ML25171A122

Risk-Informed Completion Time TSTF-505 and 10 CFR 50.69 Audit Questions (EPIDs L-2025-LLA-0025, L-2025-LLA-0058)

ML25171A122
Person / Time
Site:	Seabrook  (NPF-086)
Issue date:	06/20/2025
From:	Lantigua R Plant Licensing Branch 1
To:	Mack J NextEra Energy Seabrook
Lantigua, R, NRR/DORL/LPLI,
References
EPID L-2025-LLA-0025, EPID L-2025-LLA-0058
Download: ML25171A122 (1)

Text

1 Ricardo Lantigua From:

Ricardo Lantigua Sent:

Friday, June 20, 2025 12:24 PM To:

Mack, Jarrett

Subject:

Seabrook 1 - Risk-Informed Completion Time TSTF-505 and 10 CFR 50.69 Audit Questions (EPIDs L-2025-LLA-0025, L-2025-LLA-0058)

Jarrett, By letters dated February 10, 2025 and March 28, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML25080A172 and ML25034A143), 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 Speci"cations 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.

On March 28, 2025 and April 21, 2025, the NRC sta issued audit plans (ML25057A434 and ML25114A113) that conveyed intent to conduct a regulatory audit to support its review of the subject licensing actions. Based on the commonalities between the LARs and subsequent overlap in technical content and review personnel, the sta is conducting a combined audit that addresses both LARs. In the audit plan, the NRC sta requested an electronic portal setup and provided a list of documents to be added to the online portal. The audit plan also indicated that the NRC may request information and audit meetings/interviews throughout the audit period. The NRC sta has performed an initial review of the list of documents and is developing a list of audit questions.

The "rst set of audit questions are provided below. Please post the response(s) for the questions to the online portal as the response is completed (but no later than one week before the date of the scheduled audit meeting). The dates and times for the audit meetings have not been set; however, the sta is targeting middle of July, 2025 to conduct the initial audit discussions via MS Team teleconference call.

The proposed agenda for the audit discussions will also be provided on a later date. Please contact me at any time prior if a clari"cation discussion is needed. We look forward to discussing these questions and NextEras responses during the virtual audit meeting.

Thank you, Ricardo Lantigua, Project Manager, LPL1 Division of Operating Reactor Licensing Oice of Nuclear Reactor Regulation AUDIT QUESTIONS LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO

2 ADOPT TSTF-505, REVISION 2 NEXTERA ENERGY SEABROOK STATION DOCKET NO. 50-443 APLA Questions APLA Audit Question 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? If not, explain why.

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:

a. One OPERABLE centrifugal charging pump

b. One OPERABLE safety injection pump

c. One OPERABLE RHR heat exchanger

d. One OPERABLE RHR pump

e. 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

3 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.

APLA Audit Question 02 - Determination of Key Sources of Uncertainty and Sensitivity Results The NRC staff safety evaluation to NEI 06-09 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:

4 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.

APLA Question 03 - 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 shall either 1) use the more conservative assumption at all time, or 2) be adjusted appropriately to reflect the current (e.g., seasonal or time of cycle) configuration for 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, seasonal variations, 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 weather conditions, seasonal variations, other environmental factors, or time of cycle variations need to be made in the CRMP model and what mechanism initiates these changes.

APLA Question 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.

5 b) Alternatively, describe the approach or method used by Seabrook for SSC performance monitoring, as described in NRC staff position 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.

APLA Audit Question 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.

APLB Questions APLB Audit Question 1 - Reduced Transient Heat Release Rates (HRRs)

The key factors used to justify using transient fire reduced heat release rates (HRRs) below those prescribed in NUREG/CR-6850 are discussed in the June 21, 2012, letter from Joseph Giitter, U.S. Nuclear Regulatory Commission, to Biff Bradley, NEI, "Recent Fire PRA Methods Review Panel Decisions and EPRI 1022993, Evaluation of Peak Heat Release Rates in Electrical Cabinet Fires," (ADAMS Package Accession No. ML12172A406).

If any reduced transient HRRs below the bounding 98% HRR of 317 kW from NUREG/CR-6850 were used, discuss the key factors used to justify the reduced HRRs. Include in this discussion:

a. Identification of the fire areas where a reduced transient fire HRR is credited and what reduced HRR value was applied.

b. A description for each location where a reduced HRR is credited, and a description of the administrative controls that justify the reduced HRR including how location-specific attributes and considerations are addressed. Include a discussion of the required controls for ignition sources in these locations and the types and quantities of combustible materials needed to perform maintenance. Also, include discussion of the personnel traffic that would be expected through each location.

c. The results of a review of records related to compliance with the transient combustible and hot work controls.

APLB Audit Question 2 - Treatment of Sensitive Electronics

6 FAQ 13-0004, Clarifications on Treatment of Sensitive Electronics (ADAMS Accession No. ML13322A085) provides supplemental guidance for application of the damage criteria provided in Sections 8.5.1.2 and H.2 of NUREG/CR-6850, Volume 2, for solid-state and sensitive electronics.

a. Describe the treatment of sensitive electronics for the FPRA and explain whether it is consistent with the guidance in FAQ 13-0004, including the caveats about configurations that can invalidate the approach (i.e., sensitive electronics mounted on the surface of cabinets and the presence of louver or vents).

b. If the approach cannot be justified to be consistent with FAQ 13-0004, then justify that the treatment of sensitive electronics has no impact on the RICT calculations.

c. As an alternative to item b above, add an implementation item to replace the current approach with an acceptable approach prior to the implementation of the RICT Program. Include a description of the replacement method along with justification that it is consistent with NRC accepted guidance.

APLB Question 3 - Obstructed Plume Model NUREG-2178, Volume 1 "Refining and Characterizing Heat Release Rates from Electrical Enclosures During Fire (RACHELLE -FIRE), Volume 1: Peak Heat Release Rates and Effect of Obstructed Plume," (ADAMS Accession No. ML16110A14016) contains refined peak HRRs, compared to those presented in NUREG/CR-6850, and guidance on modeling the effect of plume obstruction. Additionally, NUREG-2178 provides guidance that indicates that the obstructed plume model is not applicable to cabinets in which the fire is assumed to be located at elevations of less than one-half of the cabinet.

a. If obstructed plume modeling was used, then indicate whether the base of the fire was assumed to be located at an elevation of less than one-half of the cabinet.

b. Justify any modelling in which the base of an obstructed plume is located at less than one half of the cabinet's height.

c. As an alternative to item b above, add an implementation item to remove credit for the obstructed plume model in the FPRA prior to the implementation of the RICT Program.

APLB Audit Question 4 - Well-Sealed MCC cabinets Guidance in Frequently Asked Question 08-0042 from Supplement 1 of NUREG/CR-6850 applies to electrical cabinets below 440 V. With respect to Bin 15 as discussed in Chapter 6, it clarifies the meaning of "robustly or well-sealed." Thus, for cabinets of 440V or less, fires from well-sealed cabinets do not propagate outside the cabinet. For cabinets of 440 V and higher, the original guidance in Chapter 6 remains and requires that Bin 15 panels which house circuit voltages of 440 V or greater are counted because an arcing fault could compromise panel integrity (an arcing fault could burn through the panel sides, but this should not be confused with the high energy arcing fault type fires)." Fire PRA FAQ 14-0009, Treatment of Well-Sealed MCC Electrical Panels Greater than 440V (ADAMS Accession No. ML15119A176) provides the technique for evaluating fire damage from MCC cabinets having a voltage greater than 440V. Therefore, propagation of fire outside the ignition source panel must be evaluated for all MCC cabinets that house circuits of 440 V or greater.

a.

Describe how fire propagation outside of well-sealed MCC cabinets greater than 440 V is evaluated.

7

b.

If well-sealed cabinets less than 440 V are included in the Bin 15 count of ignition sources, provide justification for using this approach as this is contrary to the guidance.

APLB Audit Question 5 - Influence Factors for Transient Fires NUREG/CR-6850 Section 6 and FAQ 12-0064 Hot Work/Transient Fire Frequency Influence Factors" (ADAMS Accession No. ML12346A488) describe the process for assigning influence factors for hot work and transient fires. Provide the following regarding application of this guidance:

a) Indicate whether the methodology used to calculate hot work and transient fire frequencies applies influencing factors using NUREG/CR-6850 guidance or FAQ 12-0064 guidance.

b) Indicate whether administrative controls are used to reduce transient fire frequency, and if so, describe and justify these controls c) Indicate whether you have any combustible control violations and discuss your treatment of these violations for the assignment of transient fire frequency influence factors. For those cases where you have violations and have assigned an influence factor of 1 (Low) or less, indicate the value of the influence factors you have assigned and provide your justification.

d) If you have assigned an influencing factor of "0" to Maintenance, Occupancy, or Storage, or Hot Work for any fire PAUs provide justification e) If a weighting factor of "50" was not used in any fire PAU, provide a sensitivity study that assigns weighting factors of "50" per the guidance in FAQ 12-0064.

APLB Audit Question 6 - Fire Scenario Treatment of the Main Control Board Traditionally, the cabinets on front face of the Main Control Board (MCB) have been referred to as the MCB for purposes of fire PRA. Appendix L of NUREG/CR-6850, EPRI/NRC Fire PRA Methodology for Nuclear Power Facilities (ADAMS Accession Nos. ML052580075) provides a refined approach for developing and evaluating those fire scenarios. Fire PRA FAQ 14-0008, Main Control Board Treatment dated July 22, 2014 (ADAMS Accession No. ML14190B307) clarifies the definition of the MCB and effectively provides guidance for when to include the cabinets on the back side of the MCB as part of the MCB for fire PRA. It is important to distinguish between MCB and non-MCB cabinets because misinterpretation of the configuration of these cabinets can lead to incomplete or incorrect fire scenario development. This FAQ also provides several alternatives to NUREG/CR-6850 for using Appendix L to treat partitions in an MCB enclosure. Therefore, address the following:

a. Briefly describe the main control room MCB configuration, and use the guidance in FAQ 14-0008, to determine whether cabinets on the rear side of the MCB are a part of the MCB. Provide your justification using the FAQ guidance.

b. If the cabinets on the rear side of the MCB are part of a single integral MCB enclosure using the definition in FAQ 14-0008, then confirm that guidance in FAQ 14-0008 was used to develop fire scenarios in the MCB and determine the frequency of those scenarios

c. If the cabinets on the rear side of the MCB are part of a single integral MCB enclosure and the guidance in FAQ 14-0008 was not used to develop fire scenarios involving the MCB, then provide a description of how the fire scenarios for the backside cabinets are developed and an explanation of how the treatment aligns with NRC accepted guidance.

d. If in response to parts (c) above, the current treatment of the MCB and those cabinets on the rear side of the MCB cannot be justified using NRC accepted guidance, then justify that the treatment has no impact on the RICT calculations. Alternatively, propose a mechanism that ensures that the fire PRA is updated to treat the MCB enclosure consistent with the guidance in FAQ 14-0008, prior to implementation of the RICT program.

8 APLC Questions APLC Audit Question 01 Section 2.1.2, Tornado Missiles, in Enclosure 4, Information Supporting Justification of Excluding Sources of Risk Not Addressed by PRA Models, of the LAR includes the following statement:

The conservative estimate of missiles potentially damaging safety-related SSCs is on the order of 1E-6/yr or lower. Since additional failures or unavailability of SSCs would be needed to result in a core damage sequence, and the tornado strike frequencies used in the probabilistic missile analysis are conservative, it is estimated that the CDF associated with missiles is less than 1E-6/yr. CDFs associated with RICT LCO configurations are also estimated to be less than 1E-6/yr. Even if a single missile strike could result in core damage with other SSCs unavailable, individual SSC failure frequencies from tornado missiles are all less than 1E-6/yr, and most are less than 1E-8/yr. Therefore, the tornado missile hazard can be screened using Criteria C1 and PS4 from Table E4-9.

To confirm the licensees conclusion, the licensee is requested to provide sample calculations demonstrating that the configuration risk due to tornado missiles remains negligibly small for RICT LCO configurations considered in the TSTF-505 LAR.

APLC Audit Question 02 Section 2.2, Seismic, in enclosure 4 of the LAR states that the Seabrook IPEEE SPRA is not used as the direct basis for the SCDF and 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 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.

APLC Audit Question 03 In estimating the SLERF penalty for RICT calculations, identification of the limiting containment fragility is a key consideration. This includes evaluating potential failure modes associated with containment integrity, isolation, and bypass.

Section 2.2.7, Seismic Large Early Release Frequency, in enclosure 4 of the LAR states that the containment structure has high seismic capacity and that seismic-induced failure of containment isolation was considered in SLERF penalty determination. However, the LAR does not appear to consider containment bypass as a potential failure mode, which could result in early, unmitigated release.

The licensee is requested to provide an assessment of containment bypass and its potential impact on the SLERF penalty estimation.

9 APLC Audit Question 04 (Editorial)

In Section 2.0, Technical Approach, in enclosure 4 of the LAR, the licensee states:

• 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 and update the LAR to ensure consistent terminology between the text and table E4-9.

STSB Questions STSB Audit Question 01 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.

STSB Audit Question 02 A variation to TSTF-505, Revision 2, listed in section 2.4.3, 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 but instead inserts a note.

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 TS 3.8.1.1 Action b.2 markup 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.

STSB Audit Question 03 A variation to TSTF-500, Revision 2, listed in section 2.4.3, 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.

10 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.

EICB Question EICB Audit Question 01 (this question was asked as a portal document request, so an answer has already been provided; merely inserting it onto this list to indicate we would like to discuss the answer during the audit)-

Justifications of the conservative claims of using the SSPS as surrogates for RPS and ESFAS in the PRA modeling EEEB Questions EEEB Audit Question 01-TS LCO 3.8.2.1, Action a.

LCO 3.8.2.1.a and b indicate that there are two trains for direct current (dc) electrical sources-Each train A and B has 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, 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 elements require two of 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.

Design success criteria (DSC) in Table E1-1 for TS LCO 3.8.2.1, Actions b. (only question 1. below) and a.

1. 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.

2. 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 the worst-case conditions for which battery operates and whether that is captured in PRA model?

3. The DSC should indicate minimum SSCs required to operate for safe shutdown for DBA, but the DSC seems only to indicate one dc load group of one train which corresponds to one

11 NSSS instrument channel with two channels from one train required for safe shutdown.

Please clarify and explain this discrepancy.

EEEB Audit Question 02 - TS LCO 3.8.3.1, Action c 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.

1. 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 inconsistency.

EEEB Audit Question 03 - TS LCO 3.8.1.1, Action c.2 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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 14 days based on SEPS availability serves to remove the ambiguity that would result from RICT calculations of less than 14 days.

However, 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 mentioned above proposed changes are not consistent. Please clarify what the proposed change for TS 3.8.1.1, Action c is. If the proposed change is reflected in the TS markup of Attachment 2, please explain the following:

The current TS 3.8.1.1, Action c.3 has the front stop of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Therefore, c.3.(a) and c.3(b) are applied to extend the CT from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 14 days. However, Action c.2 has the front stop of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the proposed TS 3.8.1.1, Actions c.2(a) and c.2(b) have the front stop of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Please explain why the proposed Action c.2 has two front stops (i.e., 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />)

12 SNSB Question SNSB Audit Question 01 LAR Attachment 5, Evaluation of Plant-Specific Variations from TSTF-505, Revision 2, contains the following statement in the description of TS 3.3.1, Table 3.3-1, Reactor Protection System Instrumentation Variations, Power Range Neutron Flux:

Seabrook TS 3.3.1, Table 3.3-1, ACTION 2, requires the applicable power range neutron flux channel to be placed in the tripped position within one hour of inoperability. ACTION 2 differs from the STS of NUREG 1431 (Reference 2), and thereby TSTF-505, Revision 2, in not additionally requiring either a thermal power reduction to less than 75% of RATED THERMAL POWER and the power range neutron flux trip setpoint is reduced to less than or equal to 85% of RATED THERMAL within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or the QUADRANT POWER TILT RATIO must be verified within limit every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

However, this does not appear to be accurate based on comparing Seabrook TS 3.3.1 with the STS of NUREG 1431. The differences are summarized in the table below:

Table 3.3.1-1 Function Seabrook TS Action Westinghouse STS Condition Power Range Neutron Flux, High Setpoint Action 2: With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable Channel is placed in the tripped condition within 72

hours,

b. The Minimum Channels OPERABLE requirement is met; however, one channel may be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing per Specification 4.3.1.1, and

c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per Specification 4.2.4.2.

Condition D: D.1.1 Place channel in trip. [within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or RICT]

AND D.1.2 Reduce THERMAL POWER to 75% RTP.

[within 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> or RICT]

OR D.2.1 Place channel in trip.

[within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or RICT]

AND D.2.2

---

NOTE--------------

Only required to be performed when the Power Range Neutron Flux input to QPTR is inoperable.

Perform SR 3.2.4.2. [once per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s]

Power Range Neutron Flux, Low Setpoint Action 2: same as above Condition E: Place channel in trip [within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or RICT]

Power Range Neutron Flux, High Positive Rate Action 2: same as above Condition E: same as above

13 Explain the difference between the statement in the LAR, Attachment 5, and the Seabrook TS/STS and provide any additional justification necessary for this site-specific variation.

AUDIT QUESTIONS TO SUPPORT THE REVIEW OF LICENSE AMENDMENT REQUEST TO ADOPT 10 CFR 50.69, RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS NEXTERA ENERGY SEABROOK STATION, UNIT 1 DOCKET NO. 50-443 By letter dated March 21, 2025 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML25080A172), NextEra Energy Seabrook, LLC (NextEra, the licensee) submitted a license amendment request (LAR) for Seabrook Station, Unit 1 (Seabrook). The proposed LAR 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.

By letter dated March 28, 2025, the U.S. Nuclear Regulatory Commission (NRC) staff issued an audit plan (ML25057A434) for the LAR to adopt TSTF-505 that conveyed its intent to conduct a regulatory audit to support its review of the associated license amendment request. In the audit plan, the NRC staff requested that an electronic portal be set up and provided a list of documents to be added to the portal. By email dated April 21, 2025 (ML25114A113), the NRC staff expanded on the audit request for the TSTF-505 LAR to include information for the 50.69 LAR so that audits for both LARs could be conducted concurrently. The NRC staff from the Division of Risk Assessment, Probabilistic Risk Assessment (PRA) Licensing Branch A (APLA) and Branch C (APLC) have performed an initial review of the 50.69 LAR and the documents in the portal and identified the following topics and developed the following questions to discuss during the audit.

Background for Audit Questions 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 March 21, 2025, states that NextEra will implement the risk categorization process of 10 CFR 50.69 in accordance with NEI 00-04, Revision 0, as endorsed by RG 1.201.

The following questions and topics for discussion are intended to help the NRC staff determine if the licensee has implemented the guidance appropriately in NEI 00-04, as endorsed by RG 1.201, as a means to demonstrate compliance with all of the requirements in 10 CFR 50.69, including acceptability of the PRA models. The following questions and topics for discussion may result in formal requests for additional information (RAIs) or requests for confirmation of information (RCIs).

Additional Background for APLA Topics of Discussion and Audit Questions

14 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 RA-Sa-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.

As discussed in RG 1.200, it is recognized that a PRA may not satisfy each technical requirement to the same degree (i.e., capability category as used in the ASME/ANS PRA Standard); that is, the capability category achieved for the different technical requirements may vary. This variation can range from (1) the minimum needed to meet the characteristics and attributes for each technical element to (2) the minimum to meet current good practice (i.e., state-of-practice) for each technical element. Further, which capability category is needed to be met for each technical requirement is dependent on the specific application. In general, the NRC staff considers Capability Category II (CC-II) of the ASME/ANS PRA Standard to provide a level of detail that is acceptable for the majority of applications. However, for some applications, Capability Category I (CC-I) may be acceptable for some requirements.

APLA Topic for Discussion 1 - Peer review and gap assessment history In order to review the underlying technical adequacy of a licensees TSTF-505 and 10 CFR 50.69 programs, the NRC staff need to have an understanding of the details of PRA peer reviews and gap assessments which were performed in accordance with RG 1.200. To have a better understanding of the Seabrook peer review results, the NRC staff reviewed the information in Section 3.3 of the 50.69 LAR against information presented on the portal (i.e., SBK-PRA-QUALITY R0, Seabrook Station PRA 2005 Peer Review, and Enercon Report NEECORP00012-REPT-005) and in previous submittals (i.e., LAR to adopt TSTF-425 (ML13155A002)).

Specifically, Section 3.3 of the 50.69 LAR enclosure states: The Seabrook internal events and internal flood PRA models were peer reviewed in August 2005 (full scope), in January 2010 (internal flood focused scope),

and in June 2012 (LERF focused scope) applying NEI 05-04, the ASME/ANS PRA Standard and RG 1.200, Revision 2. However, it is the staffs understanding that the 2005 peer review was not full scope and was not against ASME/ANS PRA Standard RA-Sa-2009 and RG 1.200, Revision 2. Therefore, the staff would like to confirm its understanding of the PRA peer review history.

APLA Audit Question 01 - Internal Events and Internal Flood PRA peer review history Please confirm the following and provide clarifications/corrections as needed:

In 1999, a review of all technical elements was performed using the industry PSA Certification process, the precursor to the PRA Standard.

In 2005, a focused peer review was performed for the elements AS, SC and 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 Flooding and was done using the currently endorsed PRA Standard (ASME/ANS RA-Sa-2009) and RG 1.200, Revision 2.

15 In 2012, a focused peer review was performed for the element 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 to CAFTA. 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. 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 and Appendix X to NEI 05-04, 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. 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.

APLA Topic for Discussion 2 - Credit for FLEX and other portable equipment of the LAR enclosure (Disposition of Key Assumptions/Sources of Uncertainty) states that FLEX

[diverse and flexible coping strategies] is not currently credited in the PRA models. However, the information provided on the portal in response to Audit Request 3 explains that the internal events and fire PRA model credit two FLEX actions related to providing DWST [demineralized water storage tank] gravity feed to [the]

CST [condensate storage tank] and to locally control EFW [emergency feedwater] pumps to match decay heat. It is the NRC staffs understanding that these actions do not credit portable FLEX equipment and instead rely on existing permanently installed equipment.

Additionally, the information provided on the portal explains that there is a separate action involving a portable CT [cooling tower] makeup pump credited in the PRA models that is not based on FLEX procedures.

APLA Audit Question 02 - Portable FLEX equipment Please confirm that fixed and/or portable FLEX equipment is not credited in the Seabrook PRA models. If it is modeled in any internal or external event PRA models, discuss its modeling details and impact to both TSTF-505 and 10 CFR 50.69 programs.

APLA Audit Question 03 - Portable cooling tower (CT) makeup (MU) pump credit In order to support risk-informed categorization, PRA models need to reflect the as-built, as-operated plant.

The staff is particularly interested in understanding the basis for equipment failure probabilities and human error probabilities (HEPs) associated with portable equipment credited in the PRA models. In order to determine if additional information is required, please address the following:

a. Please confirm that the portable pump referred to on the portal is part of the same portable cooling tower makeup system referred to in the Seabrook Technical Specifications.

b. Describe the basis for the failure probability associated with the portable equipment.

c. Describe the basis of for the HEPs associated with use of the portable equipment.

d. Discuss how crediting the portable equipment is expected to impact 50.69 SSC categorization (e.g., any significant impacts on importance rankings) and whether crediting the CT MU pump in the PRA is considered a key assumption or source of uncertainty for the baseline PRA model or the 50.69 application.

APLA Topic for Discussion 3 - Disposition of open peer review Facts and Observations (F&Os) finding F&O 03-002 In order to support risk-informed categorization, PRA models need to reflect the as-built, as-operated plant. In of the LAR enclosure (Disposition and Resolution of Open Peer Review Findings and Self-

16 Assessments Open Items), the disposition for F&O 03-002 states, in part, that the impact is expected to be primarily related to enhanced documentation with a limited potential impact on risk results. It is the NRC staffs understanding that the additional documentation has yet to be completed for instrumentation power supplies.

The NRC staff would like to discuss the status of this implementation item and why it is expected there would be limited potential impacts on risk results with respect to the application (i.e., risk-informed categorization of SSCs in accordance with 10 CFR 50.69). The intent is to understand if impacts to the model are identified (e.g., an instrumentation power supply is not explicitly modeled) how such impacts would be resolved (e.g., as part of PRA maintenance or during categorization) and how SSC categorization could be potentially impacted (e.g., if the component is part of a system that is being categorized, or if its omission could affect risk results that affect categorization of other SSCs).

APLC Audit Questions APLC-AQ1 In Section 3.2.3, Seismic Hazards, of the LAR, the licensee states:

NextEra will follow the same alternative seismic approach in the 10 CFR 50.69 categorization process for Seabrook as that which was approved by the NRC staff for LaSalle [ML21082A422],

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

Seabrook site-specific seismic capacity information is described above herein.

However, the seismic capacity discussions presented in the LAR appear to be generic and not specific to the Seabrook site. The licensee is requested to identify and clarify which portions of the LAR contain Seabrook-specific seismic capacity information referenced in the above statement.

APLC-AQ2 Section 3.2.4, Other External Hazards, of the LAR states that external hazards were screened for applicability to Seabrook based on a plant-specific evaluation in accordance with Generic Letter 88-20 (ML20118C090) and updated to use the criteria in ASME/ANS RA-Sa-2009 (PRA standard). Section 6-3, Peer Review for Screening and Conservative Analysis, of the PRA standard provides guidelines for a peer review of external hazard screening assessments. However, the LAR does not appear to include a discussion of such a peer review.

The licensee is requested to provide information regarding any peer reviews conducted on the external hazards screening assessment for Seabrook or justify why no peer reviews were performed.

APLC-AQ3 Section 3.2.4, Other External Hazards, of the LAR states:

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

Table 3-2, Summary of Components Credited in External Hazards Screening, provides a high-level summary of doors credited in screening external hazards such as local intense precipitation (LIP) flooding, hurricane-induced probable maximum storm surge (PMSS) flooding, and extreme winds/tornadoes. However, these doors are not specifically identified in the LAR, but they are referenced in non-docketed documentation. As indicated in NEI 00-04 Figure 5-6, Other External Hazards, such components (doors) are likely to be categorized as HSS if credited in external hazard screening. The licensee is requested to provide a list of the

17 specific doors credited in screening the external hazards discussed in table 3-2 and Attachment 4, External Hazards Screening, of the LAR, including their physical locations and credited safety functions.