ML21260A074

Regulatory Audit Plan in Support of License Amendment Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times

ML21260A074
Person / Time
Site:	Brunswick
Issue date:	09/22/2021
From:	Andrew Hon Plant Licensing Branch II
To:	Krakuszeski J Duke Energy Progress
Hon A
References
EPID L-2021-LLA-0060
Download: ML21260A074 (26)

Text

September 22, 2021 Mr. John A. Krakuszeski Site Vice President Brunswick Steam Electric Plant Duke Energy Progress, LLC 8470 River Rd., SE (M/C BNP001)

Southport, NC 28461

SUBJECT:

BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 - REGULATORY AUDIT PLAN IN SUPPORT OF LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT RISK-INFORMED COMPLETION TIMES (EPID L-2021-LLA-0060)

Dear Mr. Krakuszeski:

By letter dated April 1, 2021, Duke Energy Progress, LLC (Duke Energy or the licensee) requested to amend license No. DPR-71 and DPR-62 for Brunswick Steam Electric Plant, Units 1 and 2, respectively, to adopt Technical Specifications Task Force (TSTF) Traveler 505 (TSTF-505), Provide Risk-informed Extended Completion Times, RITSTF Initiative 4b, to permit the use of risk-informed technical specification completion times for certain actions required when limiting conditions for operation are not met.

The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed Duke Energys application and determined that a regulatory audit would assist in the timely completion of the review. The NRC staff will conduct a regulatory audit to support its review in accordance with the enclosed audit plan. A regulatory audit is a planned activity that includes the examination and evaluation of primarily non-docketed information.

The NRC staff will conduct the audit to increase its understanding of the application and identify information that will require docketing to support the NRC staffs regulatory findings. The NRC staff will conduct the audit virtually during September 27 - 30, 2021.

J. Krakuszeski If you have any questions, please contact me at (301) 415-8480 or by e-mail at Andrew.Hon@nrc.gov.

Sincerely,

/RA/

Andrew Hon, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-325 and 50-324

Enclosure:

Audit Plan cc: Listserv

Enclosure REGULATORY AUDIT PLAN BY THE OFFICE OF NUCLEAR REACTOR REGULATION IN SUPPORT OF LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT RISK-INFORMED COMPLETION TIMES DUKE ENERGY PROGRESS, LLC BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 DOCKET NOS. 50-325 AND 50-324 1.0

Background

By letter dated April 1, 2021 (Reference 1), Duke Energy Progress, LLC (Duke Energy or the licensee) submitted a License Amendment Request (LAR) to amend license No. DPR-71 and DPR-62 for the Brunswick Steam Electric Plant (BNP), Units 1 and 2, respectively. The proposed amendment would modify the plants Technical Specifications (TSs) to permit the use of risk-informed completion times (RICTs) in accordance with Technical Specification Task Force (TSTF) Traveler TSTF-505, Provide Risk-informed Extended Completion Times, RITSTF [Risk-Informed Technical Specification Task Force] Initiative 4b (Reference 2). These RICTs would apply to certain actions required when limiting conditions for operation are not met.

The staff from the Nuclear Regulatory Commissions (NRC) Office of Nuclear Reactor Regulation (NRR) has initiated its review of the LAR in accordance with NRR Office Instruction LIC-101, License Amendment Review Procedures (Reference 3).

2.0 Regulatory Audit Basis A regulatory audit is a planned license-or regulation-related activity that includes the examination and evaluation of information of primarily non-docketed information. An audit is conducted to gain understanding, to verify information, and to identify information that will require docketing to support the basis of a licensing or regulatory decision. An audit will assist the NRC staff in efficiently conducting its review and gaining insights to the licensees processes and procedures. Information that the NRC staff relies upon to make the safety determination must be submitted on the docket. This audit will be conducted in accordance with NRR Office Instruction LIC-111, Regulatory Audits, Revision 1, dated October 2019 (Reference 4), with exceptions noted within this audit plan.

The NRC staff will perform the audit to support its evaluation of whether the licensees request can be approved per Title 10 of the Code of Federal Regulations (10 CFR), Section 50.90, Application for amendment of license, construction permit, or early site permit. The staffs review will be informed by Standard Review Plan Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis (Reference 5). The audit will assist the NRC staff with understanding the licensees proposed program to implement RICTs for certain TSs.

3.0 Regulatory Audit Scope and Methodology NRCs objectives of the audit are the following:

Learn how the licensees proposed program implements TSTF-505 and conforms to NRC-endorsed guidance in Nuclear Energy Institute (NEI) report NEI 06-09, Revision 0-A, Risk-Informed Technical Specification Initiative 4b, Risk-Managed Technical Specification Guidelines (Reference 6).

Gain a better understanding of the detailed calculations, analyses, and bases underlying the LAR and confirm the staffs understanding of the LAR.

Gain a better understanding of plant design features and their implications for the LAR.

Identify any information needed to enable the staffs evaluation of the technical acceptability of the probabilistic risk assessment (PRA) used for this application.

Identify any information needed to enable the staffs evaluation of whether the proposed changes challenge design-basis functions or adversely affect the capability or capacity of plant equipment to perform design-basis functions.

Identify questions and requests that may become formal requests for additional information (RAIs) per NRR Office Instruction LIC-115, Processing Requests for Additional Information (Reference 7).

The NRC staff will audit the PRA methods that the licensee would use to determine the risk impact from which the revised completion times would be obtained, including the licensee assessments of internal events (including internal flooding) and fire PRAs. The NRC will also audit the licensees quantification of risk from significant external events, whether the licensee uses PRA or bounding methods, and the licensees evaluation of defense-in-depth.

4.0 Information and Other Material Necessary for The Audit The NRC staff will request information and interviews throughout the audit period. The NRC staff will use an audit items list to identify the information (e.g., methodology, process information, and calculations) to be audited and the subjects of requested interviews and meetings. The NRC staff will provide the final audit items list as an enclosure to the audit summary report, which will be publicly available. The attachments to this audit plan include the initial audit items list and questions that the staff intends to address prior to or during the audit.

Throughout the audit, the NRC staff will supplement this list with audit questions and audit-related requests so that the licensee can better prepare for audit discussions with NRC staff. Any information accessed through the licensees portal will not be held or retained in any way by NRC staff. The NRC will use the audit items list to support the periodic audit meetings with the licensee, which the NRC staff will schedule as needed. The NRC staff requests the licensee to have the requested audit information listed in the audit items list to be readily available and accessible for the NRC staffs review via a Web-based portal.

5.0 Team Assignments The audit team will consist of the following NRC staff from NRR.

William (Bill) Jessup, PRA Licensing Branch A (APLA), Lead Auditor Andrew (Andy) Hon, Plant Licensing Branch II-2 (LPL2-2), Project Manager Mihaela Biro, APLA Bernard (Bernie) Grenier, PRA Licensing Branch B (APLB)

Milton Valentin, PRA Licensing Branch C (APLC)

De (Wesley) Wu, APLC Edmund Kleeh, Electrical Engineering Branch (EEEB)

Khoi Nguyen, EEEB Joseph Ashcraft, Instrumentation and Controls Branch B (EICB)

Ming Li, EICB Kaihwa (Robert) Hsu, Mechanical Engineering and Inservice Testing Branch (EMIB)

Gurjendra Bedi, EMIB Brian Lee, Containment and Plant Systems Branch (SCPB)

David Nold, SCPB Summer Sun, Nuclear Systems Performance Branch (SNSB)

Khadijah West, Technical Specifications Branch (STSB)

Mark Wilk, Pacific Northwest National Laboratory (PNNL)

Garill Coles, PNNL Steve Short, PNNL 6.0 Logistics To support the schedule established when the NRC staff accepted the LAR for technical review, audit activities will be performed remotely and virtually using Microsoft Teams, teleconference, and any Web-based portals or meeting spaces created by the licensee. NRC information requests and communications with licensee staff will be coordinated through the NRCs licensing project manager.

The audit will start at 11:00 a.m. with an entrance meeting between the NRC and the licensee.

The NRCs licensing project manager will inform the licensee of the entrance and exit meeting dates when they are established. The NRC intends to establish periodic (e.g., biweekly) meetings on mutually agreeable dates and times (to be determined) to discuss information needs and questions arising from the NRCs review of the audited items.

The NRC staff requests the licensee to have the information referenced in Section 4.0 of this audit plan available and accessible for the NRC staffs review via an internet-based portal within 2 weeks of the date of this audit plan. The NRC staff requests that any supplemental information requested be available and accessible for the NRC staffs review within 1 week of the date of the NRCs notification to the licensee of the new requests. The NRCs licensing project manager will inform the licensee via routine communications when the NRC staff no longer needs access to the portal. The NRC staff requests the licensee to notify the review team when an audit item is added to its portal by sending an e-mail to the review team e-mail address1 or the NRC licensing project manager.

1 BrunswickTSTF-505@usnrc.onmicrosoft.com

7.0 Special Requests The NRC requests access to requested documents and information through a Web-based portal that allows the NRC staff and contractors to access documents over the Internet. The following conditions associated with the online portal must be maintained while the NRC staff and contractors have access to the online portal:

The online portal will be password-protected. A separate password will be assigned to each member of the NRC staff and NRC contractors participating in the audit.

The online portal will prevent the NRC participants from printing, saving, downloading, or collecting any information directly from the online portal.

Conditions of use of the online portal will be displayed on the login screen and will require acknowledgment by each user.

Username and password information should be provided directly to members of the NRC staff and contractors. The NRC licensing project manager will provide the licensee names and contact information of the NRC staff and contractors participating in the audit. All other communications should be coordinated through the NRC project manager.

8.0 Deliverables The NRC staff will develop any RAIs, as needed, via NRR Office Instruction LIC-115 and issue such RAIs separately from audit-related correspondence. The NRC staff will issue an audit summary report prior to completing its review of the LAR.

9.0 References

1. Letter from J. Krakuszeski to the U.S. Nuclear Regulatory Commission, License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion TimesRITSTF Initiative 4b, April 1, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21091A053).

2. TSTF Comments on Draft Safety Evaluation for Traveler TSTF-505, "Provide Risk-Informed Extended Completion Times" and Submittal of TSTF-505, Revision 2, July 2, 2018 (ADAMS Accession No. ML18183A493).

3. U.S. Nuclear Regulatory Commission, NRR Office Instruction LIC-101, Revision 6, License Amendment Review Procedures, July 31, 2020 (ADAMS Accession No. ML19248C539).

4. U.S. Nuclear Regulatory Commission, NRR Office Instruction LIC-111, Revision 1, Regulatory Audits, October 31, 2019 (ADAMS Accession No. ML19226A274).

5. U.S. Nuclear Regulatory Commission, NUREG-0800, Standard Review Plan, Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance, June 2007 (ADAMS Accession No. ML071700658).

6. Nuclear Energy Institute, NEI 06-09, Revision 0-A, Risk-Informed Technical Specification Initiative 4b, Risk-Managed Technical Specification Guidelines, November 2006 (ADAMS Accession No. ML12286A322).

7. U.S. Nuclear Regulatory Commission, NRR Office Instruction LIC-115, Processing Requests for Additional Information, November 6, 2019 (ADAMS Accession No. ML19242B237).

AUDIT ITEMS LIST BRUNSWICK UNITS 1 AND 2 TSTF-505 LICENSE AMENDMENT REQUEST ITEM #

AUDIT REQUEST R1 Reports of full-scope and focused-scope peer reviews (and findings and observations closure reviews referred by date in license amendment request (LAR) Enclosure 2, Section 4.1) and any self-assessment performed for the internal events, internal flooding, and fire PRAs cited in the LAR R2 For the internal events, internal flooding, and fire PRAs, plant-specific documentation (e.g., uncertainty notebooks) related to:

a. The review of the PRA model assumptions and sources of uncertainty

b. Identification of key assumptions and sources of uncertainty for the application R3 PRA notebooks for the modeling of FLEX equipment and FLEX human error probabilities credited in the PRA R4 Documentation supporting the example RICT calculations presented in LAR,, Table E1-2 R5 Any draft or final RICT program procedures (e.g., for risk management actions, PRA functionality determination, and recording limiting conditions for operation)

R6 Plant and PRA configuration control procedures R7 Documentation supporting the development of the real-time risk tool and benchmarking it against the PRA R8 BNP-PSA-094, Revision 4, PSA Model External Flooding Analysis R9 Duke Energy Design Basis Document, Revision 4, DBD-106, Hazard Analysis, December 2018 R10 BNP-PSA-100, Revision 4, BNP High Wind Probabilistic Risk Assessment (HWPRA): Quantification R11 One-line diagrams for alternating current (AC) and DC power distribution systems R12 Information on Emergency Diesel Generators (EDGs) and associated supporting systems (i.e., buses and loads) with regard to:

a. When required

b. Minimum required for each unit R13 Definitions of train, division, subsystem, and load group(s) for proposed TS changes

ITEM #

AUDIT REQUEST R14 Lists of any loads shared between the following:

a. Units

b. ESF AC divisions - medium and low voltage levels (480V)

c. ESF DC divisions R15 Presentation(s) on the following topics:

a. Normal and emergency lineups for offsite AC sources addressing the following:

i.

Offsite primary and secondary feeds for each unit and other possibilities ii.

Any shared equipment or cross-tie(s)

b. Required Engineered Safety Feature (ESF) buses for each unit, to include the following:

i.

4.16 kV ESF buses dual tie breakers, when used, and under what conditions ii.

Dual breakers between Balance-of-Plant (BOP) and ESF 4.16 kV buses iii.

Sharing of ESF loads between units with offsite power available and not available for capability to safely shutdown each unit for design basis accidents and transients

1. Any three EDGs for safe shutdown of non-accident unit and accident unit with station loss of offsite power (LOOP) iv.

Jet assist feature and when applicable

v.

Supplemental diesel generator (DG) capacity and when applicable

c. Automatic starting of EDGs from same division in both units for loss of voltage on BOP bus supplying ESF bus(es)

d. DC system and power distribution systems/subsystems addressing the following:

i.

Critical loads ii.

DC equipment losses that necessitate entry into LCO 3.8.4 Conditions A, B, or C and LCO 3.8.7 Conditions C or D iii.

Sharing of DC loads between two units iv.

Updated Final Safety Analysis Report, Section 8.3.2.1.2, which refers to a 125/250 volts direct current (VDC) system with two batteries and chargers:

1. Full redundancy of each battery for its division of AC ESF equipment

2. Number of chargers per battery and how operated

3. Distribution system for each battery a)

Number of trains or subsystems for each unit and requirements for accidents

AUDIT QUESTION LIST BRUNSWICK UNITS 1 AND 2 TSTF-505 LICENSE AMENDMENT REQUEST PROBABILISTIC RISK ASSESSMENT LICENSING BRANCH A INTERNAL EVENTS PROBABILISTIC RISK ASSESSMENT QUESTIONS APLA Q1 - Instrumentation and Controls Modeling in the probabilistic risk assessment (PRA)

The U.S. Nuclear Regulatory Commission (NRC) Safety Evaluation Report (SER)2 for Nuclear Energy Institute (NEI) Topical Report NEI 06-093 specifies that the License Amendment Request (LAR) for a Risk Informed Completion Time (RICT) program should provide a comparison of the Technical Specification (TS) functions to the PRA modeled functions and that justification should be provided to show that the scope of the PRA model is consistent with the licensing basis assumptions.

Table E1-1 in Enclosure 1 of the LAR identifies the TS Limiting Conditions for Operation (LCOs) and corresponding Conditions proposed to be included in the RICT program and describes how the structures, systems and components (SSCs) covered in the TS LCO are implicitly or explicitly modeled in the PRA. For certain TS LCO Conditions, the table explains that the associated SSCs are not modeled in the PRAs but will be conservatively represented using a surrogate event.

For several TS LCO Conditions, Table E1-1 indicates that instrumentation and control (I&C) detail in existing PRA models is insufficient to explicitly model the Condition. In these cases, the LAR indicates that the inoperability of the associated SSC (e.g., channel) will therefore be modeled using a surrogate event. For other TS LCO Conditions in the RICT program, it is not clear to NRC staff whether I&C is always modeled in sufficient detail to support implementation of Technical Specifications Task Force (TSTF) Traveler 505 (TSTF 505)4, based on documentation in the LAR. Address the following points regarding I&C modeling in the PRA that supports the proposed RICT program:

a. For certain TS LCO Conditions, LAR Table E1-1 states SSCs are modeled consistently with the TS scope but it does not provide any additional details. For these conditions discuss the following:

i. Scope of the I&C SSCs that are explicitly included in the PRA (e.g., bistables, relays, sensors, integrated circuit cards).

ii. Description of the level of detail modeled (e.g., Are all channels of an actuation circuit modeled?).

iii. Discussion of what data are used and whether plant specific data are used.

iv. Discussion of the associated TS functions for which a RICT can be applied.

2 NRC SER for NEI Topical Report 06-09-A, dated May 17, 2007 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML071200238).

3 NEI Topical Report NEI 06-09, Revision 0-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Industry Guidance Document, dated November 2006 (ADAMS Accession No. ML122860402).

4 Traveler TSTF-505, Revision 2, Provide Risk-informed Extended Completion Times, RITSTF Initiative 4b, dated June 14, 2011 (ADAMS Accession No. ML17290A082).

AUDIT QUESTION LIST BRUNSWICK UNITS 1 AND 2 TSTF-505 LICENSE AMENDMENT REQUEST

b. Table E9-1 in Enclosure 9 to the LAR identifies digital feedwater water control system modeling as a potential key source of uncertainty (Item #12). This uncertainty was dispositioned with a sensitivity study. However, it is not clear to the NRC staff whether there are other digital systems (e.g., steam leak detection modules, reactor recirculation speed controls) that are credited in the BNPs PRA.

i. Confirm that no other digital I&C systems are credited in the PRA models that will be used in the RICT program beyond the feedwater control system.

ii. If other digital I&C systems are credited in the PRA models that will be used in the RICT program, then:

1. Identify those systems and provide the results of a sensitivity study on the SSCs in the RICT program demonstrating that the uncertainty associated with modeling digital I&C systems has inconsequential impacts on the RICT calculations.

2. Alternatively, identify which LCOs are determined to be impacted by the digital I&C system modeling for which risk management actions (RMAs) will be applied during a RICT. Explain and justify the criteria used to determine what level of impact to the RICT calculation required additional RMAs.

c. For TS LCO 3.3.5.1 (Emergency Core Cooling System (ECCS) Instrumentation)

Conditions E and F concerning the Automatic Depressurization System (ADS) Trip System, Table E1-1 states that the ADS system instrumentation is not modeled in detail in the PRA, and therefore, a surrogate will be used that represents failure to depressurize the reactor.

Explain further the surrogate proposed to model TS LCO 3.3.5.1 Conditions E and F and explain how the surrogate depressurization function modeled is appropriate for accident sequences.

d. For TS LCO 3.3.5.2 (Reactor Core Isolation Cooling (RCIC) Instrumentation)

Condition B, Table E1-1 states that the individual elements of the RCIC initiation logic are not incorporated in the BNP PRA model. The table further states multiple surrogates that represent common cause failure of the RCIC initiation system would be utilized.

Explain further what component failures will be used as a surrogate to model TS LCO 3.3.5.2 Condition B. Include in this discussion how the surrogate method modeled is appropriate for accident sequences.

e. For LCO 3.3.6.1 (Primary Containment Isolation (PCI) Instrumentation) Condition A, Table E1-1 states that the associated SSCs are not modeled in the PRA and proposes to use a surrogate event that represents failure of the PCI electrical system. The functions covered by this condition operate differently in response to failures of the electrical systems that power the instrumentation logic circuits (e.g., loss of power results in isolation for the Reactor Water Cleanup (RWCU) system, loss of power results in an inability to isolate for the RCIC system).

AUDIT QUESTION LIST BRUNSWICK UNITS 1 AND 2 TSTF-505 LICENSE AMENDMENT REQUEST Describe how the generic electrical system failure surrogate applies to all the functions covered by this condition, given that the various functions operate differently in response to failure of the electrical system.

APLA Q2 - System and Surrogate Modeling Used in the PRA Models The NRC SER for NEI 06-09 specifies that the LAR should provide a comparison of the TS functions to the PRA modeled functions and that justification should be provided to show that the scope of the PRA model is consistent with the licensing basis assumptions. Table E1-1 in of the LAR identifies the TS LCOs and corresponding Conditions proposed to be included in the RICT program and describes how the systems and components covered in the TS LCO are implicitly or explicitly modeled in the PRA. For certain TS LCO Conditions, the table explains that the associated SSCs are not modeled in the PRAs but will be conservatively represented using a surrogate event. For some LCOs, the LAR did not provide sufficient information regarding surrogate PRA modeling that will be used in the RICT calculations for NRC staff to assess the acceptability of the surrogate approaches. To address this observation, address the following:

a. Table E1-1 states that the primary containment air lock is not incorporated in the BNP model and a large pre-existing leak failure surrogate will be used in the PRA model to support a RICT for TS LCO 3.6.1.2 (Primary Containment Air Lock) Condition C. It is unclear to the NRC staff what constitutes a large leak and what PRA model function the pre-existing failure would represent.

Explain further the proposed surrogate to model TS LCO 3.6.1.2 Condition C and discuss how the surrogate is equivalent or bounding for the airlock.

b. Table E1-1 states that not all Primary Containment Isolation Valves (PCIVs) are incorporated in the BNP model and that a pre-existing containment failure surrogate will be used in the PRA model to support a RICT for TS LCO 3.6.1.3 (PCIVs) Condition A. It is unclear to the NRC staff what constitutes a containment failure and what PRA model function the pre-existing failure would represent.

Explain further the surrogate proposed to model TS LCO 3.6.1.3 Condition A and discuss how the surrogate is equivalent or bounding for the non-modeled PCIVs.

c. Table E1-1 states that a vapor suppression function surrogate will be used in the PRA model to support a RICT for TS LCO 3.6.1.6 (Suppression Chamber-to-Drywell Vacuum Breakers) Condition A. It is unclear to the NRC staff what constitutes the vapor suppression function and the SSCs associated with this function.

Explain further the proposed surrogate used to model TS LCO 3.6.1.6 Condition A and discuss how the surrogate vapor suppression function modeled is appropriate for the relevant accident sequences.

APLA Q3 - PRA Modeling Success Criteria The NRC SER for NEI 06-09 specifies that the LAR should provide a comparison of the TS functions to the PRA modeled functions and that justification be provided to show that the scope of the PRA model is consistent with the licensing basis assumptions.

AUDIT QUESTION LIST BRUNSWICK UNITS 1 AND 2 TSTF-505 LICENSE AMENDMENT REQUEST Table E1-1 in Enclosure 1 of the LAR states for TS LCO 3.5.1 (ECCS - Operating) Conditions F, G, and H, the PRA success criteria are three safety relief valves (SRVs), including ADS valves, for non-Anticipated Transient Without Scram (ATWS) scenarios and ten SRVs, including ADS valves, for ATWS scenarios. The PRA success criterion of three SRVs appears to support the depressurization function required by TS LCO 3.5.1 while the criterion of ten SRVs appears to support the over-pressurization protection function required by TS LCO 3.4.3 SRVs.

However, only seven of the eleven SRVs are equipped to provide the automatic depressurization function required by the ADS. Further, Attachment 5 of the LAR indicates that LCO 3.4.3 Condition A will not be included in the RICT program since plant shutdown is required within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if one required SRV is inoperable. It is unclear to the NRC staff how the PRA treatment of SRVs that do not perform the ADS function will be utilized in RICT calculations.

a. Clarify what PRA success criteria are associated with the ADS functions covered by TS LCO 3.5.1 Conditions F, G, and H.

b. Explain why non-ADS SRVs are included in the success criteria for TS LCO 3.5.1. If the non-ADS SRVs are credited for meeting the associated success criteria, include in this explanation how a non-ADS SRV can be credited for satisfying this TS LCO.

APLA Q4 - Unspecified RICT Estimates Table E1-2 in Enclosure 1 of the LAR provides RICT estimates for TS Action Statements proposed to be in the scope of the RICT program. However, RICT estimates for several LCO Conditions and Required Actions (3.3.1.1 A and B, 3.3.5.1 E and F, 3.3.6.1.A, 3.8.4.A, 3.8.7.B, C and D) are not provided. In addition, Note 3 of Table E1-2 states:

The RICT estimate in this table represents the most limiting RICT calculation based on the most limiting component. In accordance with NEI 06-09-A, depending upon the specific inoperable SSC which causes the TS LCO to be not met, the level of risk calculated varies, and a different RICT may be calculated for different inoperable SSCs within the Action.

a. Explain what is meant by depending upon the specific inoperable SSC which causes the TS LCO to be not met, the level of risk calculated varies. Provide examples and associated RICT estimates.

b. For some Conditions with unspecified RICT estimates (LCO 3.3.1.1 Conditions A and B, LCO 3.3.5.1 Conditions E and F, LCO 3.3.6.1 Condition A), the NRC staff notes that Table E1-1 of the LAR proposes surrogate events for modeling failure of the associated SSCs. It appears that the proposed surrogate events may be restricting planned entry in an RICT for these conditions consistent with the guidance in NEI 06-09-A, which prohibits planned entry in a RICT when the instantaneous Core Damage Frequency (CDF) limit of 1E-03 per year or Large Early Release Frequency (LERF) limit of 1E-04 per year are exceeded.

Confirm BNPs intent to use the proposed surrogates for the RICT program.

AUDIT QUESTION LIST BRUNSWICK UNITS 1 AND 2 TSTF-505 LICENSE AMENDMENT REQUEST APLA Q5 - PRA Model Update Process Section 2.3.4 of NEI 06-09 specifies that criteria shall exist in PRA configuration risk management to require PRA model updates concurrent with implementation of facility changes that significantly impact RICT calculations.

LAR Enclosure 7 states that if a plant change or a discovered condition is identified and can have significant impact on the RICT calculations then an unscheduled update of the PRA models will be implemented. More specifically, the LAR states that if the plant changes meet specific criteria defined in the plant PRA and update procedures then the change will be incorporated into applicable PRA models without waiting for the next periodic PRA update.

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

APLA Q6 - Total Risk Consideration of State-of-Knowledge Correlation and Modeling Updates Regulatory Guide (RG) 1.1745 provides the risk acceptance guidance for total CDF (1E-04 per year) and LERF (1E-05 per year). Table E5-1b of Enclosure 5 to the LAR shows that the total LERF for BNP Unit 2 is 8.51E-06 per year using the baseline Model of Record (MOR) PRA.

Based on RG 1.174 and Section 6.4 of NUREG-18556, Revision 1, for a Capability Category II risk evaluation, the mean values of the risk metrics (total and incremental values) need to be compared against the risk acceptance guidelines. The mean values referred to in this context are the means of the probability distributions that result from the propagation of the uncertainties on the PRA input parameters and model uncertainties explicitly reelected in the PRA models. In general, the point estimate CDF and LERF values obtained by quantification of the cutset probabilities using mean values for each basic event probability do not produce a true mean of the CDF and LERF. Under certain circumstances, a formal propagation of uncertainty may not be required if it can be demonstrated that the State of Knowledge Correlation (SOKC) is unimportant (i.e., the risk results are well below the acceptance guidelines).

Demonstrate that the total risk for Unit 2 will conform to the RG 1.174 risk acceptance guidelines (i.e., CDF < 1E-04 per year and LERF < 1E-05 per year) after the internal events and fire PRA models are updated to include the potential increases in risk associated with SOKC and updates to PRA models performed in response to NRC staff requests. Include identification of the fire PRA parameters for which SOKC was applied in the parametric uncertainty analysis of fire events.

APLA Q7 - Supplemental Diesel Sensitivity Analysis Topical Report NEI 06-09 and the NRCs SER for NEI 06-09 specify that an LAR for RICT program implementation should identify key assumptions and sources of uncertainty and should assess/disposition each as to its impact on the RICT program. Section 2.3.4 of NEI 06-09-A states that sensitivity studies should be performed on the base PRA model prior to initial 5 NRC RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, dated January 2018 (ADAMS Accession No. ML17317A2565).

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

AUDIT QUESTION LIST BRUNSWICK UNITS 1 AND 2 TSTF-505 LICENSE AMENDMENT REQUEST implementation of the RICT program on uncertainties that could potentially impact the results of an RICT calculation. NEI 06-09-A also states that the insights from the sensitivity studies should be used to develop appropriate risk management actions (RMAs), including highlighting risk significant operator actions, confirming availability and operability of important standby equipment, and assessing the presence of severe or unusual environmental conditions.

to the LAR identifies key assumptions and sources of uncertainty associated with the PRA MOR. Item #8 of Table E9-1 identified the use assumed failure rates for the non-safety Supplemental Diesel Generator (SUPP-DG) as a source of uncertainty and provided results of sensitivity study on RICT estimates. The Case A47-1 (Distribution Systems - Operating, Unit 1-One alternating current (AC) electrical power distribution subsystem inoperable for planned maintenance due to either inoperable load group E3 bus(es) or inoperable load group E4 bus(es)) demonstrates a reduction of 2.9 days of the RICT calculation which constitutes an 18.8% impact.

Address the following related to the SUPP-DG:

a. Discuss whether the RICTs for other TS LCOs (i.e., those in scope of the RICT program but not evaluated in Table E9-1 Item #8 of the LAR) and for plant configurations involving more than one LCO entry are significantly impacted by the SUPP-DG uncertainties. For those TS LCOs that are significantly impacted by this source of uncertainty, identify the LCOs and how this source of uncertainty impacts the RICT (e.g.,

describe and provide the results of a sensitivity study). Also, discuss the basis for the chosen plant configurations involving more than one LCO entry.

b. Describe how sources of uncertainty associated with SUPP-DG will be addressed in the RICT program. Provide updated RMAs that may be considered during a RICT program entry to minimize any potential adverse impact from SUPP-DG uncertainties and explain how these RMAs are expected to reduce the risk associated with this source of uncertainty.

OR

c. Provide a detailed justification that the sensitivities of the computed RICTs to SUPP-DG uncertainties do not need to be addressed in the RICT program as required by Section 2.3.4 of NEI 06-09-A.

APLA Q8 - Supplemental Diesel Failure Data Item #8 of Table E9-1 notes that the PRA MOR uses generic industry failure data for standard Emergency DGs (EDGs), despite also acknowledging that non-safety related DGs typically have higher failure probabilities than EDGs. In addition to the sensitivity analyses discussion above, provide justification for using failure probabilities for EDGs in lieu of using non-safety related DG failure probabilities. This justification should focus on surveillance frequencies, quality, maintenance activities and other factors that typically differentiate commercial and safety grade equipment.

APLA Q9 - PRA Success Criteria for Service Water Systems The descriptions of the Nuclear Service Water (NSW) and Conventional Service Water (CSW) systems for TS LCO 3.7.2 (Service Water and Ultimate Heat Sink) Condition B in Table E1-1 of

AUDIT QUESTION LIST BRUNSWICK UNITS 1 AND 2 TSTF-505 LICENSE AMENDMENT REQUEST to the LAR suggest that the success criteria for these systems are a function of various plant configurations. Additionally, based on the information provided in LAR Table E1-1 it appears the PRA success criteria credit an operator action to throttle the Turbine Building Closed Cooling Water (TBCCW) heat exchanger outlet valve to reduce the required number of CSW pumps. Relative to the NSW and CSW systems as they are used in the PRA and the proposed RICT program, address the following:

a. Provide a further detailed discussion/explanation of the modeling of the NSW/CSW systems in the PRA and the associated success criteria. Explain whether and how the different success criteria are captured in the Configuration Risk Management Program (CRMP) for the real-time plant configuration.

b. Discuss the operator action to throttle the TBCCW heat exchanger outlet valve and summarize the Human Reliability Analysis (HRA) analysis for this action.

APLA Q10 - HRA for FLEX Operator Actions Section 4.4 of Enclosure 9 to the LAR discusses how FLEX strategies were used in the current PRA model to support implementation of a RICT program. This section notes that the FLEX equipment currently credited in the PRA model includes permanently installed diesel generators, portable pumps and portable air compressors. It also explains that post-initiator operator actions modeled include failure to load shed, failure to align and start FLEX diesel generators, failure to refuel FLEX diesel generators (if needed), failure to align and start FLEX portable pumps and failure to align and start FLEX air compressors.

The staff notes that the Electric Power Research Institute (EPRI) issued Technical Update 30020130187 which includes examples and guidance for how to perform HRA for the use of onsite portable equipment in a variety of contexts. Address the following items related to FLEX strategies and the HRA used to support implementation of the RICT program:

a. Describe the HRA methodology used for crediting operator actions related to FLEX equipment.

b. Describe the credited operator actions related to FLEX equipment and discuss the methodology used to assess the associated human-error probabilities and the licensee personnel that performs these actions. The discussion should include a summary of how the licensee evaluated the impact of the plant-specific human error probabilities and associated scenario-specific performance shaping factors listed in (a)-(j) of supporting requirement HR-G3 of American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) Standard RA-Sa-20098, as endorsed by RG 1.2009.

c. Regarding FLEX pre-initiators evaluation, discuss whether maintenance procedures for the portable equipment were reviewed for possible pre-initiator human failures that 7 EPRI Technical Update 3002013018, Human Reliability Analysis (HRA) for Diverse and Flexible Mitigation Strategies (FLEX) and Use of Portable Equipment: Examples and Guidance, November 30, 2018.

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

9 NRC RG 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, dated December 2020 (ADAMS Accession No. ML20238B871).

AUDIT QUESTION LIST BRUNSWICK UNITS 1 AND 2 TSTF-505 LICENSE AMENDMENT REQUEST renders the equipment unavailable during an event, and whether the probabilities of the pre-initiator human failure events were assessed as described in HLR-HR-D of ASME/ANS RA-Sa-2009, as endorsed by RG 1.200.

d. Discuss FLEX strategy initiation. Discuss whether the procedures for the initiation or entry into mitigation strategies are explicit. Discuss the technical bases for probability of failure to initiate mitigating strategies. Include in this discussion the cue to declare an Extended Loss of AC Power (ELAP) and how this action is incorporated into the PRA model.

APLA Q11 - FLEX Equipment in PRA Model Section 4.4 of Enclosure 9 to the LAR discusses how FLEX strategies were credited in the current PRA model to support implementation of a RICT program. This section notes that the FLEX equipment currently credited in the PRA model includes permanently installed diesel generators, portable pumps and portable air compressors. Address the following:

a. Discuss whether the FLEX diesel generators are similar to other permanently installed plant equipment (i.e., SSCs with sufficient plant-specific or generic industry data).

Compare failure data of the FLEX diesel generators with that used for similar plant equipment credited elsewhere in the PRA (e.g., EDGs).

b. Describe the events for which portable equipment is credited in the PRA models (e.g.,

ELAP only, internal events, and external hazards that are within or beyond design basis). Additionally, describe the sources of data used for any credited portable FLEX equipment and denote whether any plant-specific failure rates are higher than expected based on generic industry data.

APLA Q12 - PRA Model Upgrades with FLEX Strategies Section 4.4 of Enclosure 9 to the LAR discusses how FLEX strategies were credited in the current PRA model to support implementation of a RICT program. However, no information is provided that denotes whether crediting FLEX strategies in the PRA constitutes a PRA upgrade.

a. Describe whether incorporation of FLEX equipment into the supporting PRA model constitutes an upgrade to the PRA, along with the basis for the decision including the source of the definition of upgrade used.

b. If it is determined that inclusion of FLEX strategies constitutes an upgrade, describe supporting peer reviews that were done, as well as any Finding Closure Reviews and provide the disposition of remaining open findings.

APLA Q13 - Open Phase Protection Modeling In response to the January 30, 2012, Open Phase Condition (OPC) event at the Byron Generating Station, the NRC issued Bulletin 2012-0110. As part of the initial Voluntary Industry Initiative (VII) for mitigation of the potential for the occurrence of an OPC in electrical 10 NRC Bulletin 2012-01, Design Vulnerability in Electric Power System, dated July 27, 2012 (ADAMS Accession No. ML12074A115).

AUDIT QUESTION LIST BRUNSWICK UNITS 1 AND 2 TSTF-505 LICENSE AMENDMENT REQUEST switchyards11, licensees have made the addition of an Open Phase Isolation System (OPIS).

As per SRM-SECY-16-006812, the NRC staff was directed by the Commission to ensure that licensees have appropriately implemented OPIS and that licensing bases have been updated accordingly. Inspections of OPIS by NRC staff are currently underway. From revised voluntary initiative13 and resulting industry guidance in NEI 19-0214 on estimating OPC and OPIS risk, it is understood that the risk impact of an OPC can vary widely dependent on electrical switchyard configuration and design. Considering this observation, provide the following information:

a. A discussion of the risk impact of OPC and OPIS at BNP.

b. Discuss if the risk impact of OPC and OPIS have been or, are to be, incorporated as part of the plant MOR. If so, provide the following:

i.

The schedule for the inclusion of OPC and OPIS to the MOR.

ii.

The impact, if any, to key assumptions and sources of uncertainty.

iii.

A discussion of the HRA methods and assumptions used for OPIS alarm manual response.

iv.

The impact to external events, e.g., fire, seismic, flooding, high winds, tornado, other external events, etc.

v.

A discussion of the risk impact of inadvertent OPIS actuation and justification for its exclusion.

c. If OPC and OPIS are not planned to be included in the MOR, provide justification why the risk impact is not included by performing either a qualitative or sensitivity analysis.

APLA Q14 - Generic Fire PRA Question RG 1.200 states NRC reviewers, [will] focus their review on key assumptions and areas identified by peer reviewers as being of concern and relevant to the application. NRC has issued updated Fire PRA (FPRA) guidance subsequent to approval of the LAR for implementation of National Fire Protection Association (NFPA) 80515 at BNP. This updated set of guidance documents includes the following:

11 Letter from Anthony R. Pietrangelo (NEI) to Mark A. Satorius, Industry Initiative on Open Phase Condition - Functioning of Important-to-Safety Structures, Systems and Components (SSCs), dated October 9, 2013 (ADAMS Accession No. ML13333A147).

12 NRC Staff Requirements Memorandum SECY-16-0068, Interim Enforcement Policy for Open Phase Conditions in Electric Power Systems for Operating Reactors, dated March 9, 2017 (ADAMS Accession No. ML17068A297).

13 Letter from Doug True (NEI) to Ho Nieh (NRC), Industry Initiative on Open Phase Condition, Revision 3, dated June 6, 2019 (ADAMS Accession No. ML19163A176).

14 Technical Report NEI 19-02, Revision 0, Guidance for Assessing Open Phase Condition Implementation Using Risk Insights, April 2019 (ADAMS Accession No. ML19122A321).

15 NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants.

AUDIT QUESTION LIST BRUNSWICK UNITS 1 AND 2 TSTF-505 LICENSE AMENDMENT REQUEST NUREG-218016 regarding the updated approach to credit incipient fire detections systems.

NUREG-2230 (EPRI Report 3002016051)17 regarding electrical cabinet fires.

NUREG-2178 (EPRI Report 3002016052)18 regarding heat release rates.

State where whether the updated guidance outlined above (i.e., NUREG-2180, NUREG-2230, and NUREG-2178) has been incorporated into the BNP FPRA that will be used to support the RICT program. If the updated guidance has not been incorporated into the FPRA, then discuss the potential impact on the proposed RICT program if the revised guidance were to be incorporated into the existing BNP FPRA methodology.

16 NUREG-2180, "Determining the Effectiveness, Limitations, and Operator Response for Very Early Warning Fire Detection Systems in Nuclear Facilities (DELORES-VEWFIRE)," December 2016 (ADAMS Accession No. ML16343A058).

17 NUREG-2230, Methodology for Modeling Fire Growth and Suppression Response for Electrical Cabinets Fires in Nuclear Power Plants, dated June 2020 (ADAMS Accession No. ML20157A148).

18 NUREG-2178, Volume 2, Refining And Characterizing Heat Release Rates From Electrical Enclosures During Fire (RACHELLE-FIRE) Volume 2: Fire Modeling Guidance for Electrical Cabinets, Electric motors, Indoor Dry Transformers, and the Main Control Board, dated June 2020 (ADAMS Accession No. ML20168A655).

AUDIT QUESTION LIST BRUNSWICK UNITS 1 AND 2 TSTF-505 LICENSE AMENDMENT REQUEST PRA Licensing Branch C (APLC)

External Hazards APLC Q1 - Seismic Core Damage Frequency Calculation As clarified in the NRC SER for NEI 06-09, other sources of risk (i.e., seismic and other external events) must be quantitatively assessed if they contribute significantly to configuration-specific risk. The SER for NEI 06-09 also states that bounding analyses or other conservative quantitative evaluations are permitted where realistic PRA models are unavailable.

Section 6.1 of the Enclosure 4 to the LAR, the licensee provided seismic bounding analysis, and calculated seismic core damage frequency (SCDF) and seismic large early release frequency (SLERF). The licensee used the review-level earthquake (RLE) spectral ratios that were developed for IPEEE assessment, as shown in Table E4-1 of the LAR. These RLE spectral ratios are very similar to those in Table C-2 of the safety/risk assessment results for Generic Issue 19919 or in Table 2 of a letter20 from EPRI to NEI regarding seismic hazard estimates with an assumption of the same ratio between 5 hertz (Hz) and 2.5Hz. By using these spectral ratios, the calculated SCDF penalties, ranging from 1.8E-7 /yr to 9.5E-6 /yr, are very different among the five frequencies, peak ground acceleration (PGA), 10, 5, 2.5 and 1 Hz shown in Table E4-3 of the LAR. This difference is likely caused by using spectral ratios developed from seismic hazard curves that are different from the seismic hazard curves used in this application.

In addition, the seismic bounding analysis used an average of 5 frequencies of seismic hazard curves (PGA, 10, 5, 2.5 and 1 Hz), instead of an average of 4 frequencies of seismic hazard curves (PGA, 10, 5, and 1 Hz) proposed in Generic Issue 199. The licensee compared the difference between the two methods, with a non-conservative value from the 5-frequency method (2.81E-6 vs 3.46E-6 for SCDF and 1.35E-6 vs 1.67E-6 for SLERF penalties). However, the licensee did not provide the rationale for selecting a non-conservative averaging method.

a. Regarding the RLE approach:

i. Provide justification of why the spectral ratios developed from the RLE are applicable to the seismic hazard curves used in this application.

ii. Alternatively to Part (i), if the justification cannot be provided, calculate the spectral ratios based on the seismic hazard curves used for this application and use them to obtain and provide updated seismic risk penalty values.

b. Regarding the averaging method:

i. Provide the rationale for selecting a non-conservative averaging method.

ii. Alternatively to Part (i), if the rationale cannot be provided, use an averaging method 19 Safety/Risk Assessment Results for Generic Issue 199, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants, dated September 2, 2010 (ADAMS Package Accession No. ML100270582).

20 Letter from Stuart Lewis (EPRI) to Anthony R. Pietrangelo (NEI), Fleet Seismic Core Damage Frequency Estimates for Central and Eastern U.S. Nuclear Power Plants Using New Site-Specific Seismic Hazard Estimates, dated March 11, 2014 (ADAMS Accession No. ML14083A586).

AUDIT QUESTION LIST BRUNSWICK UNITS 1 AND 2 TSTF-505 LICENSE AMENDMENT REQUEST that is consistent with commonly accepted method and provide updated seismic risk penalty values.

AUDIT QUESTION LIST BRUNSWICK UNITS 1 AND 2 TSTF-505 LICENSE AMENDMENT REQUEST ELECTRIC ENGINEERING BRANCH (EEEB)

AUDIT QUESTIONS EEEB Q1 - Design Basis Accident for EDG Loading Address the following regarding EDG loading under Design basis Accident (DBA) conditions:

a. Is the licensing basis of the plant, a [loss-of-coolant accident] LOCA in one unit and a station loss of offsite power [LOOP] and is this worst-case accident for loading of EDGs?

b. Please provide total loading (final load sum) of each DG for worst-case licensing basis accident for ESF bus(es).

EEEB Q2 - LCO 3.8.1 Address the following inquiries regarding TS LCO 3.8.1:

a. Explain the applicability of Insert 2 in TS LCO 3.8.1, Condition D.4.

b. The LAR states that TS LCO 3.8.1, Required Action D.2 was added as part of amendment numbers 264 and 292 to support extension of the completion time for Required Action D.1 (restoration of an inoperable EDG). The justification for this amendment denoted that the Supplemental Diesel Generator (SUPP-DG) provided additional defense-in-depth. Please provide justification for the use of RICT with the proposed removal of the Supplemental DG from LCO 3.8.1, Required Action D.2.

c. For Table E1-1 in LAR, LCO 3.8.1 Conditions C, D, E, and F, please explain why each of these require three emergency buses to be available for all events. Additionally, please discuss what all events refers to in the context of this inquiry.

d. Please explain why in table E1-1, for TS 3.8.1, Condition D, it is stated that an An EDG is adequate for each bus. Three emergency buses are adequate for all events when only three emergency buses are needed.

EEEB Q3 - RICT Impacts and PRA Modeling of Multi-Unit AC Distribution Systems In general, the LCOs and Conditions described in Table E1-1 of Enclosure 1 to the LAR address the safety functions applicable to one unit (i.e., the unit experiencing a design basis accident in in the context of safety analyses context) and the AC sources necessary for safe shutdown.

Considering the limitations of PRA for dynamic modeling of more than one unit, please explain how the information provide in Table E1-1 addresses the capability to shut down both units.

EEEB Q4 - Electric Distribution Systems Design Success Criteria The design success criteria column in Table E1-1 of Enclosure 1 to the LAR includes short descriptions of the existing assumptions of success for each of the TS LCO conditions proposed for inclusion into the RICT program, including those relative to the electrical distribution system. Address the following as they relate to the design success criteria for the LCOs below:

AUDIT QUESTION LIST BRUNSWICK UNITS 1 AND 2 TSTF-505 LICENSE AMENDMENT REQUEST

a. Please explain the need for three trains of DC power per unit, as discussed in Table E1-1 for LCO 3.8.4, Condition A

b. Please explain need for three of four DC distribution systems as discussed in Table E1-1 for LCO 3.8.7, Conditions C and D.

c. Please explain need for three of four load groups as discussed in Table E1-1 for LCO 3.8.7, Condition A.

EEEB Q5 - Conservatism in RICT Estimates Note 3 in Table E1-2 of Enclosure 1 to the LAR indicates that the RICT estimates for certain TS action statements were derived from the most limiting RICT calculation based on the most limiting component. For action statements in Table E1-2 where Note 3 is applicable, provide the configurations of the associated SSCs and identify limiting components including their RICT estimates.

EEEB Q6 - TS Markup for LCO 3.8.4 Table E1-1 of Enclosure 1 to the LAR indicates that LCO 3.8.4 Conditions B and C will be included as part of the RICT program. However, the TS markups in Attachments 2 and 3 of the LAR do not include these TS conditions. Clarify whether LCO 3.8.4 Conditions B and C are within scope of the LAR and, if so, whether a note for loss of function is necessary for the proposed insert relative to TS 3.8.4 Condition C.

EEEB Q7 - Modeling of Electric Plant for LCO 3.8.7 The proposed insert to Section 5.5 of the BNP TSs includes the following statement A RICT may only be utilized in MODE 1 and 2 (see Attachments 2 and 3 of the LAR). Attachment 1 of the LAR on page 6 indicates, in part, that LCO 3.8.7 Condition A is a variation from the LCOs referenced in TSTF-505, Revision 2 (i.e., Standard TSs in NUREG-143321). The description of this variation indicates with the opposite unit (i.e., the shutdown unit) in MODE 4 or 5 and one AC electrical power distribution subsystem inoperable for planned status maintenance, the remaining AC electrical power distribution load groups can support the minimum safety functions necessary to shut down the operating unit and maintain both reactors in a safe condition. If the opposite unit (Unit 1 or Unit 2) is in MODE 1, 2 or 3, then LCO 3.8.7 Required Action A.1 for the opposite unit requires restoration of the associated AC electrical power distribution subsystem within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the inoperability.

Clarify whether the configurations associated with TS 3.8.7 Condition A for the opposite unit are explicitly modeled in the BNP PRA and whether the opposite unit (either operating or shutdown) is modeled in PRA when affected unit is in the RICT program for this TS Condition.

21 NUREG-1433, Standard Technical Specifications General Electric Plants (BWR/4), dated April 2012 (ADAMS Accession No. ML12104A192).

AUDIT QUESTION LIST BRUNSWICK UNITS 1 AND 2 TSTF-505 LICENSE AMENDMENT REQUEST TECHNICAL SPECIFICATIONS BRANCH (STSB)

AUDIT QUESTIONS STSB Q1 - Technical Specification Markups Attachments 2 and 3 include markups of the Technical Specifications (TSs) to support the proposed implementation of a RICT program at BNP Unit 1 and Unit 2, respectively. Address the following inquiries and observations relative to these markups:

a. Example 1.3-8 does not match the formatting in TSTF-505, Revision 2, in some places.

In TSTF-505, the title is in all capital letters and underscored, and logical connectors are underscored.

b. The proposed administrative controls in TS 5.5.15 paragraph c.2 of Insert 3 states Action Completion Time instead of Required Action Completion Time, the latter of which is provided in TSTF-505, Revision 2.

c. The proposed administrative controls in TS 5.5.15 paragraph e of Insert 3 include the phase this license amendment. In lieu of the phrase this license amendment, discuss whether the phrases Amendment # xxx or, as discussed in the TSTF-505 model SE, this program would provide more clarity for this paragraph.

STSB/SCPB Q2 - Table E1-3 Additional Justifications Required (starts on pdf page 206)

The proposed RICT program includes LCO 3.7.6 which addresses operability of the BNP main turbine bypass valves (LCO 3.7.7 in Standard TSs). Condition A is entered when the LCO is not met. Table E1-3 in Enclosure 1 of the LAR provides the additional justification required for this LCO consistent with Table 1 in TSTF-505, Revision 2. For Unit 2, discuss further how the common cause failure basic event for all bypass valves proposed for the RICT program ensures that the PRA success criteria bound the design-basis success criteria.

STSB/EICB Q3 - Potential Loss of Function For LCO 3.3.5.1, entry into Condition A (one or more channels inoperable) could be required during scenarios involving a Loss of Function (LOF). For LCO 3.3.5.1, LOF occurs when there is a loss of initiation capability resulting from a loss of one or more required channels. Required Action A.1 directs entry into the conditions listed in Table 3.3.5.1-1. This includes entry into Conditions B, C, D, E and F which are within the scope of the proposed RICT program. Given that TSTF-505, Revision 2, does not allow LOF conditions, discuss how potential LOF scenarios will be treated in the proposed RICT program.

STSB/EICB/SNSB Q4 - Table E1-1 Questions The following inquiries relate to Table E1-1 in Enclosure 1 of the LAR:

a. For LCO 3.3.5.1 Condition C (functions 1.d, 2.f), state the minimum number of channels required for success.

AUDIT QUESTION LIST BRUNSWICK UNITS 1 AND 2 TSTF-505 LICENSE AMENDMENT REQUEST

b. For LCO 3.3.5.1 Condition E, the design success criteria (DSC) are not described.

Please provide information regarding the DSC for this condition.

c. For LCO 3.3.5.1 Condition F, Table 3.3.5.1-1 of the BNP TS indicate that functions 4.b and 5.b have one (1) required channel per function. State the DSC for these functions (i.e., minimum number of channels required.)

d. For LCO 3.3.5.1 Condition F (functions 4.d, 4.e, 5.d, 5.e), Table 3.3.5.1-1 of the BNP TSs indicates that there are six (6) instrumentation channels used for detection of running Core Spray and Residual Heat Removal (RHR) pumps for each Automatic Depressurization System (ADS) trip system (A and B). Correspondingly, the DSC description for this condition states that 12 channels are provided to ensure that no single instrument failure can preclude ADS initiation. State the DSC (i.e., minimum number of channels required).

e. For LCO 3.3.6.1 Condition A (10 rows in Table E1-1 for varying isolation functions), state the minimum number of channels required to meet the DSC applicable to each isolation function.

f.

For LCO 3.5.1 Condition D, state the minimum equipment required to meet the DSC.

g. For LCO 3.5.3 Condition A, state the minimum complement of equipment required to meet the DSC.

ML21260A074 OFFICE NRR/DRA/APLA NRR/DRA/APLA/BC NRR/DORL/LPL2-2/PM NAME WJessup RPascarelli AHon DATE 9/16/2021 9/16/2021 9/16/2021 OFFICE NRR/DORL/LPL2-2/LA NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM NAME RButler DWrona AHon DATE 9/21/2021 9/22/2021 9/22/2021