ML21067A224

Issuance of Amendment Nos. 305 and 333 to Revise License Conditions to Adopt 10 CFR 50.69, Risk Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors

ML21067A224
Person / Time
Site:	Brunswick
Issue date:	04/30/2021
From:	Andrew Hon Plant Licensing Branch II
To:	Krakuszeski J Duke Energy Progress
Hon A
References
EPID L-2020-LLA-0152
Download: ML21067A224 (26)

Text

April 30, 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 - ISSUANCE OF AMENDMENT NOS. 305 AND 333 TO REVISE LICENSE CONDITIONS TO MODIFY APPROVED 10 CFR 50.69, RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS, AND COMPONENTS FOR NUCLEAR POWER REACTORS CATEGORIZATION PROCESS (EPID L-2020-LLA-0152)

Dear Mr. Krakuszeski:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment Nos. 305 and 333 to Renewed Facility Operating License Nos. DPR-71 and DPR-62 for the Brunswick Steam Electric Plant, Units 1 and 2, respectively. These amendments are in response to your license amendment request dated July 9, 2020, as supplemented by a letter dated November 24, 2020.

The amendments revise license conditions to the Renewed Facility Operating Licenses to allow the implementation of Title 10 of the Code of Federal Regulations (10 CFR) 50.69, Risk-informed categorization and treatment of structures, systems and components [SSCs] for nuclear power reactors. Specifically, the revised license conditions replace the use of the external flood Probabilistic Risk Assessment for categorization of SSCs under Duke Energy Progress, LLCs previously approved 10 CFR 50.69 program with external flood hazard screening.

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register Notice.

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

Enclosures:

1. Amendment No. 305 to DPR-71

2. Amendment No. 333 to DPR-62

3. Safety Evaluation cc: Listserv DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 305 Renewed License No. DPR-71 1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment filed by Duke Energy Progress, LLC (the licensee), dated July 9, 2020, as supplemented by a letter dated November 24, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes as indicated in the attachment to this license amendment, and paragraph 3 of Renewed Facility Operating License No. DPR-71 is hereby amended to read as follows:

Additional Conditions The Additional Conditions contained in Appendix B, as revised through Amendment No. 305, are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Additional Conditions.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 120 days.

FOR THE NUCLEAR REGULATORY COMMISSION David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Appendix B, Additional Conditions Date of Issuance: April 30, 2021 David J.

Wrona Digitally signed by David J. Wrona Date: 2021.04.30 15:39:56 -04'00'

Renewed License No. DPR-71 Amendment No. 305 ATTACHMENT TO LICENSE AMENDMENT NO. 305 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-71 DOCKET NO. 50-325 Replace the following pages of Renewed Facility Operating License No. DPR-71 and Appendix B, Additional Conditions, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.

Renewed Facility Operating License Remove Page Insert Page 10 10 Appendix B, Additional Conditions Remove Page Insert Page App. B-5 App. B-5

Renewed License No. DPR-71 Amendment No. 305 3.

Additional Conditions The Additional Conditions contained in Appendix B, as revised through Amendment No. 305, are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Additional Conditions.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:

1. Unit 1 - Technical Specifications - Appendices A and B Date of Issuance: June 26, 2006

Brunswick Unit 1 App. B-5 Amendment No. 305 Amendment Number 305 Additional Conditions Duke Energy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and high winds; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in Unit 1 License Amendment No. 305 dated April 30, 2021.

Duke Energy will complete the implementation items listed in Attachment 1 of Duke letter to the NRC dated November 24, 2020, prior to implementation of 10 CFR 50.69 in accordance with the categorization process described above. The issues identified in the attachment will be addressed and any associated changes will be made prior to implementation of 10 CFR 50.69 in accordance with the categorization process described above.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

Implementation Date Upon implementation of Amendment No. 305.

DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 333 Renewed License No. DPR-62 1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment filed by Duke Energy Progress, LLC (the licensee), dated July 9, 2020, as supplemented by a letter dated November 24, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes as indicated in the attachment to this license amendment, and paragraph 3 of Renewed Facility Operating License No. DPR-62 is hereby amended to read as follows:

Additional Conditions The Additional Conditions contained in Appendix B, as revised through Amendment No. 333, are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Additional Conditions.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 120 days.

FOR THE NUCLEAR REGULATORY COMMISSION David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Appendix B, Additional Conditions Date of Issuance: April 30, 2021 David J.

Wrona Digitally signed by David J. Wrona Date: 2021.04.30 15:40:24 -04'00'

Renewed License No. DPR-62 Amendment No. 333 ATTACHMENT TO LICENSE AMENDMENT NO. 333 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-62 DOCKET NO. 50-324 Replace the following pages of Renewed Facility Operating License No. DPR-62 and Appendix B, Additional Conditions, with the attached revised pages. The revised pages are identified by amendment number and contains marginal lines indicating the area of change.

Renewed Facility Operating License Remove Page Insert Page 10 10 Appendix B, Additional Conditions Remove Page Insert Page App. B-5 App. B-5

Renewed License No. DPR-62 Amendment No. 333 M.

Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(1)

Fire fighting response strategy with the following elements:

1. Pre-defined coordinated fire response strategy and guidance

2. Assessment of mutual aid fire fighting assets

3. Designated staging areas for equipment and materials

4. Command and control

5. Training of response personnel (2)

Operations to mitigate fuel damage considering the following:

1. Protection and use of personnel assets

2. Communications

3. Minimizing fire spread

4. Procedures for implementing integrated fire response strategy

5. Identification of readily-available pre-staged equipment

6. Training on integrated fire response strategy

7. Spent fuel pool mitigation measures (3)

Actions to minimize release to include consideration of:

1. Water spray scrubbing

2. Dose to onsite responders N.

The licensee shall implement and maintain all Actions required by to NRC Order EA-06-137, issued June 20, 2006, except the last action that requires incorporation of the strategies into the site security plan, contingency plan, emergency plan and/or guard training and qualification plan, as appropriate.

3.

Additional Conditions The Additional Conditions contained in Appendix B, as revised through Amendment No. 333, are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Additional Conditions.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:

1. Unit 2 - Technical Specifications - Appendices A and B Date of Issuance: June 26, 2006

Brunswick Unit 2 App. B-5 Amendment No. 333 Amendment Number 333 Additional Conditions Duke Energy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and high winds; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in Unit 2 License Amendment No. 333 dated April 30, 2021.

Duke Energy will complete the implementation items listed in Attachment 1 of Duke letter to the NRC dated November 24, 2020, prior to implementation of 10 CFR 50.69 in accordance with the categorization process described above.

The issues identified in the attachment will be addressed and any associated changes will be made prior to implementation of 10 CFR 50.69 in accordance with the categorization process described above.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

Implementation Date Upon implementation of Amendment No. 333.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 305 AND 333 TO RENEWED FACILITY OPERATING LICENSE NOS. DPR-71 AND DPR-62 DUKE ENERGY PROGRESS, LLC BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 DOCKET NOS. 50-325 AND 50-324

1.0 INTRODUCTION

By application dated July 9, 2020 (Reference 1), as supplemented by letter dated November 24, 2020 (Reference 2), Duke Energy Progress, LLC (Duke, the licensee) submitted a license amendment request (LAR) regarding the Brunswick Steam Electric Plant (BSEP), Units 1 and 2.

The proposed amendment would modify the licensing basis to implement a change to the approved voluntary implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.69, Risk-informed categorization and treatment of structures, systems, and components [SSCs] for nuclear power reactors. Specifically, the proposed amendments would revise a license condition to replace the use of the external flood (XF)

Probabilistic Risk Assessment (PRA) for categorization of SSCs under the licensees previously approved 10 CFR 50.69 program with XF hazard screening. The licensee applied the guidance in Section 5.4 of Nuclear Energy Institute (NEI) 00-04, 10 CFR 50.69 SSC Categorization Guideline (Reference 3), endorsed by the U.S. Nuclear Regulatory Commission (NRC) staff in Regulatory Guide (RG) 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance, for Trial Use (Reference 4) for XF hazard screening into the previously approved 10 CFR 50.69 categorization process.

The affected license condition was added when the NRC approved the licensees use of 10 CFR 50.69 on September 17, 2019 (Reference 5). As stated in 10 CFR 50.69, a licensee may voluntarily comply with 10 CFR 50.69 as an alternative to compliance with the following requirements for certain SSCs after it submits, and the NRC approves, an application for license amendment: (i) 10 CFR Part 21; (ii) a portion of 10 CFR 50.46a(b); (iii) 10 CFR 50.49; (iv) 10 CFR 50.55(e); (v) certain requirements of 10 CFR 50.55a; (vi) 10 CFR 50.65, except for paragraph(a)(4); (vii) 10 CFR 50.72; (viii) 10 CFR 50.73; (ix) 10 CFR Part 50, Appendix B; (x) certain containment leakage testing requirements; and (xi) certain requirements of 10 CFR Part 100, Appendix A.

The current license conditions (Amendment Nos. 292 for Unit 1 and 320 for Unit 2) are as follows:

Duke Energy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, high winds, and external flood; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non PRA evaluations that are based on the IPEEE [individual plant examination of external events] Screening Assessment for External Hazards, i.e.,

seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards updated using the external hazard screening significance process identified in ASME [American Society of Mechanical Engineers]/ANS

[American Nuclear Society] PRA Standard RA-Sa-2009; as specified in Unit 1

[Unit 2] License Amendment No. 292 [320] dated September 17, 2019.

Duke Energy will complete the implementation items list in Attachment 1 of Duke letter to NRC dated April 8, 2019 prior to implementation of 10 CFR 50.69. All issues identified in the attachment will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG [Regulations Guide] 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 CFR 50.69 categorization process.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

The licensee proposed in the LAR to amend the above license conditions as follows:

Duke Energy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and high winds; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards updated using the external hazard screening process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in Unit 1 License Amendment No. [XXX] dated [DATE] [Unit 2 License Amendment No. [YYY] dated [DATE)).

Duke Energy will complete the implementation items listed in Attachment 1 of Duke letter to NRC dated November 24, 2020, prior to implementation of 10 CFR 50.69 in accordance with the categorization process described above.

The issues identified in the attachment will be addressed and any associated changes will be made prior to implementation of 10 CFR 50.69 in accordance with the categorization process described above.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g. a change from a seismic margins approach to a seismic probabilistic risk assessment approach).

The licensee also stated that the previously approved amendment included a license condition that stated that the licensee will complete the implementation items list in Attachment 1 of Duke Energys letter to the NRC (Reference 6) dated April 8, 2019, prior to implementation of 10 CFR 50.69. In the enclosure of this LAR (Reference 1) (page 5 of 23), the licensee stated that items ii and iii in Attachment 1 of Duke Energys letter to the NRC dated April 8, 2019 (Reference 6) were completed and the current LAR, if approved, would render item i in that attachment as no longer applicable.

The NRC staff reviewed this LAR for the licensees approach for screening the XF hazard in its previously approved 10 CFR 50.69 categorization process to assess the safety significance components in accordance with RG 1.201. In addition, the NRC staff utilized prior XF assessments from the Fukushima Dai-ichi Near-Term Task Force (NTTF) Recommendation 2.1, Flood Hazard Reevaluation results in this review (Reference 7) (Reference 8).

The supplements provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on September 8, 2020 (85 FR 55506).

2.0 REGULATORY EVALUATION

On November 22, 2004, the NRC added 10 CFR 50.69 to its regulations to address the risk-informed categorization and treatment of SSCs for nuclear power plants (Reference 9).

Implementation of 10 CFR 50.69 requires that licensees first categorize safety-related and non-safety-related SSCs according to their safety significance. SSCs are classified into high safety significant (HSS) and low safety significant (LSS) SSCs. Alternative treatments per 10 CFR 50.69(b)(1) and 10 CFR 50.69(d) can then be applied consistently with the categorization of the SSCs.

In May 2006, the NRC endorsed NEI 00-04 (Reference 3), Revision 0, with conditions, by RG 1.201 (Reference 10). NEI 00-04 describes in detail a process for determining the safety significance of SSCs and for categorizing them into the four RISC categories defined in 10 CFR 50.69. NEI 00-04 guidance allows licensees to implement different approaches, including non-PRA type approaches.

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities(Reference 11) endorses, with clarifications, the ASME/American Nuclear Society (ANS) PRA standard, ASME/ANS RA-Sa-2009 (Reference 12).

In the safety evaluation (SE) to License Amendments 292 (Unit 1) and 320 (Unit 2), dated September 17, 2019 (Reference 5), the NRC staff concluded that the licensee's process, as supplemented by the license conditions for that LAR, was consistent with the NRC-endorsed NEI 00-04 guidance and thus satisfies the requirements of 10 CFR 50.69(c). In the previously approved categorization process, the licensee used the XF PRA for the consideration of the external flooding risk in the categorization process. A license condition incorporated into the license as part of the NRC staffs decision states that NRC prior approval is required for a change to a categorization process that is outside the bounds specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment).

In the LAR (Reference 1), the licensee proposed the use of the guidance in Section 5.4 of NEI 00-04 to replace the use of the XF PRA for categorization of SSCs under the licensees previously approved 10 CFR 50.69 program. Section 5.4 of NEI 00-04 (Reference 3) provides a screening process for categorizing SSCs.

3.0 TECHNICAL EVALUATION

The NRC staff reviewed the proposed change to the previously-approved BSEP 10 CFR 50.69 program. In its LAR, the licensee stated that the XF hazard screening considers proposed modifications to the flood protection at BSEP as described in Duke Energy's September 27, 2018 (Reference 13) letter. Further, the licensee stated that screening of the XF hazard follows the guidance in NEI 00-04 and RG 1.201, Revision 1.

The licensee stated that, except for screening of the XF hazard, all other previously approved screening and categorization methods are not affected by this LAR. The NRC staffs review confirmed that the current LAR does not change any other aspect of the licensees categorization process except for the approach to consider the XF hazard (i.e., change from use of XF PRA to the guidance in Section 5.4 of NEI 00-04). Therefore, the NRC staffs review and decisions in the letter dated September 17, 2019 (Reference 5) on the licensees categorization process other than those related to the consideration of the XF hazard remain unchanged and valid. Consequently, the NRC staff did not separately review the licensees categorization process other than the change requested in the LAR.

As stated in RG 1.201, Revision 1, if a licensee wishes to change its categorization approach, the staffs review of the resulting submittal will focus on the acceptability of the methodology and analyses relied upon in the application. The following sections summarize the NRC staff's review of the acceptability of the proposed use of the guidance in Section 5.4 of NEI 00-04 for the XF hazard instead of an XF PRA into the previously-approved 10 CFR 50.69 categorization process.

3.1 Acceptability of the XF Hazard The licensee stated that BSEP XF hazards were re-evaluated following the accident at Fukushima Dai-ichi in accordance with present-day guidance, as documented in Dukes letter dated March 12, 2015 (Reference 14). This re-evaluation was performed in response to the NRCs request for information pursuant to 10 CFR 50.54(f) dated March 12, 2012 (Reference 15).

The BSEP flood hazard reevaluation report (FHRR) identified that several flood causing mechanisms exceeded the plant's original design basis by virtue of either not being considered

in the original design or exceeding the water elevations of mechanisms that were included. The FHRR was previously reviewed by the NRC in detail on April 16, 2018 (Reference 16); the NRC staff concluded that the licensee conducted the XF hazard reevaluation using present-day methodologies and regulatory guidance and that the FHRR reevaluation was supplemented with the adjustments to the surge hazards. The reevaluated XF hazard showed that certain XF causing mechanisms were not bounded by the plant licensing basis. Section 3.1.1, External Flood Hazard Summary of the enclosure to the LAR lists the XF causing mechanisms that could generate flood levels above floor elevations. The LAR cites the licensees re-evaluated XF hazard to support the proposed change. The NRC staff confirmed the list against the letter dated April 16, 2018 as well as the NRC staffs assessment of the licensees focused evaluation (FE) dated January 16, 2019 (Reference 17).

Because there are no changes to the approaches used to evaluate XF causing mechanisms and the corresponding results from the FHRR, the NRC staffs previous review and conclusion is applicable to this LAR. Thus, the NRC staff finds that the XF hazard and XF causing mechanisms used to support the proposed change are acceptable for this application.

3.2 Acceptability of the Proposed XF Screening Process to the LAR, Progressive Screening Approach for Addressing External Hazards, includes the screening criteria proposed by the licensee, and outlines the licensees implementation of the screening approach for various XF causing mechanisms. The NRC staff reviewed this screening approach against NEI 00-04, considered available information pertinent to the XF hazard, and the sites flood protection to determine the acceptability of the proposed screening criteria. The NRC staff finds that the licensees progressive screening criteria is the same as that in ASME/ANS RA-Sa-2009 for screening non-seismic external hazards.

The licensee explained that its FE contained details regarding the assessment of all unbound flood causing mechanisms at BSEP, justification for adequate available physical margin, the basis for determining that the flood response is appropriate, and proposed modifications to the plant to address the impacts from the re-evaluated flooding hazards. Section 3.1.2 of the enclosure to the LAR, Disposition of Mechanisms with No Impacts to the Site for 50.69 discusses the impact on the site from tsunami, failure of dams and onsite water control/storage structures, and flooding in streams and rivers. The licensee explained that these mechanisms were calculated to have a lower water surface elevation than the nominal plant grade. The licensee stated that these XF causing mechanisms can be screened because the damage potential is less than events for which the plant is designed. The licensee identified this screening basis as preliminary in the LAR. In its supplement dated November 24, 2020, the licensee explained that the screening basis was in fact the final criteria to be (Reference 2) used for tsunami, failure of dams and onsite water control/storage structures, and flooding in streams and rivers.

The NRC staffs assessment of the licensees FE concluded that the reevaluated flood level from tsunami, dam failure, and flooding in streams and rivers is lower than the site grade elevation and that no impacts to key SSCs were identified. The same staff assessment further states that BSEP relies on the passive protection of site topography to provide protection from these three XF causing mechanisms. Because there are no changes to the approaches used to evaluate XF causing mechanisms and the corresponding results from the FHRR, the NRC staffs previous review and conclusion for the licensees FE is still applicable to this LAR.

Therefore, the NRC staff finds that tsunami, failure of dams and onsite water control/storage

structures, and flooding in streams and rivers have damage potential less than events for which the plant is designed.

Section 3.1.3, Local Intense Precipitation Disposition of the enclosure to the LAR discusses the licensees proposed consideration of Local Intense Precipitation (LIP) in the approved 10 CFR 50.69 program. The licensee stated that each units Turbine Building and Reactor Building (RB) are affected by water accumulation from the LIP. The licensee explained that water infiltration at the Turbine Building was evaluated in the FE and was determined to not adversely impact any SSCs required for safe shutdown. The licensee further explained that LIP would produce standing water above only two door thresholds in the RB. The licensee stated that the in-leakage through these normally closed airlock doors was expected to be minimal and that the resulting impact was evaluated in the FE. The FE determined that a maximum depth of 0.3 feet (ft) of water could accumulate in the equipment access area behind door D-3. Also, the in-leakage would be handled via permanent passive floor drains and the RB sump system (the licensee took no credit for actively running sump pumps). Additionally, the height of lowest safety-related SSC in the RB is 1.0 ft off the finished floor, providing 0.7 ft of available physical margin.

The licensee stated that the LIP event can be screened utilizing the screening criteria where the event damage potential is less than that for which the plant was designed. Section 3.1.3 of the enclosure to the LAR discusses the initial screening criteria for LIP. In its supplement dated November 24, 2020 (Reference 2), the licensee explained that the screening basis was in fact the final criteria to be used for LIP. The licensee identified the failure of specific RB doors as resulting in an unscreened XF scenario for LIP. The licensee included those doors in a table provided in Section 3.1.5 of the enclosure to the LAR and explained that if these doors were to be categorized, they would be categorized as HSS in accordance with the guidance provided in NEI 00-04 Figure 5-6.

The conclusions on the impacts of LIP in the NRC staffs assessment of the licensees FE are consistent with the information provided by the licensee in Section 3.1.3 of the enclosure to the LAR. Because there are no changes to the approaches used to evaluate XF causing mechanisms and the corresponding results from the FHRR, NRC staffs previous review and conclusion for the licensees FE is still applicable to this LAR. Further, the licensee will use Figure 5-6 in NEI 00-04 for categorization of SSCs, such as the RB doors, impacted by LIP.

Therefore, the NRC staff finds that the licensees SSC categorization process will evaluate the safety significance of SSCs for the LIP mechanism consistent with the guidance provided in NEI 00-04, as endorsed by the NRC.

The licensee stated that the combined effects (CE) storm surge maximum still water elevation level and associated effects such as primarily wind-driven waves and tidal effects were estimated to produce flooding levels above the door thresholds throughout the site. The licensee explained that the plant has a combination of permanently installed water-tight doors and temporary barriers to provide protection from the CE storm surge event, as documented in the BSEP Updated Final Safety Analysis Report (UFSAR) (Reference 18). The licensee further explained that while the temporary barriers currently protect to an elevation of 26.0 ft, the height of these barriers will be increased to accommodate the impacts from the reevaluated CE storm surge event. Additionally, two of the doors in the Control Building will be replaced with permanently installed water-tight doors, eliminating the need for temporary barriers to be installed prior to the arrival of the CE storm surge event.

The licensee provided a list of flood barriers with proposed modifications in Attachment 1 of the supplement letter dated November 24, 2020, and stated that the modified flood barriers will provide adequate protection to the site during a CE storm surge event. The adequacy of these protective features was evaluated as part of the staffs assessment of the licensees FE. The licensees proposed license condition controls the completion of the modifications prior to the implementation of 10 CFR 50.69 at BSEP.

The licensee identified its procedure 0AI-68, Brunswick Nuclear Plant Response to Severe Weather Warnings, as establishing unambiguous procedural triggers, a clear organizational response, and a timeline for carrying out the actions required for the arrival of a CE storm surge event. According to the procedure and evaluation in the licensees FE, the site receives a warning for the storm at least 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> before arrival; the barriers must be installed at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to the arrival of a storm that could produce a surge of 20 feet or greater. The licensee stated that these actions were time validated and the total time required for installation was 15.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Additionally, the Units are placed in Mode 3 (hot shutdown) at least 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> before the storm is forecasted to arrive on site and preparations to move to Mode 4 (cold shutdown) continue after reaching Mode 3. By the time the storm arrives at the site, the Units will be in Mode 4 for the entire event duration. The NRC staffs assessment of the licensees FE stated that, because the licensee proposed modifications to the flood protection barriers, it is expected that Administrative Instruction 0Al-68 be modified to incorporate such changes.

However, the changes to the procedure were not discussed by the licensee in its LAR and a corresponding item was not present in Attachment 1 of the enclosure to the LAR. In its supplement dated November 24, 2020, the licensee stated that the timeline for installation of the flood barriers as defined in 0AI-68 is not expected to change significantly after the modifications to the flood protection barriers proposed in the FE and the LAR, and any change to the time is expected to decrease the barrier installation time. The licensee also provided a description of the overall action timeline in procedure 0AI-68 to confirm that modifications would not affect the sites response to the XF event. Based on the supplemental information and the conclusions in the prior staffs assessment to the licensees FE, the NRC staff concludes that the licensees timeline for implementation of 0AI-68 will not be negatively impacted by the proposed modifications to the flood barriers and no additional human factors review is necessary.

In its supplement dated November 24, 2020 (Reference 2), the licensee explained that the re-evaluated flood hazard constitutes part of the BSEP design basis and provides significant margin in comparison to the licensing basis flood elevation described in the UFSAR. The licensee further stated that, once the plant modifications in Attachment 1 of the November 24, 2020, supplement are completed, these features will provide physical protection to the site such that the event damage potential (post-modification) will also be less than the event for which the plant was initially designed for (pre-modification).

Section 3.1.4, Combined Effects Storm Surge Disposition of the enclosure to the LAR discusses the preliminary initial screening criteria for the CE storm surge mechanism. As discussed above, in its supplement dated November 24, 2020, the licensee explained that the screening criteria in the LAR are the final criteria for the CE storm surge event.

The licensee stated that the list of flood barriers with proposed modifications in Attachment 1 of the enclosure to the LAR was a subset of the complete list of commitments presented in of the FE (Items #1-4, and 6). The licensee explained that the proposed modifications to the smoke removal vents on the Fuel Oil Tank Chamber (FOTC) roof (Item #5, of the FE) are not required for implementation of 10 CFR 50.69 because the lowest elevations of these vents are above the maximum reevaluated still water elevation for the CE

storm surge. However, the licensees FE states that the increase in the elevation of the smoke removal vents is to provide permanent passive protection from the CE storm surge event.

Further, the FE also provides the basis for installing a new debris barrier to protect all penetrations on the FOTC roof from debris loading associated with the CE storm surge event.

In its supplement dated November 24, 2020, the licensee stated that debris and hydrodynamic loading from the CE storm surge on the FOTC smoke removal vents and other FOTC components are addressed by installation of a permanently installed debris barrier. The debris barrier is credited in this screening evaluation and included in the proposed modifications in of the letter dated November 24, 2020. The licensees proposed license condition controls the completion of the modifications prior to the implementation of 10 CFR 50.69 at BSEP. In addition, the licensee stated that raising the FOTC smoke removal vents is only related to increasing the available physical margin of the vents above the maximum reevaluated still water elevation for the CE storm surge. Based on its review, the NRC staff finds that raising the FOTC smoke removal vents does not impact the proposed change.

Based on its review, the NRC staff finds that there are no changes to the approaches used to evaluate CE storm surge and the corresponding results from the FHRR, and therefore, NRC staffs previous review and conclusion for CE storm surge in the licensees FE is applicable to this LAR. The NRC staffs review also finds that (1) the proposed modifications provide physical protection to the site from the effects of CE storm surge; (2) the licensees procedure provides reasonable assurance of timely response to CE storm surge; (3) the licensee proposed a license condition that controls the completion of the proposed modifications for protection from CE storm surge effects prior to the implementation of 10 CFR 50.69 at the site; and (4) the licensee will use the guidance in NEI 00-04 for categorization of SSCs for the CE storm surge.

Further, the NRC staff finds that the licensees SSC categorization process will evaluate the safety significance of SSCs for the CE storm surge mechanism consistent with the guidance provided in NEI 00-04, as endorsed by the NRC.

In its supplement dated November 24, 2020, the licensee stated that there are no SSCs uniquely modeled in the XF PRA, or that are not included in the internal events and/or internal fire PRA. The licensee also stated that the flood protection features credited in the FE and LAR are the SSCs needed to mitigate the XF hazard. Section 3.1.5 of the enclosure to the LAR, SSCs Required for External Flood Screening, provides a list of SSCs that if failed would result in an unscreened XF scenario for the CE storm surge event. In its supplement dated November 24, 2020, the licensee stated that the list in Section 3.1.5 of the LAR was intended to be only a representative list based on the identified modifications. The licensee also stated that when a system is categorized, the components of that system will be evaluated against the XF hazard scenario and any component whose failure would cause the scenario to become unscreened will be assessed as HSS. This ensures that modeled components will continue to be evaluated per the NRC-approved process. Components assessed as HSS for non-modeled hazards will remain HSS and cannot be over-ridden, and changed to LSS, by the Integrated Decision-Making Panel (IDP). The NRC staffs review finds that the proposed approach is consistent with the guidance in NEI 00-04, as endorsed by the NRC.

Item (v) in Attachment 1 of the enclosure to the LAR stated that BSEP will confirm all barriers conform to the requirements for flood protection features specified in NEI 16-05 [External Flooding Assessment Guidelines (Reference 19)] including Appendix B Section 3.1.4, Combined Effects Storm Surge Disposition of the enclosure to the LAR states that the operator response, warning time and time required for installation were evaluated against the NRC endorsed NEI 16-05 Appendix C. NEI 16-05, Revision 1 was endorsed by NRCs JLD-ISG-2016-1, Guidance for Activities Related to Near-Term Task Force Recommendation 2.1,

Flooding Hazard Reevaluation Focused Evaluation and Integrated Assessment (Reference 20),

describes the flooding impact assessment process for use by licensees to close out the 10 CFR 50.54(f) request for information arising from Near-Term Task Force Recommendation 2.1. In its supplement dated November 24, 2020, the licensee stated that NEI 16-05 provides a systematic evaluation technique to assess the impact of XF hazards and has become the industry practice for evaluating the vulnerabilities and response to XF hazard impacts at nuclear power plants. The licensee also stated that NEI 16-05 systematic process uses standards appropriate for reliable performance to meet the credited mitigation and protection function. The licensee also stated that these standards are consistent with design and timing analysis used for other PRA activities and are appropriate for crediting in screening the XF hazard for the 10 CFR 50.69 program. As documented in NRC staff assessment of the licensees FE, the licensee demonstrated that effective flood protection, if appropriately implemented, exists at the site, and followed the guidance in NEI 16-05, Revision 1 to implement its flood protection strategies. Based on that evaluation and the reasons provided in the supplement dated November 24, 2020, the NRC staff finds the licensees use of NEI 16-05, Revision 1, is consistent with prior NRC staff assessments and thus acceptable for evaluation of plant modifications for this application.

3.3 Summary of Technical Evaluation The NRC staff reviewed the licensees approach for screening the XF hazards in its previously approved 10 CFR 50.69 categorization process to assess the safety significance components.

The NRC staffs review finds the proposed approach to be acceptable and consistent with RG 1.201, Revision 1, and the NRC-endorsed guidance in NEI 00-04. The NRC staff finds that the use of the proposed approach for consideration of XF risk in the licensees LAR is acceptable for the licensees 10 CFR 50.69 categorization process. Thus, the NRC staff finds that implementation item i in Attachment 1 of Duke Energys letter to the NRC dated April 8, 2019, is no longer applicable because it was related to the XF PRA, which will not be used for categorization.

In the enclosure to its letter dated July 9, 2020 (page 5 of 23), the licensee stated that items ii and iii in Attachment 1 of Duke Energys letter to the NRC dated April 8, 2019 were completed.

As noted in its SE dated September 17, 2019 (Reference 5), the completion of those items by the licensee did not change or impact the bases for the safety conclusions made by the NRC staff, and through an onsite audit or future inspections, the NRC staff may choose to examine the closure of the implementation items. Thus, the NRC staff finds that items ii and iii in of Duke Energys letter to the NRC dated April 8, 2019 can be removed from the list of implementation items.

The NRC staff has reviewed the licensees proposed change to its previously approved 10 CFR 50.69 categorization process. Use of the guidance in Section 5.4 of NEI 00-04 for external flooding risk in the licensee's categorization process, in conjunction with the license condition stated in Section 1.0 of this SE, is consistent with the NRC-endorsed NEI 00-04 guidance and thus satisfies the requirements of 10 CFR 50.69(c). Based on this evaluation, the NRC staff finds that the proposed license condition and its referenced implementation items are acceptable because they adequately implement 10 CFR 50.69 using models, methods, and approaches consistent with the applicable guidance that has previously been endorsed as acceptable by the NRC. With the exception of the approach for XF, the licensees use of other PRA and non-PRA approaches approved by the September 17, 2019 (Reference 5) letter in the

10 CFR 50.69 categorization process to assess the safety significance of active and passive components will remain unchanged.

3.4 Periodic Review The NRC staff, through an onsite audit or during future inspections, may choose to examine the closure of the implementation items with the expectation that any variations discovered during this review, or concerns regarding adequate completion of the implementation item, would be tracked and dispositioned appropriately under the licensee's corrective action program, and could be subject to appropriate NRC enforcement action, as completion of the implementation items would be required by the proposed license conditions.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the NRC staff notified the North Carolina State official on February 23, 2021, of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, published in the Federal Register on September 8, 2020 (85 FR 55506), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

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

7.0 REFERENCES

1 Duke Energy, License Amendment Request to Modify Approved 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, Agencywide Documents Access and Management System (ADAMS) Accession No. ML20191A054, July 9, 2020.

2 Duke Energy, Response to Request for Additional Information (RAI) for License Amendment Request to Modify Approved 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Plant," (ADAMS Accession No. ML20329A466), November 24, 2020.

3 Nuclear Energy Institute, NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline (ADAMS Accession No. ML052910035), July 31, 2005.

4 U.S. Nuclear Regulatory Commission, Regulatory Guide (RG) 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance, for Trial Use," (ADAMS Accession No. ML061090627), April 28, 2006.

5 U.S. Nuclear Regulatory Commission, Brunswick, Units 1 and 2 - Issuance of Amendment Nos. 292 and 320, Adopt 10 CFR 50.69 "Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors" (EPID L-2018-LLA-0008) (ADAMS Accession No. ML19149A471), September 17, 2019.

6 Duke Energy, Response to NRC Request for Additional Information (RAI) Regarding Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures,Systems, and Components (SSCs) for Nuclear Power Reactors, (ADAMS Accession No. ML19099A035), April 8, 2019.

7 U.S. Nuclear Regulatory Commission, Brunswick Steam Electric PLant Units 1 AND 2 -

Staff Assessment of Response to 10 CFR 50.54(f) Information Request Flood-causing Mechanism Reevaluation (ADAMS Accession No. ML18089A055), April 16, 2018.

8 US Nuclear Regulatory Commission, Brunswick Steam Electric PLant Units 1 AND 2 - Staff Assessment of Flooding Focused Evaluation (ADAMS Accession No. ML18348B185),

January, 16, 2019.

9 U.S. Nuclear Regulatory Commission, 10 CFR 50.69, (Federal Register Notice 69 FR 68008), November 22, 2004.

10 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.201, Revision 1, Guidelines for Categorizing Sturctures, Systems, and Components in Nuclear Power Plants according to Their Safety Significance, (ADAMS Accession No. ML061090627), May 2006.

11 U.S. Nuclear Regulatory Commission, RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (ADAMS Accession No. ML090410014), March 1, 2009.

12 ASME/ANS, ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," Addendum A to RA-S-2008, February 2009.

13 Duke Energy, Brunswick Steam Electric Plant, Units 1 and 2, Response to March 12, 2012, Request for Information Enclosure 2 Recommendation 2.1, Flooding, Required Response 3, Flooding Focused Evaluation Summary Report, (ADAMS Accession Number ML18270A372), September 27, 2018.

14 Duke Energy, Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Flood Hazard Reevaluation Report, Response to NRC 10 CFR 50.54(f) Request for Information

Regarding Recommendations 2.1, 2.3 and 9.3 of the NEEF Review, (ADAMS Accession No. ML15079A385; non-public), March 12, 2015.

15 U.S. Nuclear Regulatory Commission, Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, (ADAMS Accession Number ML12053A340), March 12, 2012.

16 Duke Energy, Brunswick Steam Electric Plant Units 1 And 2 - Staff Assessment of Response To 10 CFR 50.54(F) Information Request Flood-Causing Mechanism Reevaluation (ADAMS Accession No. ML18086B575; non-public), April 16, 2018.

17 Duke Energy, Brunswick Steam Electric Plant Units 1 and 2 - Staff Assessment of Flooding Focus Evaluation, (ADAMS Accession No. ML18348B185), January 16, 2019.

18 Duke Energy, Submittal of Updated Final Safety Analysis Report (UFSAR), Revision 27, (ADAMS Accession No. ML20260H26), August 12, 2020.

19 Nuclear Energy Institute, NEI 16-05 "External Hazard Assessment Guidelines," (Accession No. ML16165A178), June 2016.

20 U.S. Nuclear Regulatory Commission, JLD-ISG-2016-01, "Guidance for Activities Related to Near-Term Task Force Recommendation 2.1, Flooding Hazard Reevaluation," (ADAMS Accession No. ML16162A301), June 2016.

21 National Fire Protection Association, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," Standard 805 (NFPA 805), 2001 Edition, Quincy, Massachusetts.

22 Gideon, William R., Duke Energy Progress, LLC, letter to U.S. Nuclear Regulatory Commission, "Brunswick Steam Electric Plant, Unit Nos. 1 and 2, Renewed Facility Operating License Nos. DPR-71 and DPR-62, Docket Nos. 50-325 and 50-324, Request for License Amendment for Performance-Based Fire Protection, Alternative for Thermal Insulation Material," November 15, 2017 (ADAMS Accession No. ML17331A484).

23 U.S. Nuclear Regulatory Commission, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Regulatory Guide 1.205, Revision 1, December 2009 (ADAMS Accession No. ML092730314).

24 Nuclear Energy Institute, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c)," Washington, DC, NEI 04-02, Revision 2, April 2008 (ADAMS Accession No. ML081130188).

25 U.S. Nuclear Regulatory Commission, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,"

Regulatory Guide 1.174, Revision 2, May 2011 (ADAMS Accession No. ML100910006).

26 U.S. Nuclear Regulatory Commission, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, Volume 2: Detailed Methodology," NUREG/CR-6850, September 2005 (ADAMS Accession No. ML052580118).

27 U.S. Nuclear Regulatory Commission, "Fire Probabilistic Risk Assessment Methods Enhancements," NUREG/CR-6850, Supplement 1, September 2010 (ADAMS Accession No. ML103090242).

28 Annacone, Michael J., Carolina Power and Light Company, letter to U.S. Nuclear Regulatory Commission, "Brunswick Steam Electric Plant, Unit Nos. 1 and 2, Docket Nos.

50-325, 50-324, License Amendment Request to Adopt NFPA 805 Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition)," September 25, 2012 (ADAMS Accession No. ML12285A428).

29 Gideon, William R., Duke Energy Progress, LLC, letter to U.S. Nuclear Regulatory Commission, "Brunswick Steam Electric Plant, Unit Nos. 1 and 2, Renewed Facility Operating License Nos. DPR-71 and DPR-62, Docket Nos. 50-325 and 50-324, Request for License Amendment for Performance-Based Fire Protection, Alternative for Thermal Insulation Material," May 23, 2018 (ADAMS Accession No. ML18143B743).

30 U.S. Nuclear Regulatory Commission, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, Volume 1: Summary and Overview," NUREG/CR-6850, September 2005 (ADAMS Accession No. ML052580075).

31 American Society of Testing Materials, "Standard Test for Surface Characteristics of Building Materials," Standard E-84 (ASTM E-84), West Conshohocken, Pennsylvania.

32 Duke Energy, Brunswick Steam Electric Plant, Unit 1 and 2 Flood Hazard Reevaluation Report, Response to NRC 10 CFR 50.54(f) Request for Information Regarding Recommendations 2.1, 2.3 and 9.3 of the NTTF Review, (ADAMS Accession No. ML15079A385; non-public), March 11, 2015.

33 U.S. Nuclear Regulatory Commission, Brunswick Steam Electric Plant Units 1 and 2 - Staff Assessment of Flooding Focus Evaluation, (ADAMS Accession No. ML18348B185),

January 16, 2019.

34 U.S. Nuclear Regulatory Commission, Brunswick Steam Electric Plant Units 1 And 2 - Staff Assessment of Response To 10 CFR 50.54(F) Information Request Flood-Causing Mechanism Reevaluation (ADAMS Accession No. ML18086B575; non-public), April 16, 2018.

35 U.S. Nuclear Regulatory Commission, letter to William R. Gideon, Brunswick Steam Electric Plant, "Brunswick Steam Electric Plant, Units 1 and 2 - Issuance of Amendment Regarding Transition to a Risk-Informed, Performance-Based, Fire Protection Program in Accordance with 10 CFR 50.48(c) (TAC Nos. ME9623 and ME9624)," January 28, 2015 (ADAMS Accession No. ML14310A808).

Principal Contributors: Shilp Vasavada, NRR/DRA Milton Valentin-Olmeda, NRR/DRA Date: April 30, 2021

ML21067A224 OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DRA/APLA/BC NAME AHon RButler SRosenberg DATE 03/05/2021 03/16/2021 01/29/2021 OFFICE OGC - NLO NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM NAME JMcManus DWrona AHon DATE 04/06/2021 04/30/2021 04/30/2021