RA-19-0437, Nos. 1 & 2 - License Amendment Request to Modify Approved 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Categorization Process

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Nos. 1 & 2 - License Amendment Request to Modify Approved 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Categorization Process
ML20191A054
Person / Time
Site: Brunswick  Duke Energy icon.png
Issue date: 07/09/2020
From: Krakuszeski J
Duke Energy Progress
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-19-0437
Download: ML20191A054 (26)
v • d • e


Text

John A. Krakuszeski Vice President Brunswick Nuclear Plant 8470 River Rd SE Southport, NC 28461 o: 910.832.3698 RA-19-0437 10 CFR 50.90 July 09, 2020 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-325 AND 50-324 / RENEWED LICENSE NOS. DPR-71 AND DPR-62

SUBJECT:

License Amendment Request to Modify Approved 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Categorization Process

REFERENCES:

1. US NRC Letter to Duke Energy, Brunswick Steam Electric Plant, Units 1 and 2 - issuance of Amendment Nos. 292 and 320 to 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 Number ML19149A471, September 17, 2019.

2. Duke Energy Letter to US NRC, Brunswick Steam Electric Plant (BSEP),

Unit Nos. 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.

In accordance with the license conditions added to the Renewed Facility Operating Licenses for Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2 upon issuance of Amendments 292 and 320 (Reference 1) and pursuant to the provisions of 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy), is requesting a change to the 10 CFR 50.69, "Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors" categorization process.

Specifically, the proposed amendments would revise the 10 CFR 50.69 license conditions added upon issuance of Amendments 292 and 320 (Units 1 and 2, respectively) (Reference

1) to reflect the screening of the external flood (XF) hazard.

The XF hazard screen is based on proposed modifications to the flood protection system at BSEP as described in Duke Energy's letter to the NRC dated September 27, 2018 (Reference 2). The license conditions imposed on BSEP in Reference 1 indicate that a XF probabilistic risk assessment (PRA) model must be used in the 10 CFR 50.69 categorization process.

However, based on enhanced plant protection, the XF hazard has been screened and therefore a detailed XF PRA model is no longer necessary for the categorization process. No other changes to the categorization process are being requested by this license amendment request.

RA-19-0437 Page 2 The enclosure provides a description and assessment of the proposed change. The categorization process being implemented through this change remains consistent with the original BSEP 10 CFR 50.69 Safety Evaluation Report (Reference 1) and NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision O dated July 2005, which was endorsed by the NRC in Regulatory Guide 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," Revision 1 dated May 2006. Attachment 1 of the enclosure provides a list of categorization prerequisites. Use of the categorization process on a plant system will only occur after these prerequisites are met.

Attachment 2 provides the External Hazards screening pertaining to the XF PRA model.

Attachment 3 is the Progressive Screening Approach for Addressing External Hazards.

Attachment 4 is a proposed markup of the current BSEP Facility Operating License for Units 1 and 2, which contains the current issued 10 CFR 50.69 Amendments.

Duke Energy requests approval of the proposed license amendments within one year of completion of the NRC's acceptance review. Once approved, the amendments shall be implemented within 120 days.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated North Carolina Official.

If you should have any questions regarding this submittal, please contact Art Zaremba, Manager

- Fleet Licensing at (980) 373-2062.

I declare under penalty of perjury that the foregoing is true and correct. Executed on July 09, 2020.

Sincerely, John A. Krakuszeski Site Vice President - Brunswick Steam Electric Plant

Enclosure:

1. Evaluation of the Proposed Change Attachments:
1. Evaluation of the Proposed Change
2. Eternal Hazards Screening
3. Progressive Screening Approach for Addressing External Hazards
4. Markup of Facility Operating License Conditions (Appendix B of FOL)

RA-19-0437 Page 3 Cc (with enclosure):

L. Dudes, Regional Administrator U. S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257 A.L. Hon, Project Manager (BSEP) (Electronic Copy only)

U. S. Nuclear Regulatory Commission One White Flint North, Mail Stop 8 G9A 11555 Rockville Pike Rockville, MD 20852-2738 Andrew.Hon@nrc.gov G. Smith, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 Chair - North Carolina Utilities Commission (Electronic Copy only) 4325 Mail Service Center Raleigh, NC 27699-4300 swatson@ncuc.net W. Lee Cox III, Section Chief (Electronic Copy Only)

Radiation Protection Section North Carolina Department of Health and Human Services 1645 Mail Service Center Raleigh, NC 27699-145 lee.cox@dhhs.nc.gov

RA-19-0437 Page 1 of 23 Enclosure Evaluation of the Proposed Change TABLE OF CONTENTS 1

SUMMARY

DESCRIPTION .................................................................................................. 3 2 DETAILED DESCRIPTION ................................................................................................... 3 2.1 CURRENT REGULATORY REQUIREMENTS ............................................................ 3 2.2 REASON FOR PROPOSED CHANGE ........................................................................ 3

2.3 DESCRIPTION

OF THE PROPOSED CHANGE ......................................................... 3 3 TECHNICAL EVALUATION .................................................................................................. 4 3.1 EXTERNAL FLOOD HAZARD SCREENING TECHNICAL EVALUATION .................. 5 3.1.1 External Flood Hazard Summary ................................................................... 5 3.1.2 Disposition of Mechanisms with No Impacts to the Site for 50.69 .................. 6 3.1.3 Local Intense Precipitation Disposition for 50.69 ........................................... 7 3.1.4 Combined Effects Storm Surge Disposition for 50.69 .................................... 7 3.1.5 SSCs Required for External Flood Screening ................................................ 8 4 REGULATORY EVALUATION ........................................................................................... 10 4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA .................................... 10 4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS ................................... 10

4.3 CONCLUSION

S ......................................................................................................... 12 5 ENVIRONMENTAL CONSIDERATION .............................................................................. 13 6 REFERENCES ................................................................................................................... 14

RA-19-0437 Page 2 of 23 LIST OF ATTACHMENTS : List of Categorization Prerequisites .................................................................... 16 : External Hazards Screening ................................................................................ 17 : Progressive Screening Approach for Addressing External Hazards.................... 18 : Markup of Facility Operating License Conditions (Appendix B of FOL) ............... 19

RA-19-0437 Page 3 of 23 1

SUMMARY

DESCRIPTION The 10 CFR 50.69 categorization process has been reviewed and approved by the US Nuclear Regulatory Commission (NRC) for Brunswick Steam Electric Plant (BSEP) by letter dated September 17, 2019 [Reference 1]. This proposed amendment revises the license condition that required using the BSEP XF PRA model in the categorization process to allow an XF screening evaluation based on proposed modifications which would allow the plant to mitigate the effects of all XF hazards rather than allowing water to enter into structures containing safety-related (SR) structures, systems or components (SSCs). All other aspects of the program remain as the NRC approved in Reference 1.

2 DETAILED DESCRIPTION 2.1 CURRENT REGULATORY REQUIREMENTS The approved BSEP 50.69 categorization process conforms to the guidance in NRC RG 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," Revision 1 dated May 2006 [Reference 3]. The categorization process also conforms to the guidance in NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision 0 dated July 2005 (Reference 4), as endorsed by RG 1.201. With this proposed change to screen the XF scenarios, the BSEP categorization process will continue to conform to these guidance documents. With this proposed change the BSEP categorization process also remains consistent with the previously approved program in Reference 1.

2.2 REASON FOR PROPOSED CHANGE The BSEP 10 CFR 50.69 categorization process has previously been reviewed and approved by NRC (Reference 1). The proposed change implements a modification to the process, as allowed by the 10 CFR 50.69 guidance endorsed by NRC in Regulatory Guide 1.201

[Reference 3], by screening the XF scenarios listed in section 3. The justification is based on plant modifications which will either eliminate certain XF mechanisms from affecting safety related SSCs or will allow adequate time to mitigate of the effects of remaining XF mechanisms.

2.3 DESCRIPTION

OF THE PROPOSED CHANGE Duke Energy proposes to revise the 10 CFR 50.69 license conditions that were added with BSEP Amendments 292 and 320. A markup describing the proposed revisions to the amendments can be found in Attachment 4.

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

RA-19-0437 Page 4 of 23 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 [#] License Amendment No. [XXX] dated [DATE].

Duke Energy will complete the implementation items listed in Attachment 1 of Duke letter to NRC dated [DATE] 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).

3 TECHNICAL EVALUATION 10 CFR 50.69 specifies the information to be provided by a licensee requesting adoption of the regulation. This request conforms to the requirements of 10 CFR 50.69(b)(2), which states:

A licensee voluntarily choosing to implement this section shall submit an application for license amendment under § 50.90 that contains the following information:

(i) A description of the process for categorization of RISC-1, RISC-2, RISC-3 and RISC-4 SSCs.

(ii) A description of the measures taken to assure that the quality and level of detail of the systematic processes that evaluate the plant for internal and external events during normal operation, low power, and shutdown (including the plant-specific probabilistic risk assessment (PRA), margins-type approaches, or other systematic evaluation techniques used to evaluate severe accident vulnerabilities) are adequate for the categorization of SSCs.

(iii) Results of the PRA review process conducted to meet § 50.69(c)(1)(i).

(iv) A description of, and basis for acceptability of, the evaluations to be conducted to satisfy § 50.69(c)(1)(iv). The evaluations must include the effects of common cause interaction susceptibility, and the potential impacts from known degradation mechanisms for both active and passive functions, and address internally and externally initiated events and plant operating modes (e.g., full power and shutdown conditions).

The above information was previously provided to the NRC on January 10, 2018 as part of the BSEP LAR application to implement the 10 CFR 50.69 categorization process [Reference 2].

This current proposed amendment would only incorporate screening of the external flood (XF) hazard based on proposed modifications to the flood protection system at BSEP as described in Duke Energy's letter to the NRC dated September 27, 2018 (ADAMS Accession Number ML18270A372). The 10 CFR 50.69 LAR approved on September 17, 2019 stated that an External Flooding (XF) Probabilistic Risk Assessment (PRA) model would be used in the 10 CFR 50.69 categorization process. The proposed change is limited to the change in the original

RA-19-0437 Page 5 of 23 approved LAR from using a detailed XF PRA model to screening the XF hazard based on modified plant protection.

The conditions specified in the amendments provided in Reference 1 state, "Duke Energy will complete the implementation items list in Attachment 1 of Duke letter to the NRC dated April 8, 2019 prior to implementation of 10 CFR 50.69." These implementation items are listed below and are found in Attachment 1 to Reference 15.

Brunswick 50.69 PRA Implementation Items Descriotion Resolution

i. The BSEP external flood (XF) model Duke Energy will complete a focused hazard is being updated with more scope peer review of the BSEP External detailed analytical modeling as Flood PRA model hazard development described in response to RAI 11 in Duke prior to implementation of 10 CFR 50.69.

Any findings from the focused scope peer Energy letter dated November 2, 2018. review will be resolved and closed per an The additional details need a focused NRC approved process prior to scope peer review. implementing 10 CFR 50.69.

ii. The BSEP FLEX diesel generator (DG) Duke Energy will update the applicable failure rates will be updated using plant- PRA models with FLEX DG failure rates specific data as described in Attachment as described in Attachment 3 of Duke 3 of Duke Energy letter dated April 8, Energy letter dated April 8, 2019 prior to implementing 10 CFR 50.69.

2019.

iii. The BSEP LERF model is being The operator actions and associated updated with additional containment equipment failures modeling containment venting modeling as described in venting will be added to the BSEP LERF response to RAI 4.01 I 17.01 in Duke model as described in response to RAI 4.01 I 17.01 in Duke Energy letter dated Energy letter dated April 8, 2019. April 8, 2019 prior to implementing 10 CFR 50.69.

Items "ii" and "iii" in the above table have been completed. Screening the XF model will render item "i" no longer applicable.

Except screening of the XF hazard, all other previously approved categorization methods described in Reference 1, are not affected by this amendment request. Screening of external hazards follows the guidance set forth in NEI 00-04 and RG 1.201 and therefore, the remainder of this technical evaluation is focused on establishing the technical adequacy of the BSEP XF screening methodology for this application.

3.1 EXTERNAL FLOOD HAZARD SCREENING TECHNICAL EVALUATION 3.1.1 External Flood Hazard Summary The BSEP XF hazards were reevaluated following the accident at Fukushima Daiichi in accordance with present-day guidance provided in NUREG 0800 and various flood mechanism specific guidance documents. BSEP performed a reevaluation of the flood hazard at the site to respond to the Nuclear Regulatory Commission's (NRC) Request for Information pursuant to 10 CFR 50.54(f) [Reference 5]. The flood hazard reevaluation (FHR) was completed on 03/11/2015 [Reference 6] and submitted to NRC for review. It was determined that several flood causing mechanisms exceeded the plant's original design basis by both virtue of not being considered in the original design or exceeding the water elevations of mechanisms that were included. The NRC concurred with the results of the FHR and issued a Staff Assessment (SA)

RA-19-0437 Page 6 of 23 letter dated April 16, 2018 [Reference 7] to document the parameters for use in future applications.

BSEP performed a Focused Evaluation (FE) [Reference 8] submitted to NRC on September 27, 2018. The evaluation contains details regarding the assessment of all unbound flood causing mechanisms at BSEP, justification for adequate available physical margin (APM), the basis for determining that the flood response is appropriate and proposed modifications to the plant to address the impacts from the reevaluated hazards. NRC issued a Staff Assessment (SA) letter concurring with the BSEP assessment in the FE on January 16, 2019 [Reference 9].

It was determined that certain XF mechanisms were not bounded by the original design basis (DB) and should be further evaluated for impacts to the site. The table below shows the mechanisms, the original DB, and the reevaluated still water elevation. This information can also be found in the FE. For reference, the nominal plant grade at both Units is 20.0 ft (NGVD29).

Flood Causing Mechanisms Considered in the Original Design and Beyond Design Basis of BNP Original Design Basis Beyond Design Basis Flood Causing Mechanism and Still Water Elevation (ft Still Water Elevation (ft Plant Location NGVD29) NGVD29)

Tsunami Not Included 10.21 Failure of Dams and Onsite Water Not Included 15.53 Control/Storage Structures Streams and Rivers Not Included 15.46 Local Intense Precipitation (Note 1) Not Included

a. Reactor and Control Building a. 21.07
b. Turbine Building b. 21.69 Combined Effects Probable 22 ft (Still Water (Still Water Elevation)

Maximum Storm Surge (CE PMSS) Elevation)

(Note 2) a. 26.7

a. Reactor Building b. 26.7
b. Diesel Generator Building c. 26.6
c. Service Water Building/Intake d. 26.6 Structure e. 26.9
d. AOG Building
e. Turbine Building Note 1 - Locations for LIP Elevations reported only include those locations where the water surface elevation exceeds the door thresholds. All other locations at the plant were not calculated to have water above the door thresholds.

Note 2 - All doors subjected to flooding are not directly impacted by wind driven waves as they are interior to other doors. The exception is DG Building north external personnel door and this was modified to have debris protection exterior to the door/building.

3.1.2 Disposition of Mechanisms with No Impacts to the Site for 50.69 The following mechanisms are considered screened per the initial preliminary screening criteria C1, "Event damage potential is less than events for which plant is designed" (Attachment 3).

RA-19-0437 Page 7 of 23 The mechanisms were calculated to have a lower water surface elevation (WSE) than the nominal plant grade of 20.0 ft (NGVD29) and do not have any postulated impacts based on the results of the FHR. These flood causing mechanisms, shown in the table above, were evaluated as part of the post-Fukushima activities and reviewed by NRC staff.

1. Tsunami
2. Failure of Dams and Onsite Water Control/Storage Structures
3. Flooding in Streams and Rivers 3.1.3 Local Intense Precipitation Disposition For the Local Intense Precipitation (LIP) event, water collects and ponds around the plant and produces a still water elevation level (SWEL) as shown in the table, above. Two buildings on site are affected by water accumulation from the LIP event, the Turbine (TB) and Reactor Buildings (RB). Water infiltration at the TB was evaluated in the FE and concluded that it will not adversely impact any system, structures or components (SSCs) required for safe shutdown.

The impact from ponding water around the RB doors was also evaluated in the FE. The equipment access airlock door D-2 and the personnel access airlock door D-3 have a finished floor elevation of 20 ft. The maximum WSE against those doors are 20.79 ft and 21.07 ft, respectively. The in leakage through the these normally closed airlock doors is expected to be minimal. The conclusion was that a maximum depth of 0.3 ft of water could accumulate in the equipment access area behind door D-3 that will be handled via permanent passive floor drains and the Reactor Building sump system (no credit for actively running sump pumps).

Additionally, the height of lowest SR SSC in the Reactor Building off the finished floor is 1.0 ft providing 0.7 ft of available physical margin (APM). BSEP calculation BNP-MECH-FHR-001

[Reference 10] provides detailed evaluation of the LIP event and associated in leakage.

Based on the evaluations in BNP-MECH-FHR-001 and the FE, the LIP event can be screened utilizing the initial screening criteria C1 (Attachment 3) where the event damage potential is less than that for which the plant was designed. However, the failure of the reactor building doors would result in an unscreened XF scenario for the LIP event. If the doors listed in the table below were to be categorized, they would be categorized as high safety significance (HSS) in accordance with the guidance provided in NEI 00-04 Figure 5-6. These components would not be categorized as low safety significance (LSS) unless adequate justification is provided to demonstrate that their failure would not result in an unscreened scenario in accordance with NEI 00-04 Figure 5-6.

3.1.4 Combined Effects Storm Surge Disposition The combined effects (CE) storm surge maximum Still Water Elevation Level (SWEL) and associated effects (primarily wind-driven waves and tidal effects) are estimated to produce flooding levels above the door thresholds throughout the site, as shown in the table above. As part of the BSEP plant design documented in the Updated Final Safety Analysis Report (UFSAR) and governed by procedure 0AI-68 [Reference 12], the plant has a combination of permanently installed water-tight doors and temporary barriers to provide protection from the CE storm surge event. The temporary barriers currently protect to an elevation of 26.0 ft, however, the FE outlines commitments to increase the height of these barriers 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 watertight doors, thus removing the requirement for temporary barriers to be installed prior to the arrival of the CE storm surge event.

RA-19-0437 Page 8 of 23 Procedure 0AI-68 establishes unambiguous procedural triggers, a clear organizational response, and provides a detailed timeline for carrying out the actions required for the arrival of a CE storm surge event. According to the procedure and evaluation in the 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 ft or greater. The actions were time validated in the procedure and the total time required for installation is 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 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 continue after reaching Mode 3.

By the time the storm arrives at the site, the Units will be in Mode 4 (cold shutdown) for the entire event duration.

Hurricanes are slow developing natural phenomena, allowing sufficient time to provide an adequate response by plant personnel to mitigate the threat. Given that time requirements to complete these actions are governed by procedure 0AI-068 and require all the actions to be taken well in advance of the onslaught of adverse weather, it was concluded that the site response is adequate with more than twice the time margin required to re-perform actions in the event that barrier installation is not completed correctly during the first attempt [Reference 8].

These conclusions were documented in the NRCs SA of the FE [Reference 9].

Based on the evaluation performed in the FE and conclusions of the Staff Assessment to the FE, the CE storm surge event is considered screened utilizing the initial preliminary screening criteria C1 and C5 (Event develops slowly, allowing adequate time to eliminate or mitigate the threat) (see Attachment 3). The flood barriers (with proposed modification outlined in ) will provide adequate protection to the site during a CE probable maximum storm surge (PMSS) event so that the event damage potential is less than the event for which the plant was designed. The procedure 0AI-068 provides a timeline for the event progression including warning time, site preparation and time margin available. The operator response, warning time and time required for installation was evaluated against the NRC endorsed NEI 16-05 Appendix C [Reference 11] guidance for "Evaluating Manual Actions for the Integrated Assessment (IA)" process. It has been concluded, given the timeline presented, that a CE storm surge event is slow to develop and initial screening criteria C5 is appropriate given there is adequate time for the site to prepare and mitigate the effects from the storm.

The implementation of 10 CFR 50.69 at BSEP will not occur until the modifications presented in have been completed. This list is a subset of the complete list of commitments presented in the FE (Items #1-4, & 6 Reference 8, Enclosure 2). The proposed modifications to the smoke removal vents on the Fuel Oil Tank Chamber (FOTC) roof (Item #5 Reference 8, ) are not required for implementation of 10 CFR 50.69. The lowest elevations of these vents are above the maximum reevaluated SWEL for the CE storm surge and therefore are not required to be raised prior to implementation.

3.1.5 SSCs Required for External Flood Screening There are several components that if failed would result in an unscreened XF scenario for the CE storm surge event. The components are listed in the first column of the table below and if these components are categorized, they would be categorized as high safety significance (HSS) in accordance with the guidance in NEI 00-04 Figure 5-6. These components would not be categorized as LSS unless adequate justification is provided to demonstrate that their failure would not result in an unscreened scenario in accordance with NEI 00-04 Figure 5-6.

RA-19-0437 Page 9 of 23 Temporary Applicable Exterior Door ID Building/Structure Barrier Flood Required Mechanism Door D-2 LIP & CE 1-RB1-DR-EL020-210 Exterior Yes Storm Surge Equipment Airlock Doors Door D-3 LIP & CE 2-RB2-DR-EL020-210 Yes Storm Surge Exterior Equipment Airlock Doors Reactor Building 1-RB1-DR-EL020-204 CE Storm No Surge RB1 Personnel Airlock Exterior Door 2-RB2-DR-EL020-204 CE Storm No Surge RB2 Personnel Airlock Exterior Door Door D-4 CE Storm 2-DGB-DR-EL020-114 No Surge N. DGB Personnel Door Door D-5 CE Storm 2-DGB-DR-EL023-119 No Surge N. FOTC Enclosure Door D-6 Diesel Generator CE Storm 2-DGB-DR-EL023-126 Building and FOTC Yes Surge Exterior DBG Roll-Up Door Stairwell Enclosure Door D-7 CE Storm 2-DGB-DR-EL020-101 No Surge S. DGB Personnel Door Door D-8 CE Storm 2-DGB-DR-EL023-124 No Surge S. FOTC Enclosure Door D-13 CE Storm 2-SWB-DR-EL023-2 No Surge N. SWB Personnel Door Service Water Door D-14 Building CE Storm 2-SWB-DR-EL023-1 No Surge S. SWB Personnel Door 2-CTB-DR-EL023-101A/B CE Storm TB Breezeway to U2 Cable Spread Yes Surge Room Double Doors 2-CTB-DR-EL023-102 CE Storm No1 Surge TB Breezeway to U2 CTB Stairwell Control Building 2-CTB-DR-EL023-103 CE Storm Yes Surge TB Breezeway to U1 CTB Stairwell 2-CTB-DR-EL023-104A/B CE Storm TB Breezeway to U1 Cable Spread Yes Surge Room Double Doors

RA-19-0437 Page 10 of 23 Temporary Applicable Exterior Door ID Building/Structure Barrier Flood Required Mechanism 1-CTB-DR-EL023-105 CE Storm No1 Surge Radwaste to Cable Spread 2-CTB-DR-EL023-106 CE Storm Yes Surge Radwaste to Cable Spread Note 1 - BSEP modifications will replace the temporary barriers with permanent water tight doors providing permanent passive flood protection.

4 REGULATORY EVALUATION 4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA The following NRC requirements and guidance documents are applicable to the proposed change.

  • The regulations in Title 10 of the Code of Federal Regulations (10 CFR) Part 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors."
  • NRC Regulatory Guide 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, Revision 1, May 2006.

The proposed change is consistent with the applicable regulations and regulatory guidance.

4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS The 10 CFR 50.69 categorization process has been reviewed and approved by the US Nuclear Regulatory Commission (NRC) for Brunswick Steam Electric Plant (BSEP) by letter dated September 17, 2019 [Reference 1]. This proposed amendment revises the license condition that required using the BSEP XF PRA model in the categorization process to allow an XF screening evaluation based on proposed modifications which would allow the plant to mitigate the effects of all XF hazards rather than allowing water to enter into structures containing safety-related (SR) structures, systems or components (SSCs). All other aspects of the program remain as the NRC approved in Reference 1.

Duke Energy has evaluated whether a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

RA-19-0437 Page 11 of 23

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

Response: No.

The proposed change will eliminate the detailed XF PRA model from the license conditions specified in Reference 1. The process used to evaluate SSCs for changes to NRC special treatment requirements and the use of alternative requirements ensures the ability of the SSCs to perform their design function. Screening the XF PRA model from the license conditions does not change the design and operation of the SSCs. As a result, the proposed change does not significantly affect any initiators to accidents previously evaluated or the ability to mitigate any accidents previously evaluated. The consequences of the accidents previously evaluated are not affected because the mitigation functions performed by the SSCs assumed in the safety analysis are not being modified. The SSCs required to safely shut down the reactor and maintain it in a safe shutdown condition following an accident will continue to perform their design functions.

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

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

Response: No.

The proposed change will eliminate the detailed XF PRA model from the license conditions specified in Reference 1. The proposed change does not change the functional requirements, configuration, or method of operation of any SSC. Under the proposed change, no additional plant equipment will be installed.

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

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change will eliminate the detailed XF PRA model from the license conditions specified in Reference 1. The proposed change does not affect any Safety Limits or operating parameters used to establish the safety margin. The safety margins included in analyses of accidents are not affected by the proposed change. The regulation requires that there be no significant effect on plant risk due to any change to the special treatment requirements for SSCs and that the SSCs continue to be capable of performing their design basis functions, as well as to perform any beyond design basis functions consistent with the categorization process and results.

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

RA-19-0437 Page 12 of 23 Based on the above, Duke Energy concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.3 CONCLUSION

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

RA-19-0437 Page 13 of 23 5 ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendments would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendments do not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendments meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

RA-19-0437 Page 14 of 23 6 REFERENCES

1. US NRC Letter to Duke Energy, Brunswick Steam Electric Plant, Units 1 and 2 - issuance of Amendment Nos. 292 and 320 to adopt 10 CFR 50.69, "risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors" (EPID L-2018-LLA-0008), September 17, 2019, Agencywide Documents Access and Management System (ADAMS) Accession Number ML19149A471.
2. Duke Energy Letter to US NRC, Brunswick Steam Electric Plant, Units Nos. 1 and 2, Application to Adopt 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors, January 10, 2018, Agencywide Documents Access and Management System (ADAMS) Accession Number ML18010A344.
3. NRC Regulatory Guide 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, Revision 1, May 2006.
4. NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline," Revision 0, Nuclear Energy Institute, July 2005.
5. NRC Letter, 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; dated March 12, 2012, Agencywide Documents Access and Management System (ADAMS) Accession Number ML12053A340.
6. BSEP Letter, Flood Hazard Reevaluation Report, Response to NRC 10 CFR 50.54(f)

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, dated March 11, 2015, Agencywide Documents Access and Management System (ADAMS) Accession Number ML15079A385, March 11, 2015.

7. US NRC Letter to 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 (EPID NOS. 000495/05000325/L-2015-JLD-0007 AND 000495/05000324/L-2015-JLD-0008), April 16, 2018
8. Duke Energy Letter to US NRC, Brunswick Steam Electric Plant (BSEP), Unit Nos. 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, Agencywide Documents Access and Management System (ADAMS) Accession Number ML18270A372, September 27, 2018.
9. US NRC Letter to Duke Energy, Brunswick Steam Electric Plant, Units 1 And 2 - Staff Assessment Of Flooding Focused Evaluation (EPID NOS. 000495/05000324/L-2018-JLD-OO 16 AND 000495/05000325/L-2018-JLD-OO 16), January 16, 2019 Agencywide Documents Access and Management System (ADAMS) Accession Number ML18348B185.
10. BSEP Calculation, BNP-MECH-FHR-001, Evaluation of Local Intense Precipitation, Revision 0.
11. Nuclear Energy Institute (NEI), Report NEI 16-05, Revision 1, External Flooding Assessment Guidelines, Agencywide Documents Access and Management System (ADAMS) Accession Number ML16165A178, dated June 2016.

RA-19-0437 Page 15 of 23

12. BSEP Procedure, Administrative Instruction 0AI-68, Brunswick Nuclear Plant Response to Severe Weather Warnings, Revision 52.
13. Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, March 2009.
14. ASME/ANS RA-Sa-2009, Standard for Level l/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RA-S-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, dated February 2009
15. Duke Energy Letter to US NRC, 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, dated April 08, 2019, Agencywide Documents Access and Management System (ADAMS)

Accession Number ML19099A035.

RA-19-0437 Page 16 of 23 Attachment 1: List of Categorization Prerequisites The table below identifies the items that are required to be completed prior to implementation of the proposed 10 CFR 50.69 license condition in Section 2.3 of the LAR (i.e., the proposed change) at Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2.

The issues identified below will be addressed and any associated changes made. This list is a subset of the commitments found in Enclosure 2 of the FE [Reference 8].

Brunswick 50.69 XF PRA Implementation Items Description

i. BSEP will enhance the existing temporary passive interior door flood barriers (i.e., Cliff Edge Barriers) to provide protection for SSCs relied upon to protect KSFs in the Reactor Building (interior to external doors D-2 and D-3), Diesel Generator Building (Exterior to external door D-6) and Control Building (At CB entrance doors interior to TB external doors D-19 and D-22) from the CE Storm Surge event. The top elevation of these enhanced temporary passive barriers will be at elevation 27.5 ft NGVD29 to achieve minimum APM = 0.7 ft.

ii. For Control Building door locations, 2-CTB-DR-EL023-102 and 1-CTB-DR-EL023-105, BSEP modifications will replace the temporary barriers with permanent water-tight doors providing permanent passive flood protection.

iii. For the Service Water Building at the locations near Doors D-13 & D-14, BSEP will modify the existing security delay gate doors to shield ventilation openings from waves.

iv. For the DG Building north external personnel watertight door D-4, BSEP will provide a new debris barrier that protects this door subjected to debris loading associated with the CE Storm Surge event.

v. BSEP will confirm all barriers conform to the requirements for flood protection features specified in NEI 16-05, including Appendix B, and the time requirements of 0AI-68.

RA-19-0437 Page 17 of 23 Attachment 2: External Hazards Screening Screening Result External Hazard Screening Screened?

Criterion Comment (Y/N)

(Note a)

External Flooding Y C1 See External Flood Section 3.1 above.

C1 Hurricane Y See External Flood Section 3.1 above.

C5 Note a - See Attachment 3 for descriptions of the screening criteria.

This table contains only the changed entries from the previously submitted LAR.

RA-19-0437 Page 18 of 23 Attachment 3: Progressive Screening Approach for Addressing External Hazards Event Analysis Criterion Source Comments NUREG/CR-2300 C1. Event damage potential is Initial Preliminary and ASME/ANS

< events for which plant is Screening Standard RA-Sa-designed.

2009 C2. Event has lower mean NUREG/CR-2300 frequency and no worse and ASME/ANS consequences than other Standard RA-Sa-events analyzed. 2009 NUREG/CR-2300 C3. Event cannot occur close and ASME/ANS enough to the plant to affect it. Standard RA-Sa-2009 NUREG/CR-2300 Not used to screen.

C4. Event is included in the and ASME/ANS Used only to definition of another event. Standard RA-Sa- include within 2009 another event.

C5. Event develops slowly, allowing adequate time to ASME/ANS eliminate or mitigate the Standard threat.

PS1. Design basis hazard ASME/ANS Progressive cannot cause a core damage Standard RA-Sa-Screening accident. 2009 PS2. Design basis for the NUREG-1407 and event meets the criteria in the ASME/ANS NRC 1975 Standard Review Standard RA-Sa-Plan (SRP). 2009 NUREG-1407 as PS3. Design basis event modified in mean frequency is < 1E-5/y ASME/ANS and the mean conditional core Standard RA-Sa-damage probability is < 0.1.

2009 NUREG-1407 and PS4. Bounding mean CDF is ASME/ANS

< 1E-6/y. Standard RA-Sa-2009 Screening not successful. NUREG-1407 and PRA needs to meet ASME/ANS Detailed PRA requirements in the Standard RA-Sa-ASME/ANS PRA Standard. 2009

RA-19-0437 Page 19 of 23 : Markup of Facility Operating License Conditions (Appendix B of FOL)

RA-19-0437 Page 20 of 23 Operating License and Technical Specifications Brunswick Steam Electric Plant, Unit No. 1 Renewed Facility Operating License DPR-71 Updated through Amendment 299

RA-19-0437 Page 21 of 23 Amendment Additional Conditions Implementation Number Date 292 Duke Energy is approved to implement 10 CFR Upon implementation of 50.69 using the processes for categorization of Amendment No. 292.

Risk Informed Safety Class (RISC)-1, RISC-2, [XXX]

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 , and high winds; 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 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. 292 Unit 1 License Duke Energy will complete dated September 17, 2019. Amendment No. [XXX]

the implementation items dated [DATE].

listed in Attachment 1 of Duke Energy will complete the implementation Duke letter to the NRC dated items list in Attachment 1 of Duke letter to NRC

[DATE] prior to dated April 8, 2019 prior to implementation of 10 implementation of 10 CFR CFR 50.69. All issues identified in the 50.69 in accordance with the attachment will be addressed and any categorization process associated changes will be made, focused-scope peer reviews will be performed on described above. The issues changes that are PRA upgrades as defined in identified in the attachment the PRA standard (ASME/ANS RA-Sa-2009, as will be addressed and any endorsed by RG 1.200, Revision 2), and any associated changes will be findings will be resolved and reflected in the made prior to implementation PRA of record prior to implementation of the 10 of 10 CFR 50.69 in CFR 50.69 categorization process.

accordance with the categorization process Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization described above.

process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

Brunswick Unit 1 App. B-5 Amendment No. 292

RA-19-0437 Page 22 of 23 Operating License and Technical Specifications Brunswick Steam Electric Plant, Unit No. 2 Renewed Facility Operating License DPR-62 Updated through Amendment 326

RA-19-0437 Page 23 of 23 Amendment Number Additional Conditions Implementation Date 320 Duke Energy is approved to implement 10 CFR Upon implementation of 50.69 using the processes for categorization of Amendment No. 320.

Risk Informed Safety Class (RISC)-1, RISC-2, [XXX]

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 and high winds; 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 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. 320 Unit 2 License Amendment dated September 17, 2019. No. [XXX], dated [DATE].

Duke Energy will complete Duke Energy will complete the implementation the implementation items items list in Attachment 1 of Duke letter to NRC listed in Attachment 1 of dated April 8, 2019 prior to implementation of 10 Duke letter to the NRC dated CFR 50.69. All issues identified in the attachment

[DATE] prior to will be addressed and any associated changes will implementation of 10 CFR be made, focused-scope peer reviews will be 50.69 in accordance with the performed on changes that are PRA upgrades as categorization process defined in the PRA standard (ASME/ANS RA-Sa-described above. The issues 2009, as endorsed by RG 1.200, Revision 2), and identified in the attachment any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 will be addressed and any CFR 50.69 categorization process.

associated changes will be made prior to Prior NRC approval, under 10 CFR 50.90, is implementation of 10 CFR required for a change to the categorization 50.69 in accordance with the process specified above (e.g., change from a categorization process seismic margins approach to a seismic described above. probabilistic risk assessment approach).

Brunswick Unit 2 App. B-5 Amendment No. 320

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