L-2024-061, NextEra Energy Seabrook, LLC, License Amendment Request - One Time Extension to Technical Specification 3.8.1.1.a, Allowed Outage Time with One Independent Circuit Between the Offsite Transmission Network and the Onsite Class 1E Distribut

NextEra Energy Seabrook, LLC, License Amendment Request - One Time Extension to Technical Specification 3.8.1.1.a, Allowed Outage Time with One Independent Circuit Between the Offsite Transmission Network and the Onsite Class 1E Distributio

ML24131A152
Person / Time
Site:	Seabrook
Issue date:	05/10/2024
From:	Rasmus P NextEra Energy Seabrook
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
L-2024-061
Download: ML24131A152 (1)

Text

L-2024-061 10 CFR 50.90 May 10, 2024 NextEra Energy Seabrook, LLC P.O. Box 300, Lafayette Road, Seabrook, NH 03874 Attn: Document Control Desk U.S. Nuclear Regulatory Commission Washington DC 20555-0001 RE: Seabrook Station Docket No. 50-443 Renewed Facility Operating License No. NPF-86 NextEra Energy Seabrook, LLC, License Amendment Request - One Time Extension to Technical Specification 3.8.1.1.a, Allowed Outage Time with One Independent Circuit Between the Offsite Transmission Network and the Onsite Class 1E Distribution System Out-Of-Service Pursuant to 10 CFR 50.90, NextEra Energy Seabrook, LLC (NextEra) hereby requests a one-time License Amendment to Renewed Facility Operating License NPF-86. The proposed change to Technical Specification (TS) 3.8.1.1.a, A.C. Sources - Operating, would provide a one-time allowance to change plant modes from Cold Shutdown (MODE 5) to Startup (MODE 2) while one independent circuit between the offsite transmission network and the onsite Class 1E Distribution System is out of service. Seabrook is requesting an additional 384 hours0.00444 days <br />0.107 hours <br />6.349206e-4 weeks <br />1.46112e-4 months <br /> to the 3.8.1.1.a 72-hour completion time for a total of 456 hrs.

During the next refueling outage, Seabrook is required to replace the main generator breaker in accordance with FERC rulings [Reference 1]. During this replacement, the preferred source of offsite power provided by the Unit Auxiliary Transformers (UATs) will be unavailable. Seabrook is requesting an allowance to perform outage startup activities in conjunction with breaker replacement. The one-time allowance would expire upon UAT restoration or 456 hours0.00528 days <br />0.127 hours <br />7.539683e-4 weeks <br />1.73508e-4 months <br />, whichever occurs earliest. If the 456-hour one-time allowance has not been fully utilized, by November 30, 2024, this License Amendment will expire.

The enclosure to this letter provides the description and assessment of the proposed change. to the enclosure provides the existing TS pages marked up to show the newly proposed changes. Attachment 2 provides the evaluation of risk impact for the proposed Allowed Outage Time (AOT), including a discussion of Probability Risk Assessment (PRA) scope, technical adequacy, modelling, and insights. NextEra requests approval of this License Amendment Request (LAR) by September 30, 2024.

NextEra has determined that the proposed license amendment does not involve a significant hazards consideration pursuant to 10 CFR 50.92(c), and that there are no significant environmental impacts associated with the change. The Seabrook Onsite Review Group (ORG) has reviewed the enclosed amendment request. In accordance with 10 CFR 50.91(b)(1), a copy of this LAR is being forwarded to the designee for the State of New Hampshire.

This letter contains no new or revised regulatory commitments.

Should you have any questions regarding this submission, please contact Mr. Kenneth Mack, Fleet Licensing Manager, at 561-904-3635.

Seabrook Station L-2024-061 Docket No. 50-443 Enclosure Page 1 of 8 Description and Assessment Seabrook Station NextEra Energy Seabrook, LLC., License Amendment Request - One Time Extension to Technical Specification 3.8.1.1.a, Allowed Outage Time with One Independent Circuit Between the Offsite Transmission Network and the Onsite Class 1E Distribution System Out-Of-Service 1.0

SUMMARY

DESCRIPTION................................................................................................................ 2 2.0 DETAILED DESCRIPTION................................................................................................................. 2 2.1 System Design and Operation................................................................................................... 2 2.2 Current Requirements / Description of the Proposed Changes................................................ 3 2.3 Reason for the Proposed Change............................................................................................. 3

3.0 TECHNICAL EVALUATION

............................................................................................................... 4

4.0 REGULATORY EVALUATION

........................................................................................................... 5 4.1 Applicable Regulatory Requirements........................................................................................ 5 5.0 PRECEDENTS.................................................................................................................................... 6 6.0 NO SIGNIFICANT HAZARDS CONSIDERATION............................................................................. 6 6.1 Conclusion................................................................................................................................. 7

7.0 ENVIRONMENTAL CONSIDERATION

.............................................................................................. 7

8.0 REFERENCES

.................................................................................................................................... 7 ATTACHMENTS

1.

Proposed Technical Specification Changes (mark-up)

2.

PRA Evaluation for Seabrook Station UAT LAR

Seabrook Station L-2024-061 Docket No. 50-443 Enclosure Page 2 of 8 1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, NextEra Energy Seabrook, LLC (NextEra) hereby requests a one-time License Amendment to Renewed Facility Operating License NPF-86. The proposed change to Technical Specification (TS) 3.8.1.1.a, A.C. Sources - Operating, would provide a one-time allowance to change plant modes from Refueling to Startup (MODE 2) while one independent circuit between the offsite transmission network and the onsite Class 1E Distribution System is out of service. Seabrook is requesting an additional 384 hours0.00444 days <br />0.107 hours <br />6.349206e-4 weeks <br />1.46112e-4 months <br /> to the TS 3.8.1.1.a 72-hour completion time for a total of 456 hrs.

NextEra requests approval of this LAR by September 30, 2024 to support the currently planned replacement of the main generator breaker and restoration of the offsite source to operable status, during refueling outage OR23 in the Fall of 2024.

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation The Seabrook Station Class 1E AC Electrical Power Distribution System AC sources consist of the preferred and alternate offsite power sources and the onsite standby power sources (Train A and Train B Emergency Diesel Generators (EDGs)). As required by 10 CFR 50, Appendix A, GDC 17, the design of the AC electrical power system provides independence and redundancy to ensure an available source of power to the Engineered Safety Feature (ESF) systems. The onsite Class 1E AC Distribution System is divided into redundant load groups (trains) so that loss of any one group does not prevent the minimum safety functions from being performed. Each train has connections to two offsite power sources and a single EDG. Offsite power is supplied to the unit switchyard from the transmission network by three 345kV transmission lines. From the switchyard, two electrically and physically separated circuits provide AC power, through the generator step-up transformer and/or step-down station auxiliary transformers, to the 4.16kV ESF buses. An offsite circuit consists of breakers, transformers, switches, interrupting devices, cabling and controls, required to transmit power from the offsite transmission network to the onsite Class 1E ESF buses.

Two qualified circuits between the offsite transmission network and the onsite Class 1E Electrical Power System, and separate and independent EDGs for each train ensure availability of the required power to shutdown the reactor and maintain it in a safe shutdown condition after an anticipated operational occurrence (AOO) or a postulated Design Basis Accident (DBA). One of the required, independent offsite AC sources consists of the circuit from an offsite transmission line through the Unit Auxiliary Transformers (UATs) to buses E5 and E6. Operability of this circuit requires that both UATs supply breakers be closed, energizing the emergency buses. The second required independent offsite AC source consists of the circuit from a separate offsite transmission line through the Reserve Auxiliary Transformers (RATs) to buses E5 and E6. For this circuit to be operable, each emergency bus RATs supply breaker must be either (1) closed, or (2) in standby with capability for automatic closure.

To meet the TS 3.8.1.1 requirement for two independent offsite sources, each emergency bus must be (1) energized from its UAT, and (2) have its RAT supply available via fast transfer capability.

Otherwise, the appropriate action of TS 3.8.1.1 must be entered. Examples of the application of the TS to various configurations of the offsite AC sources:

The offsite source via the UATs is operable when both buses E5 and E6 are powered via the UATs.

The offsite source via the RATs is operable when the RAT is in standby power supply for both buses E5 and E6, and both buses have operable auto transfer capability to the RAT.

Seabrook Station L-2024-061 Docket No. 50-443 Enclosure Page 3 of 8 With one emergency bus powered from a UAT with capability for auto transfer to the RAT, and the other bus powered from a RAT, the offsite source via the UAT is inoperable. The source through the UAT is not available via auto transfer on the bus that is energized from the RAT. (TS 3.8.1.1.a is applicable).

With the UAT powering one emergency bus without transfer capability to the RAT coincident with the RAT powering the other emergency bus (no auto transfer from the RAT to the UAT exists), both offsite sources are inoperable. (TS 3.8.1.1.e is applicable)

With both emergency buses powered from the RATs, the offsite source via the UATs is inoperable since no auto transfer capability to the UATs exists. (TS 3.8.1.1.a is applicable)

The AC sources are required to be operable in MODES 1, 2, 3 and 4 to ensure that:

a. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of AOOs or abnormal transients and

b. Adequate core cooling is provided, and containment operability and other vital functions are maintained in the event of a postulated DBA.

See Seabrook Technical Specifications and Bases, [Reference 2] and Updated Final Safety Analysis Report, [Reference 3].

2.2 Current Requirements / Description of the Proposed Changes In MODEs 1 through 4, Seabrook TS limiting condition for operation (LCO) 3.8.1.1 requires, as a minimum, two physically independent circuits between the offsite transmission network and the onsite Class 1E Distribution System along with two separate and independent diesel generators. With an offsite circuit inoperable, ACTION a.3. requires that the AC power sources be restored to operable status in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

is the marked-up TS. The proposed change is a one-time extension of TS 3.8.1.1, ACTION a.3, to restore the inoperable offsite circuit to operable status within 456 hours0.00528 days <br />0.127 hours <br />7.539683e-4 weeks <br />1.73508e-4 months <br />. The change would be in the form of a footnote to TS LCO 3.8.1.1.a.3 which would read as follows:

• A one-time Allowed Outage Time (AOT) extension for an inoperable offsite circuit allows 456 hours0.00528 days <br />0.127 hours <br />7.539683e-4 weeks <br />1.73508e-4 months <br /> to restore the inoperable Unit Auxiliary Transformers to OPERABLE status. Compensatory measures within NextEra Energy Seabrook, LLC letter L-2024-061, dated May 10, 2024, shall be implemented, and shall remain in effect during the extended AOT period. The one-time AOT extension shall expire upon completion of the maintenance to restore the Unit Auxiliary Transformers to OPERABLE status or by 456 hours0.00528 days <br />0.127 hours <br />7.539683e-4 weeks <br />1.73508e-4 months <br /> whichever occurs earliest. If the 456-hour one-time allowance has not been fully utilized, by November 30, 2024, this License Amendment will expire. MODE 1 operation is prohibited during the AOT.

Technical Specification SRs will demonstrate that LCOs are continually met in their associated MODEs of APPLICABILITY. Seabrook will enter MODE 2 ONLY to perform low power testing and will then reduce power to MODE 3, awaiting breaker replacement. While in the extended AOT, operation in MODE 1 will be prohibited.

2.3 Reason for the Proposed Change The New England Clean Energy Connect (NECEC) Project is a fully permitted transmission line project that will connect 1200MW from Quebec to the New England grid. It includes 145 miles of new high voltage direct current (HVDC) transmission line to a converter station in Lewiston, ME. ISO New England Inc. (the system operator) performed an impact study of the additional load on the existing system and determined that, in addition to upgrades to existing substations and transmission lines,

Seabrook Station L-2024-061 Docket No. 50-443 Enclosure Page 4 of 8 the Seabrook main generator breaker would be considered over-duty by 2510A. This would exceed the current, PKG2, air blast breakers interrupting capabilities. Therefore, a new, SF6, gas type breaker will be installed during the OR23 outage in the Fall of 2024.

3.0 TECHNICAL EVALUATION

NRC Regulatory Guide (RG) 1.93, Availability of Electrical Sources, provides specific guidance to licensee to address situations in which the number of available electric power sources are less than those required by LCOs. [Reference 4] RG 1.93 re-emphasizes General Design Criteria (GDC) 17, Electric Power Systems, which calls for: 1) on and offsite electrical power to permit functioning of systems, structures and components (SSCs) important to safety; 2) onsite electric power supplies, including batteries, with sufficient independence, redundancy and testability to perform their safety functions, assuming a single failure; 3) electric power from the transmission network, supplied by two physically independent circuits to minimize their simultaneous failure; and 4) provisions included to minimize the probability of losing electric power from any remaining electrical supplies. [Reference 5]

1) Seabrook TS 3.8.1.1.a BASES state, in part, that with one offsite circuit inoperable, the potential for a loss of offsite power is increased. However, In this condition the remaining operable offsite circuit and EDGs are adequate to supply electrical power to the onsite Class 1E Distribution System, supporting safety functions. This satisfies the first part of GDC 17.

2) Onsite electric power supplies, including batteries, will be in Operations guarded status during the time that one offsite circuit is out of service.

The following is an abbreviated list of guarded equipment:

Both Emergency Diesel Generators Supplemental Emergency Power System (SEPS)

Seabrooks 345kV Switchyard Breaker Enclosure Building 3A and 3B Reserve Auxiliary Transformers Seabrooks Relay Room By guarding of these areas, Seabrook will ensure that no unapproved work occurs in those areas that could threaten any of the electrical power supplies. Operations will perform a Senior Reactor Operator (SRO) walkdown of all guarded equipment once per shift. Risk significant configurations will be avoided in order to preserve safety margins and defense in depth strategies. [Reference 6]

The non-safety related SEPS is relied upon for defense-in-depth as a backup power source when an EDG is inoperable. It can supply power to the safety and non-safety related loads in the event of a total loss of offsite power and if one or both EDG(s) fail(s) to start and load. SEPs is capable of mitigating the dominant core damage sequences and can effect a safe shutdown of the unit.

[Reference 7]

Seabrook Station FLEX equipment can be used for non-FLEX purposes. Use of the equipment requires management approval; and its location must be tracked. Procedure EN1855.001, FLEX Equipment Use Control and System Administration, provides instruction and forms for deployment.

Should FLEX equipment become a supportive option during the OR23 breaker replacement, it is available for use. [Reference 8]

With these defense-in-depth measure in place, there is reasonable assurance that there is sufficient independence and redundancy of onsite power sources. This satisfies the intent of GDC 17, part two.

Seabrook Station L-2024-061 Docket No. 50-443 Enclosure Page 5 of 8

3) Seabrook has three ties to the New England 345kV transmission grid. The grid is a major network of inter-connecting lines covering all of New England with cross-ties at various points and radial lines feeding outlying areas. At least two of the independent transmission lines will be aligned to the RATs during breaker replacement. The Station will be in contact with the load dispatcher in order to stay apprised of any potential grid perturbations or of any switching that might be necessary. This satisfies the intent of GCD 17, part 3, to minimize any simultaneous failures.

4) Regulatory Guide 1.177 directly addresses the PRA-based risk metric requirements for one-time TS changes. [Reference 9] Section 2.4, Acceptance Guidelines for Technical Specification Changes, states that acceptable impact on plant risk is demonstrated by an incremental conditional core damage probability (ICCDP) of less than 1x10-6 and an incremental conditional large early release probability (ICLERP) of less than 1x10-7 OR an ICCDP of less than 1x10-5 and an ICLERP of less than 1x10-6 with effective compensatory measures implemented to reduce the sources of increased risk. (Tier 1)

Seabrook evaluated the PRA-based risk metric for the one-time 456-hour AOT. There was a minor impact on the PRA results but the increase is within NRC limits established by Regulatory Guide 1.177. Attachment 2 is the Probabilistic Risk Assessment.

Section 2.4 of Regulatory Guide 1.177 acceptance guidance includes having appropriate restrictions on risk-significant configurations associated with the change (Tier 2). The licensee must also have implemented a risk-informed plant configuration control program, including procedures to use, maintain, and control such a program (Tier 3).

Seabrook has configuration control policies and procedures as well as risk management policies and procedures. The capability of the remaining power sources will be managed within these programs. There is reasonable assurance that the probability of losing electric power from any remaining sources is minimalized, meeting the intent of GDC 17, part 4.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements 10 CFR 50.36(c)(2)(i) - When a limiting condition for operation is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TS until the condition can be met.

10 CFR 50.63 - Light-water-cooled nuclear power plants licensed to operate be able to withstand for a specified duration and recover from a Station Blackout (SBO).

10 CFR 50.65 -The licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities and that they must be sufficient to provide reasonable assurance that SSCs are capable of fulfilling their intended functions.

General Design Criteria (GDC) 17 - Emergency power for ESF systems shall be provided with adequate independence, redundancy, capacity and testability to permit safe operation and safe shutdown of the unit. considers acceptable when the number of available electric power sources are less than the number of sources required by the limiting conditions for operation (LCOs) for a facility.

Regulatory Guide (RG) 1.177, Revision 2 - Describes an approach that is acceptable to the staff of the U.S. Nuclear Regulatory Commission (NRC) for developing risk-informed applications for changes to completion times (CTs) and surveillance frequencies (SFs) of plant technical specifications (TS). This RG provides specific guidance for considering engineering issues and

Seabrook Station L-2024-061 Docket No. 50-443 Enclosure Page 6 of 8 using risk information to evaluate nuclear power plant TS changes to CTs and SFs [Reference 9]

Regulatory Guide (RG) 1.93, Revision 0 - Provides guidelines that the NRC staff considers acceptable when the number of available electric power sources are less than the number of sources required by TS LCOs. [Reference 4]

Generic Letter (GL) 2006 Required licensees to submit their protocols for maintaining offsite power systems and grid reliability. The Generic letter itself contains historical information related to the largest power outage in US history.

5.0 PRECEDENT 5.1 Watts Bar - Letter from NRC to Watts Bar Nuclear Plant, Unit 1, - Issuance of Amendment Regarding Alternating Current Sources, dated September 29, 2015 (ADAMS Accession Number ML5225A094) 5.2 Palo Verde - Letter from NRC to Palo Verde Nuclear Generating Station, Unit 3, Issuance of Amendment Regarding Revision to Technical Specification 3.8.1, AC Sources -

Operating (Emergency Conditions), dated December 23, 2016 (ADAMS Accession Number ML16358A676) 5.3 Quad Cities - Letter from NRC to Quad Cities Nuclear Power Station, Units 1 and 2, Issuance of Amendments No. 298 and 294 Regarding Increased Completion Time in Technical Specification 3.8.1.B.4 (Emergency Circumstances) dated December 17, 2023 (ML23349A162) 6.0 NO SIGNIFICANT HAZARDS CONSIDERATION NextEra has evaluated if a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment,"

as discussed below:

(1)

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

Response: No As Seabrook progresses through refueling outage OR23, the reactor will disperse weeks worth of decay heat. The Unit will transition from refueling to MODE 2. Low power testing will be performed, for approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, in MODE 2. Then, the Unit will transition back to MODE 3 until breaker replacement comes to completion. Power history will be negligible at this point and decay heat will be minimal. The proposed amendment will have a negligible effect on the probability or severity of an accident.

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

(2)

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

Response: No

Seabrook Station L-2024-061 Docket No. 50-443 Enclosure Page 7 of 8 The replacement of the generator breaker does not introduce new mechanisms could initiate core damage or degrade principal safety barriers. The Unit will continue to be operated within the limits of it licensing basis. And, with negligible power history and little decay heat, the Unit will continue to be operated within the limits of its licensing basis.

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

(3)

Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No Over the course of OR23, the Station will have negligible power history, minimal decay heat, backup power sources and guarded equipment. The risk associated with this one-time extended allowed outage time has been analyzed and found to be within the bounds of regulatory guidance.

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

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

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

7.0 ENVIRONMENTAL CONSIDERATION

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

8.0 REFERENCES

[1]

NextEra Energy Seabrook, LLC, 182 FERC ¶ 61,044, order addressing rehearing arguments, 183 FERC ¶ 61,196 (2023), appeal docketed (D.C. Cir. No. 23-1094)(re-hearing pending)

Filed 02/01/23

Seabrook Station L-2024-061 Docket No. 50-443 Enclosure Page 8 of 8

[2]

NextEra Energy Seabrook, LLC, Seabrook Technical Specifications, Unit No. 1, Docket No.

50-443, Appendix A to License No. NPF-86, Amendment 173, March 2024

[3]

Seabrook Updated Final Safety Analysis Report, Section 8, Electric Power Revision 22, November 2023

[4]

Regulatory Guide 1.93 Revision 0, Availability of Electric Power Sources, December 1974

[5]

General Design Criteria 17, Electric Power Systems, of Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50

[6]

Fleet Procedure, OP-AA-102-1003, Guarded Equipment, Revision 47, January 2024

[7]

NextEra Energy Seabrook, LLC, Seabrook Station Technical Requirement 31,Supplemental Emergency Power System Availability Requirements, Revision 170, November 2023

[8]

Seabrook Procedure, EN1855.001, FLEX Equipment Use Control and System Administration, Revision 12, February 2019

[9]

Regulatory Guide 1.177, Revision 2, An Approach for Plant-Specific, Risk-Informed Decision Making: Technical Specifications, dated January 2021

Seabrook Station L-2024-061 Docket Nos. 50-443 ATTACHMENT 1

Seabrook Station L-2024-061 Docket Nos. 50-443 ATTACHMENT 2 PRA Evaluation for Seabrook Station UAT LAR (25 pages follow)

PRA Evaluation for Seabrook Station UAT LAR Probabilistic Risk Assessment Group Document Number SBK-1FJR-24-009 Revision 1 May 2024 Prepared by:

Jadie Palenzuela Date Reviewed by:

Joseph Stringfellow Date Approved by:

Keith Vincent Date

SBK-1FJR-24-009 Revision 1 Page 2 of 25 Table of Contents Section Title Page 1.0 Purpose........................................................................................................................... 3 2.0 Methodology................................................................................................................... 3 3.0 PRA Quality.................................................................................................................... 4 4.0 Assumptions................................................................................................................... 5 5.0 Evaluation....................................................................................................................... 6 6.0 Results............................................................................................................................ 6 6.1 Quantification.............................................................................................................................. 6 6.2 Risk Insights................................................................................................................................ 7 7.0 External Events.............................................................................................................. 7 8.0 Summary and Recommendations................................................................................. 8 9.0 References...................................................................................................................... 8 10.0 Appendices..................................................................................................................... 9 Appendix A: Flag File Used for Quantification..................................................................... 10 Appendix B: Seabrook PRA Open Peer Review Issues....................................................... 11 Appendix C: Seabrook Fire PRA Open Peer Review Issues................................................ 13

SBK-1FJR-24-009 Revision 1 Page 3 of 25 1.0 Purpose This document provides a Probabilistic Risk Assessment of a proposed license amendment to modify the Seabrook Station Technical Specifications (TS) to provide a one-time allowance to increase plant modes from Cold Shutdown (MODE 5) to Startup (MODE 2) while one qualified Class 1E AC Electrical Power Distribution System AC source is out of service. The proposed license amendment would extend the completion time of Technical Specification (TS) Limiting Condition of Operation 3.8.1.1.a, A.C. Sources - Operating, from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 456 hours0.00528 days <br />0.127 hours <br />7.539683e-4 weeks <br />1.73508e-4 months <br />.

The Seabrook Station Class 1E AC Electrical Power Distribution System AC sources consist of the preferred offsite power source, or unit auxiliary transformer (UAT), the alternate offsite power source, or reserve auxiliary transformer (RAT), and the onsite standby power sources, or Emergency Diesel Generators (EDGs). The Seabrook Station also has a non-safety related supplemental emergency power supply (SEPS) relied upon as a backup power source when EDG is inoperable. The offsite circuits receive power from three 345,000 volt transmission lines which terminate in a common termination yard, which is then fed to the switchyard. The switchyard is arranged through circuit breakers and transformers to form the two qualified circuits. To meet TS 3.8.1.1 requirement for two independent offsite sources, each emergency bus must (1) be energized from its UAT, and (2) have its RAT supply available via fast transfer capability. This design provides independence and redundancy to ensure the Engineered Safety Feature (ESF) bus is capable of being supplied by the offsite circuits either through the UAT or RAT.

2.0 Methodology The current Internal Events & Fire PRA models for the Seabrook station were used for this evaluation. The current internal events PRA model, SBK23, was obtained from Ref. 3, and the current fire PRA model, SBK21, was obtained from Ref 5. Each model was quantified using the default truncation limits demonstrated for convergence for that respective model.

Regulatory Guide (RG) 1.177, An Approach for Plant-Specific, Risk Informed Decision-making:

Technical Specifications" [Ref. 1] describes acceptable methods for assessing the nature and impact of proposed T.S. changes, including AOT extensions, by considering engineering issues and applying risk insights.

Reg. Guide 1.177 directly addresses the PRA-based risk metric requirements for one-time TS changes, as reproduced below:

The following TS acceptance guidelines specific to one-time-only CT changes for evaluating the risk associated with the revised CT:

a. The licensee has demonstrated that the impact on plant risk from implementing the one-time-only TS CT change is acceptable (Tier 1):

(1) an ICCDP of less than 1x10-6 and an ICLERP of less than 1x10-7, or (2) an ICCDP of less than 1x10-5 and an ICLERP of less than 1x10-6 with effective compensatory measures implemented to reduce the sources of increased risk.*

b. The licensee has demonstrated that there are appropriate restrictions on dominant risk-significant configurations associated with the change (Tier2).

SBK-1FJR-24-009 Revision 1 Page 4 of 25

c. The licensee has implemented a risk-informed plant configuration control program, including procedures to use, maintain, and control such a program (Tier 3).

• For one-time-only CT changes, the ICCDP and ICLERP acceptance guidelines of 1x10-5 and 1x10-6, respectively, are established for compatibility with the ICDP and ILERP limits of Section 11 in NUMARC 93-01, which is applicable for voluntary maintenance activities requiring risk-management actions."

Based on the available quantitative guidelines for other risk-informed applications, it is judged that the criteria shown in Table 1 represents a reasonable set of acceptance guidelines. For the purposes of this evaluation, these guidelines demonstrate the risk impacts are acceptably low.

This, combined with effective compensatory measures (for planned TS entries) to maintain lower risk, ensure the TS change meets the intent of small risk increases consistent with the Commission's Safety Goal Policy Statement. This document addresses Tier 1 of the risk associated with the change; Tiers 2 and 3 are addressed in Section 3.4 of the LAR enclosure.

Table 1: Proposed Risk Acceptance Guidelines Risk Acceptance Guideline Basis ICCDP < 1E-6 ICCDP is an appropriate metric for assessing risk impacts of out of service equipment per RG 1.177. This guideline is specified in Section 2.4 of RG 1.177.

ICLERP < 1E-7 ICLERP is an appropriate metric for assessing risk impacts of out of service equipment per RG 1.177. This guideline is specified in Section 2.4 of RG 1.177 The ICCDP associated with each OOS configuration for a new CT is given by ICCDPUAT= (CDFUAT - CDFBASE) x CTNEW

[Eq. 1]

Where, CDFUAT= the annual average CDF calculated for configuration with equipment OOS CDFBASE = baseline annual average CDF with average unavailability for all equipment.

This is the CDF result of the baseline PRA CTNEW = the new extended CT (in units of years)

NOTE: ICCDP is a dimensionless probability and ICLERP is quantified similarly.

3.0 PRA Quality The ASME / ANS PRA Standard (ASME/ANS RA-Sa-2009), [Ref. 6], has technical elements, high-level requirements (HLRs), and detailed supporting requirements (SRs). NRC Regulatory Guide 1.200 Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities [Ref. 4], endorses the standard with minor clarifications.

The Seabrook PRA has undergone peer review against the ASME PRA Standard, Parts 1 (configuration control), 2 (internal events), and 3 (internal flood events).

SBK-1FJR-24-009 Revision 1 Page 5 of 25 Peer reviews have been conducted against internal event supporting requirements as follows:

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 review was done to PRA Standard ASME RA-Sa-2003.

In 2009, a focused peer review was performed for all elements of Part 3, Internal Flooding. This review was done to PRA Standard ASME/ANS RA-Sa-2009.

In 2012, a focused peer review was performed for the element LE. This review was done to PRA Standard ASME/ANS RA-Sa-2009.

In 2019, a focused peer review was performed on all elements upgraded by the conversion from Riskman to CAFTA. This review was done to PRA Standard ASME/ANS RA-Sa-2009.

Four self-assessments against the internal event SRs in the PRA standard were performed in 2005 (ASME RA-Sa-2003), 2007 (ASME RA-Sb-2005), 2010 (ASME/ANS RA-Sa-2009) and 2011 (ASME/ANS RA-Sa-2009). The first three self-assessments considered all internal events technical elements. The SA-2011 addressed only the open findings against specific SRs.

In October 2017, all resolved findings were reviewed to Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13, Close-out of Facts and Observations (F&Os) as accepted by the NRC staff in their May 3, 2017, memorandum (ML17079A427).

Following the findings closure review, the 2019 focused scope peer review identified additional findings. Appendix B provides a listing of the remaining open findings and the status of their resolution as well as an assessment of the impact on this evaluation. Overall, the open findings do not have an impact on the results and conclusions of this LAR.

The Seabrook Fire PRA model has undergone a peer review conducted in 2023. This review was done against the requirements of Part 4 of the PRA Standard ASME/ANS RA-Sa-2009.

A self-assessment against the requirements of Part 4 of the PRA Standard ASME/ANS RA-Sa-2009 was completed in 2023 and represents the most current status of the Seabrook Fire PRA model. This assessment reviewed the results of the 2023 peer review on the Seabrook Fire PRA model.

Appendix C includes the open findings as a result of the 2023 peer review on the Fire PRA model and the status of their resolution as well as an assessment of the impact on this evaluation. Overall, the open findings do not have an impact on the results and conclusions of this LAR.

4.0 Assumptions

1. Only the UATs are OOS within the 456-hour requested completion time. Offsite power would be supplied through the RATs with both EDGs operable and available. SEPS will also remain available.

2. Following terry turbine testing in MODE 3, emergency feedwater (EFW) will remain operable.

3. No compensatory measures are credited in this assessment.

SBK-1FJR-24-009 Revision 1 Page 6 of 25

4. Flooding risk is considered to be bounded by Internal Events. The North Yard area, which includes the UATs and RATS, was screened from internal flooding hazard risk per Ref. 2.

5.0 Evaluation When the UATs are OOS, 2 RATs provide back-up power to the stations emergency buses.

3A RAT provides backup power to Train A bus 5 emergency bus whereas 3B RAT provides backup power to Train B bus 6 emergency bus. These buses can also be powered from their respective EDGs. The high degree of redundancy for these emergency buses leads to the expectation that the risk significance of the UAT OOS will be low.

6.0 Results 6.1 Quantification The internal events PRA quantification was performed on the Ref. 3 internal events PRA model with a CDF truncation of 1E-12 and LERF truncation of 1E-13. A flag file, shown in Appendix A, was used to calculate the CDF and LERF for when the UATs are OOS Table 2 shows the results of the internal events PRA model quantification.

Table 2: Results for Internal Events Delta Risk Calculation Case CDF (/yr) Delta CDF (/yr) LERF (/yr) Delta LERF (/yr)

Baseline 2.45E-06 1.05E-07 UATs OOS 2.46E-06 1.00E-08 1.05E-07 0.00E-00 As noted in Assumption 4, the flooding risk increase is assumed to be equivalent to the Internal Events change in risk. This assumption is a conservative assumption as there are no unscreened flooding scenarios in the North Yard area so no flooding events are analyzed that would result in a loss of all transformers. The Internal Events PRA model does however still have these failures, such as a loss of offsite power, so the change in risk that is calculated would bound any change in risk from the unscreened internal flooding scenarios.

The fire PRA quantification was performed on the Ref. 5 model with a CDF truncation of 1E-12 and LERF truncation of 1E-13. The flag file, shown in Appendix A, was used to calculate the CDF and LERF for when the UATs are OOS. Table 3 shows the results of the fire PRA model quantification.

Table 3: Results for Fire Delta Risk Calculation Case CDF (/yr) Delta CDF (/yr) LERF (/yr) Delta LERF (/yr)

Baseline 2.95E-5 6.22E-7 UATs OOS 4.78E-5 1.83E-5 6.27E-7 4.64E-9

SBK-1FJR-24-009 Revision 1 Page 7 of 25 The contribution to change in risk for the UATs out of service was then summed to determine the total contribution from internal events, flood, and fire in Table 4, with the fire scenarios being the leading contributor to delta CDF.

Table 4: Conditional Risk Calculations for Internal Events, Flood, and Fire PRA Metric CTnew (yrs)

Delta CDF (LERF)

(/yr) (IE +IF + FR) 19 Day ICCDP (ICLERP)

CDF 0.052 (19 days) 1.83E-05 9.52E-07 LERF 0.052 (19 days) 4.64E-09 2.42E-10 6.2 Risk Insights The Fire PRA was reviewed to identify fire scenarios with the largest risk increase associated with the UAT unavailability. One scenario, TB_FC_1234_Z_Exclusion, contributes almost 100%

of the change in CDF (1.83E-05). The assumed failures for this scenario includes control power cables for the RAT power supply to the vital switchgear, and when damaged, result in a total loss of offsite power sources when the UATs are unavailable.

7.0 External Events The spectrum of external event challenges were evaluated to determine which external event hazards should be explicitly addressed as part of this risk assessment. Internal events, including internal flooding, and fire are quantitatively addressed as described in the previous sections.

The impact due to seismic, high winds, external floods, shutdown operation, and other hazard groups are addressed here.

Seismic Seabrook does not have a seismic PRA. However, the increase in risk during a seismic event is deemed negligible. The seismic risk is typically governed by the initiating event frequency, as a cliff-edge effect occurs in that all offsite power transformers would tend to be highly reliable up to a certain ground motion, and then all offsite power would fail past this point. Given the seismic event frequency is not changing, the change in seismic risk is less likely than the change in risk due to internal events.

Other External Events Seabrook does not explicitly evaluate any other external event hazards (e.g., external flooding) in the probabilistic modeling. Per Ref. 7, SBK-PRAE-15-010, these external hazards were screened for analysis per the ASME/ANS Standard. Given this condition does not increase the frequency of these external events hazards the change in risk is considered negligible.

SBK-1FJR-24-009 Revision 1 Page 8 of 25 8.0 Summary and Recommendations Extension of the UATs allowed outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 456 hours0.00528 days <br />0.127 hours <br />7.539683e-4 weeks <br />1.73508e-4 months <br /> has a minor impact on the PRA results and the increase is within NRC limits established by Reg. Guide 1.177.

Table 6: Conditional Risk Calculation PRA Metric Incremental Probability for 19 Day AOT ICCDP ICLERP Internal Events 5.21E-10 0.00E+00 Internal Floods 5.21E-10 0.00E+00 Fire 9.52E-07 2.42E-10 External Events Negligible Negligible Total 9.52E-07 2.42E-10 Acceptance Criteria 1.00E-06 1.00E-07 Based on the risk insights, it is recommended that compensatory measures be taken to ensure that the area noted in Ref. 17 as a PRA Risk Area in the Turbine Building is protected during the time that the UATs are unavailable.

9.0 References

1. USNRC, Regulatory Guide (RG) 1.177, An Approach for Plant-Specific, Risk Informed Decision-making: Technical Specifications, Revision 2, (ADAMS Accession No. ML20164A034).

2. SSPSS-2019 Seabrook Station Probabilistic Safety Study Revision 0, September 2019.

3. GDOC SBK-1FJR-23-006 Rev. 1, Seabrook Internal Events Model, August 2023.

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

5. GDOC SBK-BFJR-23-042 Rev. 1, Seabrook Unit 1 At-Power Fire PRA, Qualitative Screening, Quantification, and Uncertainty Analysis Notebook, December 2023.

6. ASME PRA Standard RA-Sa-2009, Addenda to ASME/ANS RA-S-2009 Standard for Level 1 / Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, February 2009.

7. GDOC SBK-PRAE-15-010 Rev. 0 Seabrook Station PRA Capability Assessment, November 2015.

8. GDOC SBK-BFJR-23-034 Rev. 1 Seabrook Unit 1 At-Power Fire PRA, Plant Partitioning Notebook, December 2023.

9. GDOC SBK-BFJR-23-036 Rev. 1 Seabrook Unit 1 At-Power Fire PRA, Cable Notebook, December 2023.

10. GDOC SBK-BFJR-23-037 Rev. 1 Seabrook Unit 1 At-Power Fire PRA, Fire Risk Model Notebook, December 2023.

SBK-1FJR-24-009 Revision 1 Page 9 of 25

11. GDOC SBK-BFJR-23-038 Rev. 1 Seabrook Unit 1 At-Power Fire PRA, Ignition Frequency Notebook December 2023.

12. GDOC SBK-BFJR-23-039 Rev. 1 Seabrook Unit 1 At-Power Fire PRA, Fire Scenario Selection Notebook, December 2023.

13. GDOC SBK-BFJR-23-040 Rev. 1 Seabrook Unit 1 At-Power Fire PRA, Main Control Room Analysis Notebook, December 2023.

14. GDOC SBK-BFJR-23-041 Rev. 1 Seabrook Unit 1 At-Power Fire PRA, Fire HRA Notebook, December 2023.

15. GDOC SBK-BFJR-23-042 Rev. 1 Seabrook Unit 1 At-Power Fire PRA, Qualitative Screening, Quantification, and Uncertainty Analysis Notebook, December 2023.

16. GDOC SBK-BFJR-23-044 Rev. 0 Seabrook Unit 1 At-Power Fire PRA, Seismic-Fire Interactions Notebook, December 2023.

17. FP 2.2 Revision 24, Control of Combustible Materials, August 2020.

10.0 Appendices Appendix A:

Flag File Used for Quantification Appendix B:

Seabrook PRA Open Peer Review Issues Appendix C:

Seabrook Fire PRA Open Peer Review Issues

SBK-1FJR-24-009 Revision 1 Page 10 of 25 Appendix A: Flag File Used for Quantification

ED-X-2-A and ED-X-2-B OUT OF SERVICE (OOS) FLAG and ED-X-3-A and ED-X-3-B IN SERVICE FLAG

Take the UATs OOS XX.EDX2A.OOS EQU.T XX.EDX2B.OOS EQU.T

Align the RAT to be in service WWW-RATN EQU.F XX.EDX3A.RUN EQU.T XX.EDX3B.RUN EQU.T WWW-RATA EQU.F WWW-RATB EQU.F

SBK-1FJR-24-009 Revision 1 Page 11 of 25 Appendix B: Seabrook PRA Open Peer Review Issues Supporting Requirement Issue Evaluation DA-D4-01 DA-E2-02 The Seabrook PRA uses all operating experience when performing the Bayesian update. The use of all operating experience in the Bayesian update can provide non-conservative results for component failure probabilities.

For example, if a component has been replaced, previous operating experience is no longer applicable for that component.

The Bayesian reasonableness check does not discuss any criteria for when there are 0 failures in the plant-specific experience. For these cases, none of the checks will pass the specified criteria.

Section 1.0 of the Seabrook PRA Notebook 13, Data Analysis, was revised (to Rev. 2) to provide additional information. No impact on application.

DA-E1-01 The following documentation issues were identified:

1) Table 13.6-1 of the Data Analysis shows the Bayesian validation of the Seabrook type codes. It is noted that the Bayesian update equations used for Beta distributions are incorrect. The equation used to update the beta parameter of the beta distribution should be B_prior +

n_exposures - n_failures. The current equation used is B_prior + n_exposures. Note that the current equation used is not consistent with the CAFTA Bayesian update tool.

2) Section 13.6.2 of the Data Analysis discusses three conditions for checking the reasonableness of the Bayesian update. In the description of the conditions it should be stated 5th percentile and less than the 95th percentile of the generic/posterior distribution.

3) Section 13.6.2 states that the parameters of interest in the reasonableness check are the: mean values, 5th percentile value, and 95th percentile value. Table 13.6-1 does not provide the mean values.

The Data Notebook was revised to address this documentation-only finding.

No impact on application.

DA-E2-01 The following documentation issues were identified:

1) A review of the CAFTA database shows that there are 6 common cause groups making use of the MGL method: BUSFX, BUSFL, LINES, LINES.YR, LINESMNT, and LINESMNT.YR. A search of the System Analysis notebook states that for BUS56FX Note that MGL CCF parameters are used in the 2019 update because the 2015 update to NUREG/CR-5497 did not have information on switchgear CCF failure data. This statement does not provide a reference to the data source used, and the data notebook does not provide this information either.

2) There is no discussion regarding the selection of staggered or non-staggered testing schemes and the use of these calculation methods for the CCF groups.

The Data Notebook was revised to address this documentation-only issue.

No impact on application.

QU-C2-01 / HR-G7-01 1)Self-assessment identifies limitations with manpower requirements and there still appear to be gaps with HRAC specific inputs for manpower. Additionally, execution locations are also not identified for all actions.

2)Not being able to reproduce results. Recreating the dependency analysis using the same cutsets that were used and creating a combination event recovery rule file resulted in 860 combinations versus the documented 505 combinations in the Section 11 HR document Section 11.8.1.3.

3)Manual combination and dependency overrides lacked sufficient justification for assigned dependency levels.

For example, combination of HH.OFL0CW.FL and HH.OFL1CW.FA, the current justification taken is for The Human Reliability Analysis (HRA) was revised in support of the 2022 Seabrook model update. The three items identified were all addressed in Revision 2 of SBK-1FJR-19-041, Seabrook Human Reliability Analysis - Internal Events. Execution locations were added to the HRA Calculator, the method for performing dependency analysis was documented and reproducible, and dependency overrides were justified. No impact on application.

SBK-1FJR-24-009 Revision 1 Page 12 of 25 Supporting Requirement Issue Evaluation larger timing separation between actions, however, the override taken is equivalent to intervening success. This isnt sufficient justification for the override taken.

AS-C1-01 / DA-E1-02 /

HR-I1-01 / SY-C1-01 Section 3.0 of Section 11, Human Actions Analysis, discusses methodology and references PRA-106 PRA Model Guidelines, Section 106E Methodology for Human Reliability Analysis. PRA-106 is the modeling information for RISKMAN. No discussion could be found for dependency analysis methodology in the conversion report.

Similar issue was found to exist in Systems Analysis, Data Analysis, HRA, and Accident Sequence.

This issue is documentation only. The methods used for performing HRA and dependency analysis are documented in Rev. 2 of SBK-1FJR-19-041. No impact on application.

HR-E4-01 There are instances where the information from Appendix 11.1A does not match the HRAC. See example below.

HH.OHSB1.FA Tcog 5 minutes versus Appendix 11.A1 Tcog of 20-30 minutes.

Also, Operator interview Insights in HRAC for HH.OALT1.FL dont seem to match the interview documentation. This appears to be a systemic problem as there were other instances found.

This issue was resolved with Rev. 2 of SBK-1FJR-19-041. Operator interviews were re-performed in support of the report revision and updated in the HRA Calculator. No impact on application.

QU-B9-01 Logic flags have not been set to TRUE or FALSE for all flags prior to the generation of cutsets. The current methodology sets logic flags to TRUE in the recovery rules which occurs after the generation of cutsets.

Additional cutsets have been generated in the final results that should not exist as they are non-minimal.

This issue was resolved by flag file changes in the 2022 model update. No non-minimal cutsets were identified in the model update. No impact on application.

QU-E3-01 / QU-A3-01 SOKC is not accounted for in some type codes that use identical data sets. One example is for the type codes NICB1C and NICB1O. Both of these type codes use the same data set, but since they are different type codes UNCERT does not take the same sample for both distributions. This appears to be a common approach when the generic data doesnt delineate between the different failure modes of a component.

This issue was resolved with the 2022 model update. Type codes that shared data sets were revised to ensure functionality with UNCERT. No impact on application.

QU-F2-01

1) The FTREX.ini file was not documented. This is necessary to quantify the model and is significant because the default method is not used.

2) The criteria establishing convergence is based on

<=5% change when compared to the next decade. The example in the standard uses a <5% final change. The final change is interpreted as calculating the percent change at the current truncation level with respect to the previous decade truncation level not the next. The criteria used is adequate, but there is no documentation of definition used to establish convergence.

3)There is no discussion of the top basic events and why they make logical sense. A general statement that notes that basic events importances were reviewed to ensure they make logical sense is not sufficient evidence for the actual review taking place.

4) There is no documentation of how the circular logic is broken. A demonstration was performed that identified a couple examples of where in the model circular logic was broken. This identification and modeling technique needs to be documented.

This issue is documentation only. No impact on application.

QU-D6-01 Component importance measures were not identified.

The supporting requirement specifically requires the identification of significant SSCs.

This issue was resolved with the 2022 model update. Component importance measures were identified in the model update report. No impact on application.

SBK-1FJR-24-009 Revision 1 Page 13 of 25 Appendix C: Seabrook Fire PRA Open Peer Review Issues The open findings as a result of the 2023 peer review on the Fire PRA are included in the table below. An evaluation of the findings has been completed as part of a 2023 self-assessment and the resolutions have been applied to the current Fire PRA model. Therefore, the open findings do not have an impact on the results and conclusions of this LAR.

Supporting Requirement Issue Evaluation PP-A1 Description Areas outside the protected area were improperly included as fire compartments Basis PAUs such as the warehouses outside the protected area have no potential to adversely affect any equipment or cable item to be credited in the Fire PRA. This has the potential to dilute the transient weighting factors.

NAH-FIRE-TR-RM-000001 (SBK-BFJR-23-034) and associated GPAB attachment was revised to remove the PAUs outside the protected area that have no potential adverse effect such as MISCEXPA.

This issue was resolved in Revision 1 of SBK-BFR-23-034; therefore, no impact on this application.

ES-A2 / ES-B4 /

PRM-B9 Description Review of power supplies and interlocks for inclusion in ES.

Basis Section 5.7 of the ES notebook indicates that any additional power supplies, interlocks, and other dependencies that are identified in the CS notebook are either considered sub-components on the main component, and thus the associated cables are included with the main component, or are added to the FPRA equipment list. A complete listing of the interlocks by equipment functional state does not appear to be included in the CS notebook, and no systematic review of power supplies and interlocks appears to have been made and documented in the ES notebook. It is unclear if this review has been completed.

The FSS Database (SBK-BFJR-23-035)

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

This issue was resolved in Revision 1 of SBK-BFR-23-035; therefore, no impact on this application.

ES-B1 Description Cables should be mapped to basic events RCPCV456A[B].RS (PORV A[B] fails to reseat).

Basis In tbl_import_Equip_List, basic events RCPCV456A[B].RS (PORV A[B] fails to reseat) are screened out as Screening A with the Note type code NIVR3C - PORV, ASDV - fail to reseat; no electrical dependency. Since the PORVs can fail to close due to a spurious hot short, cables should be mapped to these events. Note that cables are mapped to basic events RCPCV456A[B].TO, but these events are used in different places in the CAFTA logic.

The Fire PRA Equipment List tbl_import_Equip_List was revised to model the reseat basic events. The cable selection was updated to map cables to the reseat basic events identified for modeling.

This issue was resolved in Revision 1 of SBK-BFR-23-035; therefore, no impact on this application.

ES-D1 Description Corrections to Table tbl_import_Equip_List.

Basis Basic event CSHCV123.TO (CS Valve CS-HCV-123 transfers open) is dispositioned in tbl_import_Equip_List as I (always failed), but the box for Modeled is checked. Note that all other entries in tbl_import_Equip_List dispositioned as I (always failed) are included in tbl_fq_FRANX_Zone_to_Raceway Zone AlwaysFailed. Also, in tbl_import_Equip_List, basic event SWAF192.PL.YR (CT Pump Room Intake FILTER F192 fails due to plugging) is screened as disposition A, but the box for Screened is not checked. Confirm this is correct.

The fire PRA Equipment List tbl_import_Equip_List was reviewed.

tbl_import_Equip_List was revised to change the disposition of basic events to N/A if screened and confirm all basic events have one box checked (screened, modeled, always failed).

Confirmed CSHCV123.TO is modeled with N/A for Screen_Basis. Confirmed Screened was checked for SWAF192.PL.YR.

This issue was resolved in Revision 1 of SBK-BFR-23-035; therefore, no impact on this application.

SBK-1FJR-24-009 Revision 1 Page 14 of 25 Supporting Requirement Issue Evaluation CS-B1 Description Coordination for 120V AC or 125V DC buses/panels has been assumed and not reviewed.

Basis Section 4.3.1.1.3 and Attachment 3 of the CS/CF notebook analyzes and confirms coordination for 4160V and 480V buses and MCCs. However the notebook does not discuss or confirm that coordination exists for voltages less than 480V (e.g., 120V AC or 125V DC buses/panels), and rather assumes that a correct coordination exists. After speaking with the Peer Review support team, additional coordination calculations were presented that confirm coordination for the remaining buses.

These additional calculations were only listed as References in the CF/CS notebook, with no discussion or point to the additional references. Experience indicates that this level of voltage historically has issues with proper coordination.

NAH-FIRE-TR-RM-000003 (SBK-BFJR-23-036) was revised consistent with the possible resolution. The coordination of 120V AC and 125 V DC panels/buses has been analyzed and included in tables 4&5 of Attachment 3. Text in Sections 4.3.1.1.4 & 4.4.1 has been modified.

CS-C1 Description Review of interlocks and other dependencies not documented.

Basis The methodology in Section 4.3.1.1.3 of the CS/CF notebook (NAH-FIRE-TR-RM-000003) does not provide sufficient detail or guidance to identify interlocks and other dependencies during the cable selection process. The section solely discusses power supplies, which are documented in. After discussing with the peer review support team, it seems that cables supporting interlocks were included in the cable selection, however the process and results for this review are not documented in the CS/CF or ES notebooks.

The Cable Notebook (SBK-BFJR 036) was revised to include documentation regarding the identification of interlocks and dependencies during cable selection process.

This issue was resolved in Revision 1 of SBK-BFR-23-036; therefore, no impact on this application.

CS-C4 Description Coordination analysis for 480V MCCs lack sufficient technical basis.

Basis of the CS/CF notebook evaluates the coordination of select 480V MCCs (E511, E512, E513, E514, E515, E611, E612, E613, E614, E615) by analyzing only the largest load (motor and non-motor). However, it could not be confirmed that this could be applied to all loads on these MCCs. Additionally, Section 4.3.1.1.3 includes a statement that discusses evaluation and assumption of short circuit current at the end of cables connected to MCC-E511. Sufficient information or a calculation could not be found to support this statement.

NAH-FIRE-TR-RM-000003 (SBK-BFJR-23-036) and associated attachments were revised consistent with the possible resolution for MCCs. Text in Section 4.3.1.1.3 has been modified to justify the resolution of this F&O.

This issue was resolved in Revision 1 of SBK-BFR-23-035; therefore, no impact on this application.

PRM-B2 / HRA-C1 Description Unpopulated HRA Calculator Tables that Impact Dependency Analysis Basis This Finding is consistent with previous Internal Events F&O HR-6-1. There are 26 combinations that are currently assessed by HRA Calculator [SBK_HRA_DAF.daf] as 2 - Complete Dependency, which implies Sufficient Staffing and Simultaneous Timing. Since no overrides were assigned, it is not clear if or how well these combinations were reviewed for appropriateness. But the quantification was done with no Crew Composition table filled out in HRA Calculator file SBK-21 Rev 2_Fire DIT-004, which is the basis for the dependency analysis staffing evaluation. While the 2-CD dependency evaluation could be considered conservative since it leads to a high(er)

Joint HEP, it is not technically correct in evaluating SR HR-G7

- availability of resources. This is a potential concern since to the Fire HRA Notebook, NAH-FIRE-TR-RM-The Internal Events HRA Calculator was revised and delivered through DIT-005.

This was incorporated into the Fire HRA Notebook (SBK-BFJR-23-041).

Confirmed execution PSFs included located such as A/B Train Switchgear Rooms SEPS Switchgear Room for HH.SEPE5XFR.FA.

This issue was resolved in Revision 1 of SBK-BFR-23-041; therefore, no impact on this application.

SBK-1FJR-24-009 Revision 1 Page 15 of 25 Supporting Requirement Issue Evaluation 000008, the Fire PRA Operator Interview Summary, Section 3.0, states that The Fire Brigade is made up of the FB leader and four of the five NSOs. (NSOs are Aux Operators in the field, so if there are two or more local actions happening simultaneously in a short timeframe, there is a question of whether there are sufficient personnel available.) In addition, the Execution Location information in HRA Calculator is not always populated, although it has been identified and documented for ex-MCR actions in Fire HRA Notebook Table 4.3-1. While this information omission does mainly involve Level 2 actions, there are local actions credited in the Fire PRA (e.g., HH.OFCRL3.REC and HH.SEPE5XFR.FA) that are impacted and may not be evaluated properly in the dependency analysis.

PRM-B11 Description Some non-credited operator actions are inadvertently being credited in the Fire PRA quantification.

Basis Section 5.5.1 (Table 5.5-1) of the ES notebook identifies operator actions that are screened from the FPRA, with the basis that the accident sequences to which they apply are not caused by fire-induced failures. Specifically, the Level 2 HFEs which were screened, which are not ANDed with their specific initiator (%SGTR LOOPX, %ALMFW, %AMFW, %ALOSP, %ASLOC, etc). One of these events (HH.OCI1A.FA) first appears in cutset #35, and has a FV of 0.0025 The Fire PRA updated the flag file (SBK-Master-Flag - No Flood.flg) to set to actions not credited in the Fire PRA to TRUE for quantification.

This issue was resolved in Revision 1 of SBK-BFR-23-042; therefore, no impact on this application.

Description Set Internal Events flag to FALSE in Fire PRA quantification.

Basis Internal Events PRA quantification appears to require flag event FL-INTERNAL set to TRUE or 1.0 (and initiator %FIRE set to FALSE or 0.0). For Fire PRA it may be advisable to set FL-INTERNAL to FALSE to ensure internal events HFEs are not inadvertently credited.

The Fire PRA updated the flag file (SBK-Master-Flag - No Flood.flg) to set FL-INTERNAL to FALSE for quantification.

This issue was resolved in Revision 1 of SBK-BFR-23-042; therefore, no impact on this application.

PRM-B11 / HRA-B2 Description Possible detrimental actions in the fire procedures should be considered in the Fire PRA.

Basis HRA Notebook Table 4.3-6 lists several prompt disablement actions that are taken by the operators (dependent on the fire location) that may be detrimental to the safe shutdown of the plant. For example, if the PORV block valves are closed (based on OS1200.00 Step 5) and then all steam generator cooling is failed due to the fire or random failures, bleed and feed cooling could not be credited unless the PORV block valves are re-opened, which would require a new operator action.

The Fire PRA was updated to review the fire procedures such as OS1200.00, OS1200.01 and OS1200.02 and model any prompt actions which may be detrimental. Appendix G was added to the PRM Notebook (SBK-BFJR-23-037) for the required model updates. was added to the Fire HRA Notebook (SBK-BFJR-23-041) for review of detrimental impact from prompt disablement actions.

This issue was resolved in Revision 1 of SBK-BFR-23-037 and SBK-BFJR 041; therefore, no impact on this application.

PRM-C1 Description Disposition of Internal Events F&O 5-1 is incomplete.

Basis According to PRM-B2, the peer review exceptions and deficiencies for the Internal Events PRA are required to be dispositioned, it is also required that the disposition does not adversely affect the development of the Fire PRA plant response model. For F&O 5-1, NAH-FIRE-TR-RM-000004, The disposition of Internal Events F&O 5-1 was revised in NAH-FIRE-TR-RM-000004 (SBK-BFJR-23-037) to confirm the updates were completed in Internal Events prior to Fire PRA.

This issue was resolved in Revision 1 of SBK-BFR-23-037; therefore, no impact on this application.

SBK-1FJR-24-009 Revision 1 Page 16 of 25 Supporting Requirement Issue Evaluation Appendix E, Table E-1 states The 2019 data update covers the period of July 1, 2013 through August 31, 2018, not all of Seabrooks operating experience. Considering the very small fraction of components in the database replaced during this time, the impact on failure rates is negligible. With respect to F&O 5-1, the fraction of components that were replaced over this time period is irrelevant if a Fire PRA component is risk significant due to its random failure probability. Updating the random failure probability basic event based on plant data may be inappropriate if the component has been replaced. An inappropriate change in random failure probability of a risk significant basic event could result in a change in fire risk. The Seabrook team provided additional information during the review that documented closure of F&O 5-1. However, NAH-FIRE-TR-RM-000004 has not been updated to reflect this closure.

FSS-A1 Description There are no fire scenarios developed for Fire Compartment FCMISCEXPA.

Basis This ID is identified as a fire compartment in the PP Notebook as the group of buildings/structures outside of the Protected Area, however, there is no fire ignition frequency or fire scenario assigned to this fire compartment.

This is inter-related with 06-012. The PP Notebook was revised to remove fire compartment FCMISCEXPA therefore ignition frequency was not assigned to a location removed from GPAB. All documentation for FCMISCEXPA was removed from Ignition Frequency (SBK-BFJR-23-038), FSS (SBK-BFJR-23-039) and Quantification Notebook (SBK-BFJR-23-042).

This issue was resolved in Revision 1 of SBK-BFR-23-038, SBK-BFJR-23-039 and SBK-BFJR-23-042; therefore, no impact on this application.

FSS-B2 Description The MCR Analysis does not specifically model the MCB scenarios for control room abandonment due to habitability and the FSS Spreadsheet screens MCBs based on a review of Control Room fire events in the Fire events Database (FEDB) reveals that none of the Control Room fires affected items much beyond the point of ignition.

Basis The potential for MCB scenarios to result in main control room abandonment is not included in the Seabrook Fire PRA.

The MCR Notebook (SBK-BFJR-23-040) and FSS (SBK-BFJR-23-039) attachments (for scenario development) were revised to assess the MCB scenarios for habitability criteria. The MCB HRR was documented and justification provided to assign abandonment probability. Section 4.3.8.5 of the FSS Notebook documents the application of MCB fire progression event tree. CFAST previously included the appropriate HRR for open cabinets and FSS documentation reflects the assumed MCR abandonment for propagation beyond single panel.

This issue was resolved in Revision 1 of SBK-BFR-23-040 and SBK-BFJR 039; therefore, no impact on this application.

FSS-C1 Description The MCR Analysis does not consider the potential for a transient fire within the MCB horseshoe impacting the open, exposed MCB panels. The MCR Analysis does not consider the potential for propagation and/or damage to MCB panels across the horsehoe corridor to the open, exposed cables in the MCB panels.

Basis The MCB and transient scenarios could result in damage to additional MCB back/front panels due to the open configuration.

The FSS Notebook (SBK-BFJR-23-039) was revised to include documentation about transients within the MCB which include single and multiple panel damage within the MCB. FSS attachments (for scenario development) were revised to model single panel and multiple panel damage within the MCB based on the layout of the MCB and transient ZOI. Scenarios are identified by CB_FC_3A_A-T-MCB* or CB_FC_3A_A-TFWC-MCB* with the node, panel, or panels impacted.

SBK-1FJR-24-009 Revision 1 Page 17 of 25 Supporting Requirement Issue Evaluation Description The fire modeling does not postulate fire spread to adjacent cabinets for electrical panels (non-switchgear and MCCs).

Basis The fire modeling currently models all electrical panels as single fire scenarios with no propagation to the adjacent section of the cabinet or adjacent cabinets.

Section 4.3.5.8 was added to FSS Notebook (SBK-BFJR-23-039).

The target sets assigned to electrical cabinet modeled with detailed fire modeling were reviewed and revised as needed to either justify why propagation cannot occur or revise the target sets to reflect propagation to adjacent electrical sections.

This issue was resolved in Revision 1 of SBK-BFR-23-039; therefore, no impact on this application.

FSS-C8 Description There is no specific documentation provided that confirms that the fire rated conduit is not subject to mechanical damage or direct flame impingement from a high-hazard ignition source.

Basis This SR requires confirmation that the credited fire wrap will not be subject to either mechanical damage or direct flame impingement from a high-hazard ignition source.

Section 4.3.5.7 of NAH-FIRE-TR-RM-000006 (SBK-BFJR-23-039) was revised to include documentation similar to what was provided during responses during peer review for the impact of high hazards and HEAF on fire wrapped conduits.

This issue was resolved in Revision 1 of SBK-BFR-23-039; therefore, no impact on this application.

FSS-D3 / FSS-G6

/ FQ-E1 Description The top CDF and LERF contributors contain known conservatisms that are not addressed. The Quantification Notebook includes a sensitivity analysis which indicates that the fire PRA is sensitive to MCA scenarios. MCA scenarios DG_FC_3AB_Z-CB_FC_2B_A and DG_FC_3AB_ZDG_FC_2A_A are included in the top 20 CDF scenarios. Based on the current results, several fire scenarios among the top contributors are utilizing conservative modeling approaches and including larger than realistic target sets. The Seabrook Fire PRA model has not refined the detailed fire scenarios to remove the modeling conservatisms in several of these top scenarios Basis The two SWGR and Cable spreading rooms are at the top of the list for risk relevant fire compartments. For LERF in particular, the cable spreading room contributes 95% of the risk. For CDF these three fire compartments make up 46% of the risk. There are known conservatisms in the fire modeling within these locations and based on this, there is potential masking of significant contributors taking place. MCA scenarios are risk significant based on the Quantification Notebook sensitivity analysis and some MCA scenarios are included in the top 20 CDF scenarios. Risk significant fire scenarios such as the Cable Spreading Room transient fires are failing a large amount of targets in the compartment. Walkdown of this compartment indicates fire modeling refinements are available to reduce these impacts.

The Fire PRA was updated to address F&Os and additional refinements were performed as necessary to remove conservatisms. The risk insights represent the as-built as-operated plant condition. MCA scenarios have been refined and additional detailed fire modeling was completed.

This issue was resolved in Revision 1 of SBK-BFR-23-042; therefore, no impact on this application.

FSS-D4 Description An initial ambient temperature of 25°C was utilized in the fire modeling calculations for several fire scenarios. This ambient temperature does not appear to be appropriate for areas that are not temperature controlled such as the PAB, Turbine Building, Service Water Pumphouse, DG Buildings, etc.

Basis The use of a higher initial ambient temperature could impact the required HRR, damage time, and therefore, the severity factors and non-suppression probabilities.

NAH-FIRE-TR-RM-000006 (SBK-BFJR-23-039) was revised to document the review of ambient temperatures for the plant. The FSS spreadsheet (and downstream calculations for HGL formation) were revised to incorporate the revised ambient temperatures in DATA worksheet of Attachment 2 to FSS Notebook.

SBK-1FJR-24-009 Revision 1 Page 18 of 25 Supporting Requirement Issue Evaluation This issue was resolved in Revision 1 of SBK-BFR-23-039; therefore, no impact on this application.

FSS-D7 Description The FSS Notebook does not include the review of the the automatic detection systems for outlier behavior relative to system unavailability to justify the use of the generic estimate for detection unavailability.

Basis This review is needed to confirm that the generic unavailability values assigned are bounding for the automatic detection systems.

Section 4.3.3.2.1 was revised to add a statement about assumed unavailability based on FP 3.1 and review of maintenance rule function failures.

This issue was resolved in Revision 1 of SBK-BFR-23-039; therefore, no impact on this application.

FSS-D8 Description Specific features of physical analysis unit and fire scenario under analysis (e.g., pocketing effects, blockages that might impact plume behaviors or the visibility of the fire to detection and suppression systems, and suppression system coverage) are not included in the detailed fire modeling to ensure effectiveness of suppression and detection with respect to specific fire scenarios. The FSS Spreadsheet does not assess the time to detector activation for fire scenarios, instead an automatic detection system probability is assigned if installed but assumes a 15-minute detection.

Basis There is a potential to credit detector or sprinkler actuation that will not be effective for an individual scenario based on the specific configuration. Automatic detection is credited in scenarios involving lower HRRs without assessment of the detection actuation timing and effectiveness for the specific scenario and fire size.

Section 4.3.3.2.3 was added to NAH-FIRE-TR-RM-000006 (SBK-BFJR 039) to document the review of fire detection and suppression system effectiveness on a scenario specific basis. The results of the detailed calculation are shown in the DATA worksheet of the FSS Spreadsheet to track the review and conclusions.

Smaller HRR were confirmed to not activate the detection or activate at later time than larger HRR or fires in the same configuration.

This issue was resolved in Revision 1 of SBK-BFR-23-039; therefore, no impact on this application.

FSS-F2 Description There is no criteria provided or justified in the exposed structural steel analysis for structural collapse in terms of fire generated conditions.

Basis There is no criteria to confirm which fire scenarios will result in exposed structural steel damage and structural collapse.

NAH-FIRE-TR-RM-000006 (SBK-BFJR-23-039) Section 4.3.5.5 documentation was revised to document and justify the criteria for structural steel collapse due to fire-generated conditions.

This issue was resolved in Revision 1 of SBK-BFR-23-039; therefore, no impact on this application.

FSS-G6 Description MCA Scenario CP-FC-1-0-TB_FC_1234_Z is included in the FSS Spreadsheet, however, CP-FC-1-0 is not a separate fire compartment and the Plant Partitioning Notebook lists CP-FC-1-0 as a fire zone within TB_FC_1234_Z. Also, there is a typo in the compartment ID for scenarios in the Service Water Pumphouse in the FSS Database.

Basis CP-FC-1-0-TB_FC_1234_Z MCA scenario is between a fire zone located within a Fire Compartment and is not a compartment boundary. Compartment IDs SW_FC_1C_A and SW_FC_1C-A are used for the same compartment and there is an issue with the two separate IDs in quantifying the MCA scenarios.

NAH-FIRE-TR-RM-000006 spreadsheet (SBK-BFJR-23-039) and database MCA was reviewed and revised to ensure the MCA scenarios are based on fire compartments.

NAH-FIRE-TR-RM-000006 spreadsheet (SBK-BFJR-23-039) and database MCA were reviewed and revised as needed to ensure there are no typos and the fire compartments are accurate and mapping targets in FRANX (see SW_FC_1C_A vs. SW_FC_1CA).

This issue was resolved in Revision 1 of SBK-BFR-23-039; therefore, no impact on this application.

FSS-H2 Description The FSS Notebook includes discussions of the basis for the Thermoset damage criteria and lack of self-ignited cable fires as the cable type being qualified to IEEE-383.

Basis NAH-FIRE-TR-RM-000006 ( SBK-BFJR-23-039) Section 4.3.5.1 was revised to remove statements about IEEE-383 and replaced with the UFSAR reference and information on thermoset cables

SBK-1FJR-24-009 Revision 1 Page 19 of 25 Supporting Requirement Issue Evaluation There are inconsistencies in the documentation that consider IEEE-383 qualified is equivalent to Thermoset cable.

included in other parts of the Fire PRA documentation.

This issue was resolved in Revision 1 of SBK-BFR-23-039; therefore, no impact on this application.

FSS-H8 Description A large number of MCA scenarios have a 1E-15 fire frequency in FRANX with no basis for the frequency.

Basis These MCA scenarios are screening due to the lack of hot gas layer and apply a default frequency value of 1E-15 in FRANX.

NAH-FIRE-TR-RM-000006 ( SBK-BFJR-23-039) was revised to document the MCA screening and HGL formation results in new table within Section 4.3.4 Multi-Compartment Analysis.

This issue was resolved in Revision 1 of SBK-BFR-23-039; therefore, no impact on this application.

IGN-A7 Description There are multiple cases where the counting of ignition sources is inconsistent with the documented methodology as well as consistency across all fire compartments.

Basis Bin 16.2 (iso phase bus duct) counting is not consistent with the documented approach as discussed in Table 4-2. The document suggests that all frequency is allocated to the Turbine Building only. However, the PRA Model team has identified that the allocated count and documentation is incorrect and needs to be updated. During the walkdown it was identified that there are discrepancies in the counting of small cabinets, specifically wall mounted cabinets. In general, these types of cabinets were included in the component count as found in Attachment 3 of NAH-FIRE-TR-RM-000005. However, as examples, there were some wall mounted cabinets that are located in the EDG room (DG_FC_2B_A) that are similar in design and size that were screened from the count with no justification provided why these were screened. During the walkdown it was identified that the remote shutdown panel count was a 1. However, based on the physical size of the cabinet compared to other similar cabinet sizes, the count is applied inconsistently. The remote shutdown panel includes 3 doors, which based on other counted panels, would be considered a count of 3. Bin 35, TGO, is another example of this discrepancy. Table 4-2 suggests that this Bin was equally split among the three locations that contain the TGO; Turbine Oil Tank Room, Lube Oil Reservoir Room, and the Main Turbine Building. However, Attachment 3 shows that the entire frequency associated with Bin 35 has been applied to the Turbine Building.

NAH-FIRE-TR-RM-000005 (SBK-BFJR-23-038) was revised to implement consistent counting of ignition sources and provide justification or document any assumptions performed during counting to ensure the frequency can be maintained and updated with the provided documentation through the expansion of information included in. The count of remote shutdown panels (MM-CP-108 A / B) were revised in Attachment 3.

Consistency in ignition source counting is demonstrated with documentation in which was populated based on prior walkdown data and supporting plant information. Table 4-2 was revised to document the approach for Bin 16.2 and Bin 35.

This issue was resolved in Revision 1 of SBK-BFR-23-038; therefore, no impact on this application.

IGN-A9 Description The transient weighting factors are inconsistently applied across the fire compartments within plant locations.

Basis Application of transient weighting factors used a 3 for almost all locations. Ductbanks used a 0 for occupancy but still had hot work, maintenance, and storage. During the walkdown it was noted that the non-essential SWGR room NES_FC_1A_Z had a fair amount of storage (compared to similar type fire compartments) which appeared to be there during normal operation. This fire compartment is characterized as average for storage, this seems to be low given what was found. The SW pump house area (SW_FC_1E_Z) is also characterized as average for storage. However, during the walkdown, it was identified that there is a significant amount of storage located in this area. The Cable Spreading Room, which also has a 3, had verylittle to no storage In the area during the walkdown.

Compared to other areas, this rating may be considered high.

NAH-FIRE-TR-RM-000005 (SBK-BFJR-23-038) Attachment 3 weighting factors within Transient Survey Results worksheet were revised to reflect the plant experience. Ductbanks occupancy was revised to 0.3, NES_FC_1A_Z storage was revised to 10 based on similar compartments. SW_FC_1E_Z storage was revised to 10 based on FLEX equipment storage. Cable Spreading (CB_FC_2A_A) was revised to 1 with Notes added for justification.

The notes within "Transient Survey Result" document the clarification when revisions were made beyond what was communicated through survey results.

SBK-1FJR-24-009 Revision 1 Page 20 of 25 Supporting Requirement Issue Evaluation This issue was resolved in Revision 1 of SBK-BFR-23-038; therefore, no impact on this application.

IGN-B3 Description The following inconsistencies in the ignition source apportioning documentation were observed:

1) In Attachment 3 of NAH-FIRE-TR-RM-000005, there is a table that identifies those sources that have been screened or removed from the analysis. There are some sources that do not provide a justification for screening or removal.

2) There are some ignition source bins discussed in Table 4-2 of NAH-FIRE-TR-RM-000005 that are incorrectly stating how the actual counting is being applied.

3) Plant walkdowns were performed initially in 2015.

Subsequent walkdowns were performed in 2022 and 2023 but only in select fire compartments.

4) The non-PRA Junction Boxes are all included in a pseudo, fire compartment"UN" Basis

1) Any ignition source that has been identified and subsequently screened or removed (and identified as such) should contain a basis for its removal. For example, ignition source 0-SY-CP-TMBR-SWMP in fire compartment TB_FC_1C_Z is listed as being screened with no basis.

2) Table 4-2 Bin 8 states that there are two EDGs at Seabrook.

However, a review of Attachment 3 indicated that there are actually 4 sources identified as part of Bin 8. The Fire PRA model team has confirmed that there are in fact 4 EDGs at the plant. Table 4-2 Bin 16.1 states that the segmented bus ducts are counted by the number of transition points. However, a review of Attachment 3 and as observed by the walkdowns, the count of the bus duct is based on number of bus duct runs for a given fire compartment. The Fire PRA model team has confirmed that the intended counting methodology is to follow what was done. However, the documentation does not accurately reflect this approach.

3) During this period of time, a number of plant changes have taken place where pieces of equipment were added and/or removed. These updates could have taken place in fire compartments that were not walkdown in 2022 and 2023. The Fire PRA model team acknowledged that the utility did perform a review of engineering changes for applicability to the Fire PRA. This was reviewed and transmitted to the Fire PRA model team in reference DIT 002-PRA.

4) The ignition frequency associated with this fire compartment

1) Attachment 3 of the IGN Notebook (SBK-BFJR-23-038) was revised to remove the Screened Ignition Sources. was expanded to include explicit notes for inclusion/exclusion of all electrical nodes in EDISON.

2) Table 4-2 was revised and Section 4.3.2.1 was added to provide additional counting clarifications.

3) NAH-FIRE-TR-RM-000005 (SBK-BFJR-23-038) was revised to explicitly document the conclusions from the review of the plant design changes for impact on the Fire PRA in Attachment 6 (newly added).

4) NAH-FIRE-TR-RM-000005 (SBK-BFJR-23-038) Attachment 3 was revised to map JB-REMAINING to FCMISC.

Section 4.3.7.6 of NAH-FIRE-TR-RM-000006 (SBK-BFJR-23-039) was revised to clarify the modeling approach for the non-PRA junction boxes and approach for maintenance of the scenarios.

This issue was resolved in Revision 1 of SBK-BFR-23-039 and SBK-BFJR 038; therefore, no impact on this application.

IGN-B4 Description Not all Potentially Challenging fires identified were dispositioned to confirm that the event can be screened from the Bayesian update process.

Basis There are 3 events identified in Attachment 1 that have no basis for excluding from the Bayesian update.

NAH-FIRE-TR-RM-000005 (SBK-BFJR-23-038) Attachment 1 was revised to document the Bayesian Update Disposition in a new column. No additional Bayesian updates were performed based on the added dispositions.

This issue was resolved in Revision 1 of SBK-BFR-23-038; therefore, no impact on this application.

IGN-B5 Description The mean and statistical representation of the uncertainty intervals is documented for most ignition source bins. However, some Bins are missing data.

Basis NAH-FIRE-TR-RM-000005 (SBK-BFJR-23-038) was revised to remove the uncertainty terms and provide a reference to worksheet IGN contained within Attachment 2 of the FSS Notebook.

SBK-1FJR-24-009 Revision 1 Page 21 of 25 Supporting Requirement Issue Evaluation The uncertainty terms for Bin 15 are missing. On an ignition source level, Bin 24 results in an"erro" since there is no 5th and 95th percentile. However, Attachment 2 of the FSS calculation contains all the necessary data.

This issue was resolved in Revision 1 of SBK-BFR-23-038; therefore, no impact on this application.

CF-A1 Description Discrepancy between CF notebook and FSS database.

Basis Upon performing a sample review of conditional probabilities in of the CS/CF notebook (NAH-FIRE-TR-RM-000003) and the FSS database (NAH-FIRE-TR-RM-000006 ), a data discrepancy was found for cables attached to basic event RCP1CBKR.FTO (Equipment 1-RC-BKR-1-C). Certain cables for this BE were listed in table tbl_import_CFMLA with conditional probabilities, however the cables were not identified in Attachment 2 of the CS/CF notebook. After consulting with the peer review support team, the missing cables were determined to be not required for this BE. Review of Attachment 1 of the CS/CF notebook and table tbl_import_Cables_Selection confirmed that the missing cables are not required for the subject BE. Similar discrepancies were found on cables for other RC-BKR equipment.

The FSS database"tbl_import_CFML" was reviewed and revised to remove cables not required per Attachments 1 and 2 of NAH-FIRE-TR-RM-000003 (SBK-BFJR-23-036).

This issue was resolved in Revision 1 of SBK-BFR-23-039; therefore, no impact on this application.

HRA-B2 Description Application of Disablement Action beyond Fire Procedure Direction Basis Fire PRA staff indicated [per response to Question EC-02] that the one disablement action modeled at this time is HH.OISPV1FR.FA, Close PORV Block Valve to Isolate Stuck-Open PORV (SLOCA)-- FIRE. This HFE is being applied to fire areas beyond those specified in Table 4.3-6 of the Fire HRA Notebook (which cites relevant fire procedure steps) Fire PRA staff clarified that HH.OISPV1FR.FA is directed by E-0 Step 7 and uses the timing based on E-0 instead of prompt disablement action timing.

The Fire HRA Notebook was revised to add the suggested paragraph as stated in the Possible Resolution.

This issue was resolved in Revision 1 of SBK-BFR-23-041; therefore, no impact on this application.

HRA-B3 Description Procedure Path for HFEs Credited for Fire Basis The primary Fire AOP OS1200.00, RESPONSE TO FIRE OR FIRE ALARM ACTUATION, specifically states up front that E-0 should not be entered if a fire induced transient causes a reactor trip. There are two fire HFEs HH.FWP37BMSFR.FA, and HH.OISPV1FR.FA that have E-0 cited as their primary Procedure in HRA Calculator. Responses to Questions EC-09 and DK-08 stated that Fire AOP OS12000.00 precautions that E-0 should not be entered until Step 9 of the procedure is processed. Based on the feedback from Seabrook operations where there is no transition out, the EOPs and Fire AOP procedures are implemented in parallel.

HH.FWP37BMSFR.FA and HH.OISPV1FR.FA were reviewed in the HRA Calculator to clarify the basis from operator insights for the modeled procedure path and timing. This is documented in Procedure and Training Notes within the HRA Calculator and documentation was added to Section 4.3.3.3 of the Fire HRA Notebook (SBK-BFJR-23-041).

This issue was resolved in Revision 1 of SBK-BFR-23-041; therefore, no impact on this application.

HRA-C1 Description Lack of Fire HEP Consistency and Reasonableness Check along with Risk-Significant HFEs Quantified with Screening Values Basis SR HR-G6 from Part 2 of the PRA Standard requires a consistency and reasonableness check of the HEPs relative to each other and given the scenario and context. This was not done. Attachment 5, Basic Event Importance Measures, of the FQ Notebook (NAH-FIRE-TR-RQ-000001, Rev. 0) was reviewed against the Fire HFE list from the Fire HRA Calculator The Fire HRA Notebook (SBK-BFJR 041) was revised to add Attachment 4 listing the Fire HFEs and HEP along with discussion on the reasonableness including a comparison with the other Fire HFEs.

This issue was resolved in Revision 1 of SBK-BFR-23-041; therefore, no impact on this application.

SBK-1FJR-24-009 Revision 1 Page 22 of 25 Supporting Requirement Issue Evaluation file SBK-21 Rev 2_Fire DIT-004. There are four HFEs and three combinations including HFEs that have FV>0.005 for CDF and are based on screening value HEPs:

HH.ODEP4.FA HH.OSIG2FR.FA HH.OSIG3FR.FA HH.SWCTSURVFR.FA COMBINATION_104 COMBINATION_135 COMBINATION_148 HRA-D1 Description MCR Abandonment HFEs not credited in Fire PRA Model Basis Seven (7) fire HFEs are identified in the Fire HRA Notebook with a basic event name suffix of.REC and involve recovery actions to restore specific systems at the RSS due to a CR fire in a specific MCR panel. However, Fire PRA staff stated (response to Question EC-02) that"The actions developed at the RSS are discussed in the HRA notebook (.REC actions).

Note that the current Seabrook MCR abandonment scenario does not credit any of the recovery actions at the RSS after MCR abandonment" Section 5.3.2 of PRM Notebook (SBK-BFJR-23-037) was revised to document the model changes to credit operator actions for MCR abandonment. The Fire HRA Notebook (SBK-BFJR-23-041) includes the operator actions.

This issue was resolved in Revision 1 of SBK-BFR-23-041; therefore, no impact on this application.

HRA-E1 Description Documentation Issues within HRA Calculator Basis This Finding is consistent with F&O 20-9 from the November 2016 Fire PRA Peer Review documented in PWROG-16042-P Rev. 0, which identified an issue to Adjust the HFE timings used for two scenarios: 1) Spurious actuation of containment spray impacts on operator action for sump recirculation, 2)

Loss of Main Feedwater impacts on Bleed & Feed actions. A new fire HFE with specific timing was developed for the first case, but notes in the HRA Calculator indicated for the Bleed &

Feed case that while an approximation was used, a new MAAP run was needed. Fire PRA staff stated [Question EC-08] that during the development of the updated IE HRA analysis, the approximation for an additional 2 minutes to transition from 30% wide range to 20% wide range was determined to be appropriate. However, the documentation in the timing analysis for the HFEs in question was not fully updated to reflect this change either in the IE or the Fire HRA.

This issue is documentation only. No impact on application.

SF-A2 Description NAH-FIRE-TR-RM-000009 does not specifically address diversion of suppressants from areas where they might be needed for those fire suppression systems associated with a common suppressant supply.

Basis SF-A2 requires assessment of the potential for diversion of suppressants from areas where they might be needed for those fire suppression systems associated with a common suppressant supply.

This F&O is related to Seismic Fire Interaction which is a qualitative assessment, therefore this F&O does not impact this application.

SF-A3 Description Other than a reference to the IPEEE walkdowns that were performed in 1992 and a statement that the associated fire PRA walkdowns also did not identify any concerns for seismic induced common cause failure of multiple fire suppression This F&O is related to Seismic Fire Interaction which is a qualitative assessment, therefore this F&O does not impact this application.

SBK-1FJR-24-009 Revision 1 Page 23 of 25 Supporting Requirement Issue Evaluation systems, no qualitative assessment for the potential was provided.

Basis SR SF-A3 requires the assessment of the potential for common-cause failure of multiple fire suppression systems due to the seismically induced failure of supporting systems such as fire pumps, fire water storage tanks, yard mains, gaseous suppression storage tanks, or building standpipes.

SF-A5 Description No discussion of the fire brigade training is described.

Basis Part (a) of the SR requires that the extent to which training has prepared firefighting personnel to respond to potential fire alarms.

This F&O is related to Seismic Fire Interaction which is a qualitative assessment, therefore this F&O does not impact this application.

FQ-A1 Description To support the fire quantification all required cables must be included in the FRANX data.

Basis There are a number of cables (E3C-F36/1 and F36-P89 as examples) that are listed as"require" in Attachment 1 of NAHFIRE-TR-RM-000003. However, these cables are not found in the FSS Database (Attacment 1 of NAH-FIRE-TR-RM-000006). Not having these cables in the FSS Database could lead to an underestimation of the risk.

The FSS database"tbl_import_Cable_Selectio" was reviewed and revised to ensure consistent with Attachment 1 of NAH-FIRE-TR-RM-000003 (SBK-BFJR 036). ErrorCheck_RequiredCables added to FSS database to confirm with future updates.

This issue was resolved in Revision 1 of SBK-BFR-23-039; therefore, no impact on this application.

FQ-E1 Description The top CDF and LERF contributors contain known conservatisms that are not addressed.

The Quantification Notebook includes a sensitivity analysis which indicates that the fire PRA is sensitive to MCA scenarios.

MCA scenarios DG_FC_3AB_Z-CB_FC_2B_A and DG_FC_3AB_ZDG_FC_2A_A are included in the top 20 CDF scenarios.

Based on the current results, several fire scenarios among the top contributors are utilizing conservative modeling approaches and including larger than realistic target sets. The Seabrook Fire PRA model has not refined the detailed fire scenarios to remove the modeling conservatisms in several of these top scenarios Basis The two SWGR and Cable spreading rooms are at the top of the list for risk relevant fire compartments. For LERF in particular, the cable spreading room contributes 95% of the risk. For CDF these three fire compartments make up 46% of the risk. There are known conservatisms in the fire modeling within these locations and based on this, there is potential masking of significant contributors taking place.

MCA scenarios are risk significant based on the Quantification Notebook sensitivity analysis and some MCA scenarios are included in the top 20 CDF scenarios.

Risk significant fire scenarios such as the Cable Spreading Room transient fires are failing a large amount of targets in the compartment. Walkdown of this compartment indicates fire modeling refinements are available to reduce these impacts.

The Fire PRA was updated to address F&Os and additional refinements were performed as necessary to remove conservatisms. The risk insights represent the as-built as-operated plant condition. MCA scenarios have been refined and additional detailed fire modeling was completed.

SBK-1FJR-24-009 Revision 1 Page 24 of 25 Supporting Requirement Issue Evaluation FQ-F1 Description The non significant cutset reviews do not sufficiently document a review of each cutset.

Basis and 7 of NAH-FIRE-TR-RQ-000001 document the review of randomly selected cutsets. These tables include all cutsets not included in the top 100 and only a small number of cutsets actually include a review. The process used to select these reviewed cutsets is not clear as there are very few cutsets actually documented. In discussion with the PRA model team, it was acknowledged that many of the insights were the same or similar. However, it is not clear from a reviewers standpoint that this is the case. This was reviewed as part of Part 2 SR QU-D5.

NAH-FIRE-TR-RQ-000001 (SBK-BFJR-23-042) was revised to enhance the documentation for the nonsignificant cutset review completed in Section 5.3.2.

This issue was resolved in Revision 1 of SBK-BFR-23-042; therefore, no impact on this application.

Description of NAH-FIRE-TR-RQ-000001 provides a table of the basic event importances for both CDF and LERF.

Basis The tables in Attachment 5 do not provide sufficient basis for the relative importance of events. There is no documentation supporting the various types of importances relative to operator action, CCF events, joint HFEs (combinations), components as examples. This was reviewed as part of Part 2 SR QU-F3.

This is tracked in the open Item section of NAH-FIRE-TR-RQ-000001 (SBK-BFJR-23-042). There is no planned update to add details to Attachment 5 of the Importances in this revision.

There is no impact to this application since this is documentation only.

UNC-A1 Description The CAFTA.rr database provided to the peer review team prior to the peer review team does not contain parameter distribution types or the error factors for the scenarios. Another database was provided during the review that did contain some error factors for scenarios (not all) but did not contain the distribution types. There are inconsistencies between the model and the documentation with respect to some of the uncertainty distributions.

Basis Without the distribution type in the.rr database, the error factor will be ignored by UNCERT code so the uncertainty in the ignition frequency is not evaluated. In the documentation of the human failure events, it is stated that the distribution type is lognormal. However, in the.rr database, lognormal and beta distributions are used.

The Fire PRA added a CAFTA.rr database to the NAH-FIRE-TR-RQ-000001 (SBK-BFJR-23-042 ) which is specific for UNCERT with the distribution types and uncertainty parameters populated. The documentation was reviewed and revised as necessary to align with the modeled distribution types.

This issue was resolved in Revision 1 of SBK-BFR-23-042; therefore, no impact on this application.

UNC-A2 Description The state of knowledge correlation (SOKC) is not specifically addressed in the Seabrook documentation.

Basis Back referenced SR QU-E3 requires taking into account the state-of-knowledge correlation. The Seabrook model does take into account the SOKC through the use of type codes to address the parametric uncertainties, but the documentation does not address it specifically.

Section 5.5.19.3 was revised in the Quantification Notebook, SBK-BJFR 042, to include statements about how SOKC is addressed.

This issue was resolved in Revision 1 of SBK-BFR-23-042; therefore, no impact on this application.

SBK-1FJR-24-009 Revision 1 Page 25 of 25 Supporting Requirement Issue Evaluation Description Case 1 Sensitivity Study: Always Failed includes two results: 1) case where cutsets contain HEPCOMB2 and 2) case where HEPCOMB2 cutsets are excluded. HEPCOMB2 is a combination of any 2 HFEs that were not included in the dependency analysis, but appeared in the sensitivity case cutsets. For this sensitivity case, the dependency analysis should have been expanded to include the unanalyzed combinations.

Basis The cutsets containing HEPCOMB2 have combinations of events that have not been analyzed. The use of this method masked the sensitivity study performed for always failed components such that its importance cannot be assessed.

The documentation for Case 1 sensitivity case was enhanced to describe the insights gained by the existing analysis.

This issue was resolved in Revision 1 of SBK-BFR-23-042; therefore, no impact on this application.

Description CAFTA.rr database contains legacy items that are no longer used.

Basis The peer team believed that the documentation was inconsistent from the CAFTA.rr database because some hot short probabilities contained uncertainy distributions types (lognormal) that did not match the distribution type discussed in the documentation (beta).

The.rr database was purged prior to the final updates to remove unused basic events and gates.

This issue was resolved in Revision 1 of SBK-BFR-23-042; therefore, no impact on this application.