.69 SSC [Structure, System, and Component] Categorization Guideline (ML052910035), as one acceptable method for use in complying with the requirements in 10 CFR 50.69. Section 3.1.1 of the LAR dated March21, 2025, states that NextEra will implement the risk categorization process of 10CFR 50.69 in accordance with NEI 00-04, Revision 0, as endorsed by RG 1.201.

The following RCIs in RAI-09 and RAI-10 are intended to help the NRC staff determine if the licensee has implemented the guidance appropriately in NEI 00-04, as endorsed by RG1.201, as a means to demonstrate compliance with all of the requirements in 10 CFR 50.69, including acceptability of the PRA models.

In 10 CFR 50.69(c)(1)(i) and (ii), the regulations require that a licensees PRA be of sufficient quality and level of detail to support the SSC categorization process, and must be subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC, and all aspects of the integrated, systematic process used to characterize SSC importance must reasonably reflect the current plant configuration and operating practices, and applicable plant and industry operational experience.

Revision 2 of RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (ML090410014), and Revision 3 of RG 1.200, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities (ML20238B871) both describe acceptable approaches for determining the acceptability of a base PRA used in regulatory decision-making for commercial light-water nuclear power plants. Both revisions of RG 1.200 endorse, with clarifications, the American Society of Mechanical Engineers (ASME) / American Nuclear Society (ANS) PRA Standard ASME/ANS RASa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications. Section 3.3 of the LAR states that the PRA models have been assessed against RG 1.200, Revision 2.

The NRC staff needs to confirm its understanding of the internal events and internal flood

PRA peer reviews, self-assessments, and independent closure reviews of peer review facts and observations (F&O) findings that were performed in accordance with RG 1.200.

Please confirm (i.e., respond with yes/no) the following summary is accurate or provide clarifications/corrections as needed:

  • In 1999, a review of all technical elements was performed for the internal events PRA using the industry PSA [probabilistic safety assessment] Certification process, the precursor to the PRA Standard.
  • In 2005, a focused peer review was performed for the internal events PRA for the elements of Accident Sequence Analysis (AS), Success Criteria (SC), and Human Reliability Analysis (HR), as well as configuration control. This assessment replaced the 1999 peer review for those elements that were in scope. This review was done using the PRA Standard that was current at the time (ASME RA-Sa-2003).
  • In 2009, a focused peer review was performed for the internal flood PRA. This review addressed all elements for internal flood PRA and was done using the currently endorsed PRA Standard (ASME/ANS RA-Sa-2009) and RG 1.200, Revision 2.
  • In 2012, a focused peer review was performed for the internal events PRA for the element of LERF Analysis (LE). This assessment replaced the 1999 peer review for that element. This review was done using the currently endorsed PRA Standard (ASME/ANS RA-Sa-2009) and RG 1.200, Revision 2.
  • In 2019, a focused peer review was performed on all elements affected by the PRA upgrade to convert from RISKMAN (support state) to CAFTA (linked fault tree). This review was done using the currently endorsed PRA Standard (ASME/ANS RA-Sa-2009) and RG 1.200, Revision 2.
  • Multiple self-assessments were performed against various versions of the PRA Standard and RG 1.200. The PRA capability assessments in 2010 and 2015 considered all internal events supporting requirements (SRs) and assessed them against ASME/ANS RA-Sa-2009 and RG 1.200, Revision 2.
  • Multiple F&O closure reviews were performed by independent assessment teams in accordance with NRC endorsed processes (i.e., NEI 17-07, Revision 2, and Appendix X to NEI0504, NEI 07-12 and NEI 12-13). The closure review completed in October 2017 assessed open F&O findings for the internal events and internal flood PRA models using the process in Appendix X to NEI 05-04. The F&O closure review performed in 2019 (completed April 2020) was performed as part of the 2019 peer review and closed additional internal flood PRA F&Os. The F&O closure performed in August 2024 assessed open internal events and fire PRA F&O findings using the process in NEI1707, Revision 2

RAI-10 (APLA 50.69-RCI-02) - Closure of peer review findings

In Attachment 3 of the LAR enclosure (Disposition and Resolution of Open Peer Review Findings and Self-Assessments Open Items), F&O findings LE-D6-01 and 03002 were identified as open. Based on its audit, the NRC staff understands that these findings have since been dispositioned and closed out using an independent assessment process.

Please confirm (i.e., respond with yes/no) the following information is accurate or provide clarifications/corrections as needed:

An independent assessment team completed an F&O closure review in July 2025 for peer review findings LE-D6-01 and ES-B4 (03002), using the process in NEI 1707, Revision 2, and confirmed that no PRA upgrades were performed as part of the resolution to the findings.

RAI-11 (APLC 505-01)

The regulation at Title 10 of theCode of Federal Regulations (10 CFR) 50.36, Technical specifications, provides the regulatory requirements for the content of the technical specifications (TS). It requires, in part, that a summary statement of the bases for such specifications shall be included by applicants for a license authorizing operation of a production or utilization facility. Specifically, 10 CFR 50.36(c) requires that TS include items in five specific categories related to station operation. These categories are (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls.

Item 7 of NEI 06-09, Revision 0-A, Section 2.3.1, Configuration Risk Management Process & Application of Technical Specifications, states, in part, that the impact of other external events risk shall be addressed in the RMTS program, and explains that one acceptable method for accomplishing this is performing a reasonable bounding analysis and applying it along with the internal events risk contribution in calculating the configuration risk and the associated RICT. The NRC staffs safety evaluation for NEI 06-09 states that where PRA models are not available, conservative or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT.

Section 2.2, Seismic, in enclosure 4 of the license amendment request (LAR) describes that the Seabrook Individual Plant Examination for External Events (IPEEE) seismic probabilistic risk assessment (SPRA) is not used as the direct basis for the seismic core damage frequency (SCDF) and seismic large early release frequency (SLERF) estimates but is used to provide input into the calculation process. The plant-level fragility and containment function fragility values used in the LAR are derived from the IPEEE SPRA. It is expected that all the information supporting the RICT analysis reflects current, as-operated plant conditions. However, the LAR does not discuss whether the licensee performed any post-IPEEE walkdowns to substantiate the continued applicability of the IPEEE information.

The licensee is requested to provide:

a. A summary of seismic walkdown(s) performed since the IPEEE or a statement that no seismic walkdowns have been performed since the IPEEE.
b. If a seismic walkdown was performed, an explanation of how the walkdown results support the validity of the seismic information used in the LAR, including identification of critical containment failure modes.
c. A discussion of any post-IPEEE identification of potential seismic-induced

spatial interactions relevant to the TSTF-505 RICT application.

RAI-12 (APLC 505-02)

Item 7 of NEI 06-09, Revision 0-A, Section 2.3.1, Configuration Risk Management Process & Application of Technical Specifications, states, in part, that the impact of other external events risk shall be addressed in the RMTS program, and explains that one acceptable method for accomplishing this is performing a reasonable bounding analysis and applying it along with the internal events risk contribution in calculating the configuration risk and the associated RICT. The NRC staffs safety evaluation for NEI 06-09 states that where PRA models are not available, conservative or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT.

Section 2.0, Technical Approach, in enclosure 4 of the LAR describes the proposed progressive screening approach, which includes:

  • A qualitative screening of hazards based on their limited potential impact, using a set of screening criteria (QL-1 to QL-6) listed in table E4-9 of this enclosure.
  • A quantitative screening based on conservative estimates of the hazard and consequences, using the screening criteria (QN-1, QN-2) listed in table E4-9 of this enclosure.

However, Table E4-9, Progressive Screening Approach for Addressing External Hazards, identifies the screening criteria as C1 to C5 for preliminary (qualitative) screening and PS1 to PS4 for progressive (quantitative) screening. The licensee is requested to clarify the discrepancy in screening criteria notation between section 2.0 and table E4-9.

RAI-13 (STSB 505-01)

The regulation at 10 CFR 50.36, Technical specifications, provides the regulatory requirements for the content of the technical specifications (TS). It requires, in part, that a summary statement of the bases for such specifications shall be included by applicants for a license authorizing operation of a production or utilization facility. Specifically, 10 CFR 50.36(c) requires that TS include items in five specific categories related to station operation. These categories are (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls.

In attachment 2 to the LAR, the proposed changes in markup for TS 3.3.1 would add a RICT to Action 11 (for turbine trip - stop valve closure), which begins, [w]ith the number of OPERABLE channels less than the Total Number of Channels The cross-reference table of proposed changes in attachment 4 to the LAR indicates that this change is similar to TSTF-505, Revision 2, NUREG-1431 TS 3.3.1 Condition R, which states, One Turbine Trip channel inoperable. Though similar, the NRC staff interprets the Seabrook TS phrase less than the total number to be equivalent to the standard TS phrase, one or more which includes a includes a loss of function condition.

Consistent with section 2.3 of TSTF-505, Revision 2, the exclusion criteria for the RICT program do not allow a Condition that represents a TS loss of specified safety function condition... Toward the end of section 2.3 of TSTF-505, Revision 2, there is an example of

one way to modify a TS that includes a loss of function condition to limit the application of a RICT. Provide justification for this variation to apply a RICT to Action 11 or revise.

RAI-14 (STSB 505-02)

A variation to TSTF-505, Revision 2, listed in section 2.4.3 of attachment 1 to the LAR describes proposed changes to TS 3.8.1.1 to delete ACTION b.2(a), and ACTION b.2(b) in their entirety However, the TS 3.8.1.1 markup in attachment 2 to the LAR does not propose to delete these sub-actions; instead a note would be added.

a. Clarify what changes are proposed to TS 3.8.1.1 Action b and its subparts so that all parts of the LAR that discuss this Action are consistent.
b. The proposed note in the markup for TS 3.8.1.1 Action b.2 states ACTIONs b.2(a) and b.2(b) shall not be applied in conjunction with a Risk Informed Completion Time. Provide a discussion or example of how this note will be implemented.

RAI-15 (STSB 505-03)

A variation to TSTF-505, Revision 2, listed in section 2.4.3 of attachment 1 to the LAR describes proposed changes to TS 3.8.1.1 to delete ACTION c.3(a), and ACTION c.3(b) provisions associated with SEPS availability. However, the TS 3.8.1.1 markup in attachment 2 to the LAR renumbers and retains these sub-actions and inserts a note.

a. Clarify what changes are proposed to TS 3.8.1.1 Action c and its subparts so that all parts of the LAR that discuss this Action are consistent.
b. The proposed note in the TS 3.8.1.1 Action c.2 markup states ACTIONs c.2(a) and c.2(b) shall not be applied in conjunction with a Risk Informed Completion Time. Provide a discussion or example of how this note will be implemented.

RAI-16 (STSB 505-04)

In attachment 2 to the LAR, the proposed changes in markup for TS 3.6.3 would add a RICT to the Action for inoperable equipment. The proposed Action begins, [w]ith one or more of the isolation valve(s) inoperable The NRC staff interprets the phrase one or more to include a loss of function condition. Consistent with section 2.3 of TSTF-505, Revision 2, the exclusion criteria for the RICT program do not allow a Condition that represents a TS loss of specified safety function condition... Toward the end of section 2.3 of TSTF-505, Revision 2, there is an example of one way to modify a TS that includes a loss of function condition to limit the application of a RICT.

Provide justification for this variation to apply a RICT to the Seabrook TS 3.6.3 Action or revise.

RAI-17 (STSB 505-05)

Proposed TS 6.7.6.p, "Risk Informed Completion Time Program," as marked up in LAR attachment 2, appears to be missing the last paragraph of the administrative controls for the RICT Program in TSTF-505, Revision 2, (NUREG-1431) Standard TS 5.5.18 (i.e., paragraph e). Additionally, the NRC staff notes that the third sentence in the paragraph e Standard TS markup has some unclear phrasing, namely PRA methods used to support this license amendment The NRC staff recognizes that the model SE for TSTF-505, Revision 2, contains improved phrasing, PRA methods approved for use with this program...

In lieu of the original phrasing in Standard TS 5.5.18 paragraph e, consider whether the phrases such as PRA methods used to support Amendment # xxx or, as discussed in the TSTF505 model SE, PRA methods approved for use with this program would provide more clarity for proposed Seabrook TS 6.7.6.p paragraph e. Overall, provide justification for the missing paragraph or revise.

RAI-18 (EEEB 505-01) - TS LCO 3.8.2.1, Action a.

10 CFR Appendix A, General Design Criterion (GDC)17 requires, in part, that both offsite and onsite electrical power systems be provided to permit the functioning of systems, structures, and components (SSCs) important to safety. The safety function for each system, assuming other is not functioning, assures fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded, and the core is cooled, and containment integrity and other vital functions are maintained for postulated accidents.

LCO 3.8.2.1.a. and b., indicates there are two trains for direct current (dc) electrical sources-Each train, A and B, have two battery banks and two battery chargers with only one battery bank and two chargers per train required to be operable. LCO 3.8.2.1, Action a. is for the inoperability of one of two battery banks in one train.

UFSAR Section 8.3.2.1 reveals that the 125 Vdc power system has battery chargers, station batteries, and the 125 Vdc distribution system. Each battery charger supplies steady-state loads with its battery being for transient loads and the reserve power source for charger failure in some way. There are four batteries (battery banks) for the Seabrook single unit plant. Each battery and thus each charger supplies a Class 1E dc bus which powers: inverter for a singular vital instrument bus; Class 1E dc loads; and controls for Class 1E systems for Engineered Safety Features (ESF). UFSAR Section 8.3.2.1.d. states each battery, battery charger, and their loads comprise a single load group with each train having two load groups.

USFAR Section 8.3.1.1.d. indicates that 120V Instrumentation and Control Power System powers four vital uninterruptible power supply (UPS) units to supply nuclear steam supply system (NSSS) instrument channels I, II, III, and IV. The inverters, supplied by the four batteries for Seabrook plant, are the same four vital UPS units. UFSAR Table 8.3-3 shows that channels I and III are for train A and channels II and IV are for Train B. UFSAR Section 7.1.2.3 indicates that there are four separate protection sets (channels) I, II, III, and IV. UFSAR Table 7.3-1 indicates that most ESF elements require two of the four channels to trip for safe shutdown - channels I and III or II and IV. Each NSSS instrument channel is powered by a dc load group which means that two load groups from one train are necessary for safe shutdown.

The staff notes the following for DSC for TS LCO 3.8.2.1, Action a, in Table E1-1 of the LAR:

a. DSC appears inconsistent with UFSAR since DSC indicates that a battery always assists its charger during an accident, but the UFSAR states that battery and charger for each redundant load group may address an accident together only if accident loads exceed charger full amperage output rating.

Please clarify or explain this inconsistency.

b. Table E1-1 provides DSC and PRA success criteria for each LCO Action for the rated DBA for Seabrook which is a concurrent loss of coolant accident (LOCA) and loss of offsite power (LOOP). Please clarify for this DSC whether PRA model assumes a battery operates for that DBA and, if so, under what conditions will that happen?
c. DSC typically indicates minimum SSCs required to operate for safe shutdown for DBA, The DSC for TS LCO 3.8.2.1, Action a seems only to indicate one dc load group of one train for safe shutdown, which corresponds to one NSSS instrument channel while UFSAR indicates that two NSSS channels from one train are required for safe shutdown. Please clarify and explain the discrepancy between the DSC and UFSAR.

RAI-19 (EEEB 505-02) - TS LCO 3.8.3.1, Action c.

GDC 17 requires, in part, that both offsite and onsite electrical power systems be provided to permit the functioning of systems, structures, and components (SSCs) important to safety. The safety function for each system, assuming other is not functioning, assures fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded, and the core is cooled, and containment integrity and other vital functions are maintained for postulated accidents.

UFSAR Section 8.3.2.1 reveals that the 125 Vdc power system has two trains with each having two battery chargers, two station batteries, and 125 Vdc distribution systems. Each battery charger supplies steady-state loads with its battery being for transient loads and reserve power source for charger failure. There are four batteries (battery banks) for the Seabrook single unit plant. Each battery and thus each charger supplies a Class 1E dc bus which powers: inverter for a singular vital instrument bus; Class 1E dc loads; and controls for Class 1E systems for Engineered Safety Features (ESF). UFSAR Section 8.3.2.1.d. states each battery, battery charger, and their loads are a load group with each train having two load groups.

USFAR Section 8.3.1.1.d. indicates that 120V Instrumentation and Control Power System powers four vital uninterruptible power supply (UPS) units which supply NSSS instrument channels I, II, II, and IV. The inverters supplied by the four batteries for Seabrook plant are the same four vital UPS units. UFSAR Table 8.3-3 shows that channels I and III are for train

A and channels II and IV are for Train B. UFSAR Section 7.1.2.3 indicates that there are four separate protection sets (channels) I, II, III, and IV. UFSAR Table 7.3-1 indicates that most ESF trip functions require two of four channels to trip - channels I and III or II and IV. The staff notes the following for DSC for TS LCO 3.8.3.1, Action C, in Table E1-1 of the LAR: DSC appears inconsistent with the UFSAR which requires two energized dc buses to power two NSSS instrument channels from one train for safe shutdown whereas the DSC only requires one energized dc bus. Please clarify or explain the inconsistency. RAI-20 (EEEB 505-03) - TS LCO 3.8.1.1, Action c.3. GDC 17 requires, in part, that both offsite and onsite electrical power systems be provided to permit the functioning of systems, structures, and components (SSCs) important to safety. The safety function for each system, assuming other is not functioning, assures fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded, and the core is cooled, and containment integrity and other vital functions are maintained for postulated accidents LAR Attachment 1, Description and Assessment of the Proposed Change, Section TS 3.8.1.1. A.C. Source - Operating states, in part (emphasize added): The proposed change revises ACTION c.2 to require restoration of at least one of the inoperable sources within 12 hours or in accordance with the Risk Informed Completion Time Program. The proposed change is appropriate since it is consistent with STS 3.8.1, ACTION D.2, of TSTF-505, Revision 2. The proposed change additionally deletes ACTION c.3 in its entirety, including the ACTION c.3(a), and ACTION c.3(b) provisions associated with SEPS availability. The proposed change is acceptable since ACTION c) need not address the operable offsite circuit(s) or EDG(s) and since deletion of the authorization to extend the front stop CT from 72 hours to 14 days based on SEPS availability serves to remove the ambiguity that would result from RICT calculations of less than 14 days. The staff notes the following: In LAR Attachment 2, Proposed Technical Specification Changes (Mark-Up), the markup for TS 3.8.1.1, Action c indicates the deletion of Action c.3 only while maintaining Actions c.3(a) and c.3(b). It appears that the proposed changes mentioned above in LAR Attachments 1 and 2 are inconsistent. Please clarify the following: The current TS 3.8.1.1, Action c.3 has the front stop of 72 hours. Therefore, c.3. (a) and c.3(b) are applied to extend the CT from 72 hours to 14 days. The proposed TS 3.8.1.1, Action c.2 has the front stop of 12 hours, and the proposed TS 3.8.1.1, Actions c.2(a) and c.2(b) have the front stop of 72 hours. Please also explain why the proposed Action c.2 has two front stops (i.e., 12 hours and 72 hours).

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