BSEP 13-0097, Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard (NFPA) 805

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Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard (NFPA) 805
ML13246A276
Person / Time
Site: Brunswick  Duke Energy icon.png
Issue date: 08/29/2013
From: Hamrick G
Duke Energy Corp
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 13-0097
Download: ML13246A276 (26)
v • d • e


Text

Letter Enclosures 6, 7, and 10 Contain DUKE DUKE Security-Related Information - George Vice Hamrick T. President ENERGY. Withhold in Accordance with 10 CFR 2.390 Brunswick Nuclear Plant P.O. Box 10429 Southport, NC 28461 o: 910.457.3698 August 29, 2013 Serial: BSEP 13-0097 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324 Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard (NFPA) 805 (NRC TAC Nos. ME9623 and ME9624)

References:

1. Letter from Michael J. Annacone (Carolina Power & Light Company) to U.S.

Nuclear Regulatory Commission (Serial: BSEP 12-0106), License Amendment Request to Adopt NFPA 805 Performance-BasedStandardfor Fire Protectionfor Light Water ReactorElectric Generating Plants(2001 Edition), dated September 25, 2012, ADAMS Accession Number ML12285A428

2. Letter from Michael J. Annacone (Carolina Power & Light Company) to U.S.

Nuclear Regulatory Commission (Serial: BSEP 12-0140), Additional Information Supporting License Amendment Request to Adopt NFPA 805 Performance-BasedStandardfor Fire Protectionfor Light Water Reactor Electric GeneratingPlants (2001 Edition), dated December 17, 2012, ADAMS Accession Number ML12362A284

3. Letter from Christopher Gratton (USNRC) to Michael J. Annacone (Carolina Power & Light Company), Request for Additional Information Regarding Voluntary Risk Initiative National Fire ProtectionAssociation Standard805 (TAC Nos. ME9623 and ME9624), dated May 15, 2013, ADAMS Accession Number ML13123A231
4. Letter from George T. Hamrick (Duke Energy) to U.S. Nuclear Regulatory Commission (Serial: BSEP 13-0066), Response to Request for Additional Information Regarding Voluntary Risk Initiative NationalFire Protection Association Standard805 (TAC Nos. ME9623 and ME9624), dated June 28, 2013, ADAMS Accession Number ML13191B271
5. Letter from George T. Hamrick (Duke Energy) to U.S. Nuclear Regulatory Commission (Serial: BSEP 13-0070), Response to Request for Additional Information Regarding Voluntary Risk Initiative NationalFire Protection Association Standard805 (TAC Nos. ME9623 and ME9624), dated July 15, 2013, ADAMS Accession Number ML13205A016 When Enclosures 6, 7, and 10 to this letter are removed, this document is no longer Security-Related

U.S. Nuclear Regulatory Commission Page 2 of 5

6. Letter from George T. Hamrick (Duke Energy) to U.S. Nuclear Regulatory Commission (Serial: BSEP 13-0083), Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard805 (TAC Nos. ME9623 and ME9624), dated July 31, 2013, ADAMS Accession Number ML13220B041 Ladies and Gentlemen:

By letter dated September 25, 2012 (i.e., Reference 1), as supplemented by letter dated December 17, 2012 (i.e., Reference 2), Duke Energy Progress, Inc., formerly known as Carolina Power & Light Company (CP&L), submitted a license amendment request to adopt a new risk-informed performance-based (RI-PB) fire protection licensing basis for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2.

During the week of April 8 through 12, 2013, the NRC conducted an audit at the Brunswick Plant to support development of questions regarding the license amendment request. On May 15, 2013 (i.e., Reference 3), the NRC provided a set of requests for additional information (RAls) regarding the license amendment request. This letter divided these RAIs into 60-day, 90-day, and 120-day responses. In subsequent telephone calls with the NRC Project Manager for BSEP, the following modifications were agreed to regarding the RAI response schedule shown in the May 15, 2013, letter:

" The 60-day RAI responses were to be submitted by July 1, 2013 (i.e., 60 days following the May 2, 2013, clarification call that was conducted with the NRC). These responses were submitted by letter dated June 28, 2013 (i.e., Reference 4). Probabilistic Risk Assessment (PRA) RAls 1A, 1B, lC, 1D, IF, 1G, 11, 1K, 1N, 10, 1P, 1Q, 1R, 4, 5, 9, 10, 17, and 18, which were included in the set of 60-day RAIs, will be addressed in a separate submittal due by July 15, 2013 (i.e., 60 days following the date of the letter).

These responses were submitted by letter dated July 15, 2013 (i.e., Reference 5).

" The 90-day RAI responses were to be submitted by July 31, 2013 (i.e., 90 days following the May 2, 2013, clarification call). These responses were submitted by letter dated July 31, 2013 (i.e., Reference 6). Fire Protection Engineering RAI 1, which was included in the set of 60-day RAIs, will be addressed as part of the 90-day RAI responses.

  • The 120-day RAI responses were to be submitted by August 30, 2013 (i.e., 120 days following the May 2, 2013, clarification call). In addition, per Reference 4, PRA RAI 1H would be addressed as part of the 120-day RAI responses, rather than with the 60-day RAI responses.

A tabulation of the individual RAIs and the planned response submittal dates is provided in . Duke Energy's responses to the set of 120-day RAIs are provided in Enclosure 2.

During preparation of the 60-, 90-, and 120-day RAI responses, Duke Energy determined that the Brunswick NFPA 805 Transition Report, along with some of its supporting attachments, would need to be revised. The revised Transition Report and the revised attachments are provided as Enclosures 3 through 11.

The revisions provided to Enclosure 6 (i.e., Attachment C) address several clarifications agreed to during the NRC audit conducted during the week of April 8 through 12, 2013. The major revision applies to fire areas whose post-transition licensing basis is NFPA 805, Section 4.2.4, wherein Variance from Deterministic Requirement (VFDR) information that was previously

U.S. Nuclear Regulatory Commission Page 3 of 5 contained in one or more locations was summarized and reformatted into a single table. Other revisions included rewording discussions related to VFDRs and credited defense-in-depth features. The revisions to the other enclosures address items described in the following RAI responses.

NFPA 805 Transition Enclosure Report Section Related RAI(s) 3 Main Report FPE-17, FPE-19, 4 Attachment A FPE-5, FPE-6, FPE-7, FPE-9, FPE-10, FPE-11, FPE-13, FPE-14, FPE-15, FPE-21 5 Attachment B SSA-4 6 Attachment C FPE-15, FPE-19, FPRA-12, SSA-8, SSA-9, SSA-10, SSA-15 7 Attachment G SSA-12, SSA-15 8 Attachment J FM-3A, FM-3B, FM-3C 9 Attachment L FPE-16, FPE-18 10 Attachment S SSA-2, FPRA-16 11 Attachment V FPRA-1 B Enclosures 6, 7, and 10 (i.e., Attachments C, G, and S) contain security-related information and should be withheld from public disclosure under 10 CFR 2.390.

Additional time is needed to complete revision of the Fire Probabilistic Risk Assessment (FPRA) model and perform review of the new FPRA sensitivity study results. As a result, the following RAI responses are being delayed:

0 Fire Modeling RAI 1E 0 FPRA RAI 1D 0 FPRA RAI 1 E 0 FPRA RAI 1 H 0 FPRA RAI 3D 0 FPRA RAI 11 B Duke Energy plans to provide these sensitivity study results in a later submittal, via an update to the Brunswick NFPA 805 Transition Report, Transitionto 10 CFR 50.48(c) - NFPA 805 Performance-BasedStandardfor Fire Protectionfor Light Water ReactorElectric Generating Plants, 2001 Edition Main Report and a revision to NFPA 805 Transition Report, Attachment W, Fire PRA Insights. This information will be submitted no later than September 30, 2013.

This document contains no new regulatory commitments.

Please refer any questions regarding this submittal to Mr. Lee Grzeck, Manager - Regulatory Affairs, at (910) 457-2487.

U.S. Nuclear Regulatory Commission Page 4 of 5 I declare, under penalty of perjury, that the foregoing is true and correct. Executed on August 29, 2013.

Sincerely, Geo THamrick

Enclosures:

1. Revised Response Schedule to NFPA 805 Request for Additional Information
2. Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association (NFPA) Standard 805
3. Revised Brunswick NFPA 805 Transition Report, Transition to 10 CFR 50.48(c) - NFPA 805 Performance-BasedStandardfor Fire Protectionfor Light Water Reactor Electric GeneratingPlants, 2001 Edition, Transition Report, August 28, 2013, Main Report Without Attachments
4. Revised NFPA 805 Transition Report, Attachment A, Table B-1, Transitionof FundamentalsFireprotection Program& Design Elements
5. Revised NFPA 805 Transition Report, Attachment B, Table B-2, Nuclear Safety CapabilityAssessment Methodology Review
6. Revised NFPA 805 Transition Report, Attachment C, NEI 04-02 Table B Fire Area Transition (Security-Related Information - Withhold from Public Disclosure)
7. Revised NFPA 805 Transition Report, Attachment G, Recovery Actions Transition(Security-Related Information -Withhold from Public Disclosure)
8. Revised NFPA 805 Transition Report, Attachment J, Fire Modeling V&V
9. Revised NFPA 805 Transition Report, Attachment L, NFPA 805 Chapter3 Requirements for Approval (10 CFR 50.48(c)(2)(vii))
10. Revised NFPA 805 Transition Report, Attachment S, Modifications and Implementation Items (Security-Related Information - Withhold from Public Disclosure)
11. Revised NFPA 805 Transition Report, Attachment V, Fire PRA Quality

U.S. Nuclear Regulatory Commission Page 5 of 5 WRM/wrm cc (with all enclosures):

U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Victor M. McCree, Regional Administrator 245 Peachtree Center Ave, NE, Suite 1200 Atlanta, GA 30303-1257 U. S. Nuclear Regulatory Commission ATTN: Mr. Christopher Gratton (Mail Stop OWFN 8G9A) 11555 Rockville Pike Rockville, MD 20852-2738 U. S. Nuclear Regulatory Commission ATTN: Ms. Michelle P. Catts, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 cc (with Enclosures 1 through 5, 8, 9, and 11):

Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 Mr. W. Lee Cox, Ill, Section Chief Radiation Protection Section North Carolina Department of Health and Human Services 1645 Mail Service Center Raleigh, NC 27699-1645 lee.cox@dhhs.nc.gov W

Enclosure 1 Page 1 of 1 Revised Response Schedule to NFPA 805 Request for Additional Information Revised Response Schedule Section Title Question Number(s) Submittal Date 60-Day Response - Non-PRA Programmatic 1, 2,3, 4,5,6,7 July 1, 2013 Safe Shutdown Analysis 3, 4, 6, 7, 8,10,12 (Complete 1A, 1E, 1F, 1G, 1H, 2A, 2B, 5A, 5B une 28, 2013)

Fire Modeling Ji "W,

... ýR I . . ....... ....... .*

60-Day.Response PRA -

Probabilistic Risk I 1A, 1B, 1C, 1D, 1F, 1G, 11, 1N, 10, July 15, 2013 Assessment 1P, 1Q, 1R, 4, 5, 9, 10, 17, 18 (Complete 1 July 15, 2013) 90 Day4Response'.

Radiation Release 1, 2, 3 Fire Protection Engineering 1, 3, 4, 5,6,7,8,9,10, 11, 12,13,14, 15, 16, 17, 18, 19, 20, 21 July 31, 2013 Safe Shutdown Analysis 1, 2, 5, 9,11, 13,14 (Complete Probabilistic Risk 1J, 1K, M, 2, 3, 6, 7, 11, 12, 13, 14, 15, July 31, 2013)

Assessment 16 Fire Modeling 1B, 2C, 5C 120 Day esponse.' "" '

Fire Protection Engineering 2 Safe Shutdown Analysis 15 Probabilistic Risk 1E, 1H, 1L, 8 August 30, 2013 Assessment Fire Modeling 1C, 1D, 1I, 2D, 3, 4, 6 120 Day Response _________________________*___'_"_ .......

PRA Sensitivity Results

________________j3D, IFPRAFPRAD,1 FPRA 1E, FPRA 1H, FPRA 118, FM 1E September 30, 2013

Enclosure 2 Page 1 of 20 Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association (NFPA) Standard 805 By letter dated September 25, 2012 (i.e., ADAMS Accession Number ML12285A428), as supplemented by letter dated December 17, 2012 (i.e., ADAMS Accession Number ML12362A284), Duke Energy Progress, Inc., formerly known as Carolina Power & Light Company, submitted a license amendment request (LAR) to adopt a new risk-informed performance-based (RI-PB) fire protection licensing basis for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2.

During the week of April 8 through April 12, 2013, the NRC conducted an audit at BSEP to support development of questions regarding the license amendment request. On May 15, 2013, the NRC provided a set of requests for additional information (RAIs) regarding the license amendment request. That letter divided the RAIs into 60-day, 90-day, and 120-day responses.

In subsequent telephone calls with the NRC Project Manager for BSEP, the following modifications were agreed to regarding the RAI response schedule shown in the May 15, 2013, letter:

" The 60-day RAI responses were to be submitted by July 1, 2013 (i.e., 60 days following the May 2, 2013, clarification call that was conducted with the NRC). These responses were submitted by letter dated June 28, 2013 (i.e., Reference 4). Fire Probabilistic Risk Assessment (FPRA) RAIs 1A, 1B, 1C, 1D, 1F, 1G, 11, 1K, 1N, 10, 1P, 1Q, 1R, 4, 5, 9, 10, 17, and 18, which were included in the set of 60-day RAIs, will be addressed in a separate submittal due by July 15, 2013 (i.e., 60 days following the date of the letter).

These responses were submitted by letter dated July 15, 2013 (i.e., Reference 5).

  • The 90-day RAI responses were to be submitted by July 31, 2013 (i.e., 90 days following the May 2, 2013, clarification call). These responses were submitted by letter dated July 31, 2013 (i.e., Reference 6). Fire Protection Engineering RAI 1, which was included in the set of 60-day RAIs, will be addressed as part of the 90-day RAI responses.
  • The 120-day RAI responses were to be submitted by August 30, 2013 (i.e., 120 days following the May 2, 2013, clarification call). In addition, per Reference 4, FPRA RAI 1H would be addressed as part of the 120-day RAI responses, rather than with the 60-day RAI responses.

Duke Energy's 120-day response to the RAls is provided below. Additional time is needed to complete revision of the FPRA model and perform review of the new FPRA sensitivity study results. As a result, the following RAI responses are being delayed:

  • Fire Modeling RAI 1E
  • FPRARAIID
  • FPRARAIIE

" FPRARAI1H

" FPRARAI 3D

  • FPRARA111B Duke Energy plans to provide these sensitivity study results in a later submittal, via an update to the Brunswick NFPA 805 Transition Report, Transition to 10 CFR 50.48(c) - NFPA 805 Performance-BasedStandardfor Fire Protectionfor Light Water Reactor Electric Generating

Enclosure 2 Page 2 of 20 Plants,2001 Edition Main Report and a revision to NFPA 805 Transition Report, Attachment W, Fire PRA Insights. This information will be submitted no later than September 30, 2013.

Fire Protection Engineering Requests for Additional Information Fire Protection Engineering RAI 2 Attachment S, Table S-1, Item #11 of the LAR identifies a modification of suspended ceiling configuration to allow for an effective increase in ceiling height and associated volume of the MCR. Provide a more detailed explanation of what this modification entails. Describe whether the suspended ceiling is to be removed. Describe how this modification will affect the fire detection systems (i.e., including the potential for stratification), both current and planned detection. Describe how this modification will be incorporated into the fire protection program.

Response

As identified in the Control Room Abandonment Calculation, there exists limited margin for risk impacts (i.e., core damage frequency/large early release frequency (CDF/LERF)) driven by temperature and visibility obscuration. This conclusion is identified in the Kleinsorg Group Report No. 1RCS04042.103.007-01, Evaluation of Main Control Room Abandonment Times at the Brunswick Nuclear Plant,which is contained in Attachment 14 of Calculation Number BNP-PSA-082, BNP Fire PRA - NFPA 805 Transition Support. The purpose of this modification (i.e., Attachment S, Table S-1, Item 11, of the LAR) is to provide access to the vast interstitial space located above the Main Control Room (MCR). Since the current configuration is included in the FPRA, the modification will provide enhanced risk margin for CDF/LERF for the MCR.

As a part of the modification process, an evaluation will be performed to determine how much of the suspended ceiling (from several ceiling tiles to the entire ceiling) will need to be removed to successfully access the interstitial space to allow the hot gases from a fire to flow, which in turn will increase abandonment time. The evaluation will also need to take into consideration:

  • Human factors that impact the operators working in the space (i.e., acoustics, lighting, etc.)

" MCR systems (i.e., ventilation systems, detection systems, etc.)

Detailed fire modeling will be completed to determine how the proposed change will affect fire conditions and scenarios, including the potential for stratification. The fire modeling may yield a change to the existing fire detection system, or the need to install a new fire detection system.

These results will then be folded back into the FPRA and fire protection program through the Engineering Change (EC) process, in accordance with procedure EGR-NGGC-0005. The EC process will also include NFPA 805 impact review, as specified by procedure FIR-NGGC-0010, Fire ProtectionProgram Change Process.

Safe Shutdown Analysis Requests for Additional Information SSA RAI 15 During the audit, it was discussed that Table B-3 of the LAR would be updated to identify dispositions for each VFDR. Provide the updated Attachment C, Table B-3 of the LAR.

Enclosure 2 Page 3 of 20

Response

The revisions to Attachments C and G address several clarifications agreed to during the audit.

The major changes applied to fire areas whose post transition licensing basis is NFPA 805, Section 4.2.4, wherein Variance from Deterministic Requirement (VFDR) information that was previously contained in one or more locations was summarized and reformatted into a single table. Minor revisions included rewording discussions related to VFDRs and credited defense in depth features.

The following fire areas were affected by the major updates:

CB-1 DG-4 DG-1 1 DUCTBANK TB-1 CB-2 DG-5 DG-12 RBI-1 CB-23E DG-7 DG-13 RB2-1 DG-1 DG-8 DG-16E SW1-1 For each affected area, a new VFDR detail table replaced the existing "VFDR List" table and either "Type 3" or "Type 4" VFDR tables, as applicable. The information previously given in the supplanted tables is summarized by affected component in the new table and supplemented with Performance Goal and Failure Impact columns. The performance goals listed in Attachment G and in the Comments column of the B-3 tables duplicated those listed in the VFDR detail tables and were deleted. The introductory paragraphs in these areas were also provided with a general description of the recovery actions and performance goals affected by the VFDRs.

Other revisions made to Attachment C are listed below:

CB-23E, Units 1 and 2: The VFDR list for the control room was modified to reflect the treatment of risk in the FPRA for the Control Building. The net result was the deletion of VFDRs associated with fire induced cable damage in keeping with the MCR abandonment scenario wherein all such damage is isolated from safe shutdown circuits at locations remote from the fire. What remains is a listing of recovery actions evaluated in the FPRA which are performed at locations other than the primary control station.

  • DG-5 and DG-8, Units 1 and 2: Additional information was added describing Defense-in-Depth (DID) fire protection features in the form of pyrocrete and Electrical Raceway Fire Barrier System (ERFBS), respectively.
  • DG-1, Units 1 and 2: The discussion in the B-3 table that referred to the fire protection features related to the non-transitioning exemption was deleted.

All Fire Areas: Any reference to or discussion of pre-transition exemptions granted under the Appendix R current licensing basis were removed.

  • Line items were added to the VFDR detail table as required so that the dispositions aligned with the components assigned to recovery actions in Attachment G. The added records do not represent new or unanalyzed VFDRs in the FPRA, but rather create an improved trail between Attachments C and G.

RB1-1, Units 1 and 2: Additional VFDRs reflected in the updated FPRA results were added.

Enclosure 2 Page 4 of 20 Attachment G was edited as follows:

Recovery Actions (RAs) that applied to multiple components were listed multiple times to match entries in the new VFDR Detail tables in Attachment C.

Two RA-DID actions in analysis area RB2-1, Unit 1 were deleted as the risk results for RB2-1, Unit 1 were below the screening threshold provided in the supplement to Attachment C.

  • Similarly, three RA-DID actions in TB-1 for Units 1 and 2 were removed.
  • Recovery action text that was previously truncated was corrected.
  • Where a recovery action for a shared component was also applicable to the opposite unit, the action was copied to the appropriate table.

Enclosure 2 Page 5 of 20 Fire Probabilistic Risk Assessment Requests for Additional Information FPRA RAI ID F&O 1-19 against FSS-Al (Not Met):

The disposition for this F&O explains that the ZOI associated with a 143 kilo-watt (kW) heat release rate (HRR) (7 5 th percentile) transient fire was used in all fires areas, except the turbine building where a ZOI for a 317 kW HRR ( 9 8 th percentile) fire was used. The disposition provides the basis for this lower HRR as existing and planned administrative controls, plant experience, and insights from a bounding sensitivity study. Provide further justification for the use of 143 kW transient fires, given that both 143 kW and 317 kW are taken from the same HRR distribution. Include further description of the administrative controls used in the different areas for managing transient combustibles, the results of reviewing plant experience and records of violations of transient combustible controls, other key factors for this reduced fire size, and the results of the bounding sensitivity study referred to in the disposition. Also, confirm that 143 kW and 317 kW HRRs were the only transient fire sizes used in the FPRA.

Response

A response to FPRA RAI 1D was provided in Duke Energy's letter dated July 15, 2013 (i.e.,

ADAMS Accession Number ML13205A016). This response indicated that a new sensitivity study would be submitted in the 120-day responses with the results of other sensitivity studies.

Additional time is needed to complete revision of the FPRA model and perform review of the new FPRA sensitivity study results. As a result, Duke Energy plans to provide these sensitivity study results in a later submittal, via an update to the Brunswick NFPA 805 Transition Report, Transition to 10 CFR 50.48(c) - NFPA 805 Performance-BasedStandardfor Fire Protectionfor Light Water Reactor Electric GeneratingPlants, 2001 Edition Main Report and a revision to NFPA 805 Transition Report, Attachment W, Fire PRA Insights. This information will be submitted no later than September 30, 2013.

FPRA RAI IE Clarify the following dispositions to fire F&Os and supporting requirement (SR) assessments identified in Attachment V of the LAR that have the potential to impact the FPRA results and do appear to be fully resolved:

e) F&O 1-20 against FSS-A1 (Not Met):

As stated in the disposition, Appendix H.2 of NUREG/CR-6850 recommends that vulnerability to transient fires be limited to cable vulnerability. However, Appendix H.2 also recommends that if sensitive electronics can be impacted, then ignition of such components should be considered. Describe how the impact on sensitive electronics from transient fires is modeled in the FPRA; as appropriate, refer to the draft FAQ under development on sensitive electronics. If this impact was not considered, provide a sensitivity study that estimates this impact on core damage frequency (CDF), large early release frequency (LERF), ACDF, and ALERF.

Enclosure 2 Page 6 of 20

Response

Additional time is needed to complete revision of the FPRA model and perform review of the new FPRA sensitivity study results. Accordingly, Duke Energy plans to provide these sensitivity study results in a later submittal, via an update to the Brunswick NFPA 805 Transition Report, Transitionto 10 CFR 50.48(c) - NFPA 805 Performance-BasedStandard for Fire Protectionfor Light Water ReactorElectric GeneratingPlants,2001 Edition Main Report and a revision to NFPA 805 Transition Report, Attachment W, FirePRA Insights. This information will be submitted no later than September 30, 2013.

FPRA RAI IH Clarify the following dispositions to fire F&Os and supporting requirement (SR) assessments identified in Attachment V of the LAR that have the potential to impact the FPRA results and do appear to be fully resolved:

h) F&O 1-32 against FSS-Cl (Cat 1):

The disposition to this F&O states, based on footnotes to NUREG/CR-6850 Table G-1, that for the 9 8 1h percentile case, an HRR associated with motor fires (69 kW) was used for pump electrical fires rather than the pump electrical HRR of 211 kW that is recommended by NUREG/CR-6850, Table G-1. Provide a sensitivity study that shows that impact on CDF, LERF, ACDF, and ALERF of using the NUREG/CR-6850 recommended HRR of 211 kW as the 98th percentile HRR for pump electrical fires.

Response

Additional time is needed to complete revision of the FPRA model and perform review of the new FPRA sensitivity study results. Accordingly, Duke Energy plans to provide these sensitivity study results in a later submittal, via an update to the Brunswick NFPA 805 Transition Report, Transition to 10 CFR 50.48(c) - NFPA 805 Performance-BasedStandard for Fire Protectionfor Light Water Reactor Electric GeneratingPlants,2001 Edition Main Report and a revision to NFPA 805 Transition Report, Attachment W, Fire PRA Insights. This information will be submitted no later than September 30, 2013.

FPRA RAI IL Clarify the following dispositions to fire F&Os and supporting requirement (SR) assessments identified in Attachment V of the LAR that have the potential to impact the FPRA results and do appear to be fully resolved:

I) F&O 2-16 against FSS-D9 (Cat 1):

Provide additional justification for not postulating smoke damage. Address in this justification the specific types of components vulnerable to smoke damage and the potential damage mechanisms presented in Appendix T of NUREG 6850.

Response

This RAI response discusses the treatment of smoke damage in the BSEP FPRA and the effects on the quantification results. Treatment for smoke damage is discussed in Appendix T of

Enclosure 2 Page 7 of 20 NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities Volume 2: Detailed Methodology. The practical implication of the guidance in Appendix T of NUREG/CR-6850 is that short term smoke damage (i.e., damage generated by exposure to smoke during the fire event or shortly after it is suppressed) is limited to electrical enclosures with high smoke concentration. In most cases, these high concentrations of smoke will happen within the electrical panels physically connected to the panel of fire origin. Examples of this configuration could include breaker cubicles within the same Motor Control Center (MCC) or switchgear where the fire started, or relay panels within the same relay panel bank where the fire started. In summary, smoke damage is not postulated outside the interconnected panels adjacent to the cabinet of fire origin. In addition, Appendix T of NUREG/CR-6850 limits the equipment vulnerable to short-term smoke damage to medium and high voltage switching or transmission equipment, and lower voltage instrumentation and control devices.

The BSEP FPRA currently accounts for smoke damage consistent with the guidance in Appendix T of NUREG/CR-6850 by failing the entire electrical bus or panel where the fire is postulated. As an example, if the fire fails all the cables entering a cabinet, all the basic events associated with the function of the cabinet will fail. This accounts for any smoke damage generated inside the panel.

Summary of Results - Smoke Damage Evaluation A qualitative evaluation has been conducted for determining if the current treatment of smoke damage in the BSEP FPRA has a significant effect on the quantified Fire Risk values (i.e., CDF and LERF). The results of the evaluation suggest that the current treatment of smoke damage in the BSEP FPRA is consistent with the guidance in Appendix T of NUREG/CR-6850. The quantification approach for ignition sources and targets inside and outside of the MCR captures the failure of credited function in the FPRA that could be generated by smoke damage (i.e.,

function failures within the panels connected to the panel of fire origin). Consequently, it is concluded that the Fire Risk results currently bounds failures of components vulnerable to smoke damage consistent with current guidance in NUREG/CR-6850, Appendix T.

PRA RAI 3D NUREG/CR-6850 Section 6 and FAQ 12-0064 describe the process for assigning influence factors for hot work and transient fires. Provide the following regarding application of this guidance:

d) Given that a weighting factor of "50" was not used in any fire area, provide a sensitivity study that assigns weighting factors of "50" per the guidance in FAQ 12-0064.

Response

A response to FPRA RAI 3D was provided in Duke Energy's letter dated July 31, 2013 (i.e.,

ADAMS Accession Number ML13220B041). The response indicated that a new sensitivity study would be submitted in the 120-day responses, assigning a weighting factor of 50 for an area of "very high" maintenance or hotwork as described in Table 6-3 of FAQ 12-0064. Additional time is needed to complete revision of the FPRA model and perform review of the new FPRA sensitivity study results. As a result, Duke Energy plans to provide these sensitivity study results in a later submittal, via an update to the Brunswick NFPA 805 Transition Report, Transition to 10 CFR 50.48(c) - NFPA 805 Performance-BasedStandardfor Fire Protectionfor Light Water Reactor Electric Generating Plants, 2001 Edition Main Report and a revision to NFPA 805

Enclosure 2 Page 8 of 20 Transition Report, Attachment W, Fire PRA Insights. This information will be submitted no later than September 30, 2013.

FPRA RAI 8 Describe how heating, ventilation, and air conditioning (HVAC) modeling was performed to support the FPRA and whether HVAC cable tracing and fire modeling were performed to support this modeling. Confirm that additional operator actions are not needed for crediting HVAC. Heat load calculations performed for the internal events probabilistic risk assessment (PRA) do not account for the additional heat load from fires. Confirm that heat loads from fires do not fail additional equipment in rooms that do not credit HVAC.

Response

HVAC modeling for the FPRA was developed from HVAC modeling for the Internal Events PRA, which is described in Attachment 21 of BNP-PSA-062, PRA Model Appendix A System Notebooks. In particular, the Diesel Generator Building HVAC was included within the scope of component selection for the FPRA, and the associated cables were routed such that the risk impacts of potential fire effects can be evaluated. As described in Attachment 16 of BNP-PSA-080, BNP Fire PRA - Quantification,different operating configurations of the Control Building HVAC System were evaluated in the timing study for MCR abandonment and credited in the risk evaluation for MCR abandonment. More realistically, the firefighting actions (e.g.,

opening doors to gain access, spraying water) can be expected to lower temperatures in the room relatively early in the fire scenario, but no such additional operator action was credited for HVAC concerns in the FPRA.

Although the heat load calculations performed for the Internal Events PRA do not account for the additional heat load from fires, the results can be considered for limited applicability to the FPRA. As described in BNP-PSA-067, PRA Model Appendix H Room Heat-UpAnalysis, and BNP-PSA-070, PSA Related GOTHIC Thermal/HydraulicCodes ForBNP, a series of room heat-up analyses based on the GOTHIC code was performed for the Reactor Building, Control Building, and Diesel Generator Building and included consideration for the diesel generators, plant batteries, Residual Heat Removal (RHR), High Pressure Coolant Injection (HPCI), Reactor Core Isolation Cooling (RCIC), Core Spray, and Control Rod Drive (CRD). Small Break Loss of Cooling Accident (SBLOCA) and Large Break Loss of Cooling Accident (LBLOCA) were identified as the bounding scenarios for heat-up based on equipment demands, and the equipment demands for fire scenarios are not expected to be as great. For the Internal Events PRA, the highest room temperatures occurred much later in those scenarios than the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> assumed for the fire duration. During the assumed 1-hour fire duration, the analysis for Internal Events PRA showed estimated margins to the limiting room temperatures of between 150 F and 80° F. The analyses concluded that no HVAC was required to support mitigating system success during the 24-hour mission time of the Internal Events PRA, except for the Diesel Generator Building HVAC system.

Equipment failures due to heat loads from fires are mitigated for many Fire Compartments in the Reactor Building and the Turbine Building by "open" features (e.g., doorways, stairways, equipment hatches, pipe chases, and floor gratings) that would not facilitate the heat buildup required to reach room temperatures that would fail additional equipment. Openings to Fire Compartments at higher elevations promote a natural convection of heat upward and away from the equipment, which would generally be located closer to the floor. In many cases, the heat

Enclosure 2 Page 9 of 20 would eventually rise to the Refueling Floor in the Reactor Building or the Turbine Deck in the Turbine Building. Because the Turbine Deck and Refueling Floor are very large with high ceilings, no additional equipment failure due to HVAC impact is postulated in these cases.

Even for Fire Compartments that are not conducive to natural convection, the conditions under which a fire has the potential to fail additional equipment indirectly due to room temperature is very limited. The equipment of interest would need to be in an enclosed room with an ignition source. But to avoid direct failure, all of the cables associated with the equipment of interest would need to be outside the zone of influence of the ignition source. The fire would need to grow large enough, perhaps propagating to surrounding cable trays, until the temperature near the equipment of interest rose to some limiting value for equipment damage during the assumed 1-hour fire duration. However during this growth and propagation phase, the fire could not damage the cables associated with the equipment of interest and could not grow large enough to form a Hot Gas Layer (HGL) because the FPRA assumes all equipment with cables in that room to be damaged at the point that an HGL forms.

Beyond these physical constraints, several assumptions in the FPRA mitigate the risk impact of additional equipment failure in rooms that do not credit HVAC:

" The FPRA assumed the condenser, condensate system, and feedwater system were failed for every scenario.

" To simplify the analysis, some fire scenarios were modeled as a HGL, regardless of whether an HGL is plausible in the Fire Compartment.

" Even assuming additional equipment failure due to room temperature, the resulting combination of ignition frequency, non-suppression probability, and either conditional core damage probability or conditional large early release probability, would need to be greater than the effective truncation for the FPRA of 1 E-9 for CDF or 1 E-1 0 for LERF in order for there to be any contribution to fire risk.

While no additional GOTHIC analysis was performed to confirm that heat loads from fires do not fail additional equipment in rooms that do not credit HVAC, an inspection of the FPRA inputs and assumptions indicated that about 99% of fire scenarios would screen out based on this qualitative assessment.

FPRA RAI IIB 6 of BNP-PSA-080 describes how the risk of MCR abandonment was calculated for fire in Fire Area CB-23E. Address the following:

b) The abandonment risk is highly sensitive to whether the MCR electrical cabinets are assumed to be single-bundle cables or multiple-bundle cables. Provide justification for the assumption that the MCR cabinets only contain single-bundle cables. If cabinets containing multiple-bundle cables are present in the MCR, provide the results of a sensitivity analysis accounting for the MCR cabinets that contain multiple-bundle cables.

Response

A response to FPRA RAI 11 B was provided in Duke Energy's letter dated July 31, 2013 (i.e.,

ADAMS Accession Number ML13220B041). This response indicated that a new sensitivity study which evaluates the treatment of the Main Control Boards (MCBs) as multi-bundle fires

Enclosure 2 Page 10 of 20 would be submitted in the 120-day responses with the results of other sensitivity studies.

Additional time is needed to complete revision of the FPRA model and perform review of the new FPRA sensitivity study results. Accordingly, Duke Energy plans to provide these sensitivity study results in a later submittal, via an update to the Brunswick NFPA 805 Transition Report, Transition to 10 CFR 50.48(c) - NFPA 805 Performance-BasedStandardfor Fire Protectionfor Light Water Reactor Electric GeneratingPlants, 2001 Edition Main Report and a revision to NFPA 805 Transition Report, Attachment W, Fire PRA Insights. This information will be submitted no later than September 30, 2013.

Enclosure 2 Page 11 of 20 Fire Modeling Requests for Additional Information Fire Modeling RAI 1C NFPA 805, Section 2.4.3.3, states: "The PSA [probabilistic safety assessment] approach, methods, and data shall be acceptable to the AHJ [authority having jurisdiction] ..." The NRC staff noted that fire modeling comprised the following:

  • The Generic Fire Modeling Treatments (GFMTs) approach was used to determine the ZOI for transient and oil spill fires in all fire areas throughout plant

" Fire Dynamics Tools (FDT's) were used for ZOI calculations of cabinet and cable tray fires throughout the plant

  • The Consolidated Fire Growth and Smoke Transport (CFAST) model was used to calculate control room abandonment times
  • Fire Dynamics Simulator used for various fire hazard calculations Section 4.5.1.2, "FPRA" of the LAR states that fire modeling was performed as part of the FPRA development (NFPA 805 Section 4.2.4.2). Reference is made to Attachment J, "Fire Modeling V&V [verification and validation]," for a discussion of the acceptability of the fire models that were used.

Specifically regarding the acceptability of CFAST for the MCR abandonment times study:

c. The sensitivity study in Appendix B of the MCR abandonment times study shows that poorly ventilated burning conditions result in a significant reduction of the MCR abandonment times in some scenarios. For instance, according to Table B-3, for a scenario involving a closed cabinet with multiple cable bundles and normal ventilation, poorly ventilated burning conditions result in a reduction of the abandonment time from 9.41 to 5.85 minutes. Explain how these abandonment time reductions affect the CDF, ACDF, LERF and ALERF; or provide justification for why these scenarios were not included in the FPRA calculations.

Response

The sensitivity analysis in Appendix B.3 of the MCR Abandonment Report (BNP-PSA-080 Revision 3, Attachment 16) investigates the impacts of poorly ventilated conditions on a local scale, such as may occur with a fire inside a cabinet enclosure, with the room as a whole considered well-ventilated. The time to abandonment for the cabinet fires with poorly ventilated conditions are simulated in CFAST by increasing the ratio of the soot yield to carbon dioxide yield from 0.184 kg soot/kg CO 2 to 0.36 kg soot/kg CO 2 . The sensitivity calculation is a conservative assessment because it assumes a fire located on the floor, rather than at the height of the top of a cabinet, as suggested in NUREG/CR-6850. When the normal ventilation case, with increased soot yield is rerun using a fire elevation of 2 meters, the time to abandonment is 7.3 minutes for the multiple bundle case and 12.4 minutes for one bundle case.

In addition, the multiple bundle results compared to the single bundle results are conservative because the majority of cabinets in the control room are single bundle cabinets. These results show that poorly ventilated conditions do not impact the FPRA calculations for the one-bundle cabinets and have only a small impact for the multiple bundle cabinets.

Enclosure 2 Page 12 of 20 Table 1. Results Sensitivity for the Poorly Ventilated Case Abandonment Time (min)

Closed Cabinet, Closed Cabinet, Ventilation Multiple Confguraion Sensitivity Parameter Value One Bundle ne BndleBundles Configuration Baseline 11.84 (V) 9.41 (T)

Poorly Ventilated: Increase soot 9.26 (V) 5.85 (V)

Normal yield and fire on floor Poorly Ventilated: Increase soot yield and fire elevation at 12.7 (T) 7.3 (T) 2 meters T = Abandonment due to temperature.

V = Abandonment due to visibility.

Fire Modeling RAI ID NFPA 805, Section 2.4.3.3, states: "The PSA [probabilistic safety assessment] approach, methods, and data shall be acceptable to the AHJ [authority having jurisdiction] ..." The NRC staff noted that fire modeling comprised the following:

  • The Generic Fire Modeling Treatments (GFMTs) approach was used to determine the ZOI for transient and oil spill fires in all fire areas throughout plant
  • Fire Dynamics Tools (FDT's) were used for ZOI calculations of cabinet and cable tray fires throughout the plant
  • The Consolidated Fire Growth and Smoke Transport (CFAST) model was used to calculate control room abandonment times

" Fire Dynamics Simulator used for various fire hazard calculations Section 4.5.1.2, "FPRA" of the LAR states that fire modeling was performed as part of the FPRA development (NFPA 805 Section 4.2.4.2). Reference is made to Attachment J, "Fire Modeling V&V [verification and validation]," for a discussion of the acceptability of the fire models that were used.

Specifically regarding the acceptability of the GFMTs approach:

d. Explain how the modification to the critical heat flux for a target that is immersed in a thermal plume described in Section 2.4 of the GFMTs document was used in the ZOI determination.

Response

The modification to the critical heat flux (i.e., modified critical heat flux) tabulated in the Generic Fire Modeling Treatments report was not used to calculate zones of influence (ZOls) at BSEP.

Calculation HNP-M/MECH-1 129, Fire Zone of Influence Calculation,was used to determine ZOls for BSEP.

Enclosure 2 Page 13 of 20 Fire Modeling RAI IE NFPA 805, Section 2.4.3.3, states: "The PSA [probabilistic safety assessment] approach, methods, and data shall be acceptable to the AHJ [authority having jurisdiction] ..." The NRC staff noted that fire modeling comprised the following:

  • The Generic Fire Modeling Treatments (GFMTs) approach was used to determine the ZOI for transient and oil spill fires in all fire areas throughout plant

" Fire Dynamics Tools (FDT's) were used for ZOI calculations of cabinet and cable tray fires throughout the plant

  • The Consolidated Fire Growth and Smoke Transport (CFAST) model was used to calculate control room abandonment times
  • Fire Dynamics Simulator used for various fire hazard calculations Section 4.5.1.2, "FPRA" of the LAR states that fire modeling was performed as part of the FPRA development (NFPA 805 Section 4.2.4.2). Reference is made to Attachment J, "Fire Modeling V&V [verification and validation]," for a discussion of the acceptability of the fire models that were used.

Regarding the acceptability of the PSA approach, methods, and data in general:

e. Explain how the effect of the increased HRR from intervening combustibles (cable trays) on the ZOI was accounted for, or provide justification for ignoring this effect.

Response

A response to Fire Modeling RAI 1 E was provided in Duke Energy's letter dated June 28, 2013 (i.e., ADAMS Accession Number ML13191B271). The response indicated that sensitivity analyses for identifying impact on CDF, ACDF, LERF and ALERF was under development, as some 90-day and 120-day RAIs would influence the quantification process. Additional time is needed to complete revision of the FPRA model and perform review of the new FPRA sensitivity study results. As a result, Duke Energy plans to provide these sensitivity study results in a later submittal, via an update to the Brunswick NFPA 805 Transition Report, Transition to 10 CFR 50.48(c) - NFPA 805 Performance-BasedStandardfor Fire Protection for Light Water Reactor Electric GeneratingPlants, 2001 Edition Main Report and a revision to NFPA 805 Transition Report, Attachment W, FirePRA Insights. This information will be submitted no later than September 30, 2013.

Fire Modeling RAI 11 NFPA 805, Section 2.4.3.3, states: "The PSA [probabilistic safety assessment] approach, methods, and data shall be acceptable to the AHJ [authority having jurisdiction] ..." The NRC staff noted that fire modeling comprised the following:

  • The Generic Fire Modeling Treatments (GFMTs) approach was used to determine the ZOI for transient and oil spill fires in all fire areas throughout plant
  • Fire Dynamics Tools (FDT's) were used for ZOI calculations of cabinet and cable tray fires throughout the plant
  • The Consolidated Fire Growth and Smoke Transport (CFAST) model was used to calculate control room abandonment times

Enclosure 2 Page 14 of 20

. Fire Dynamics Simulator used for various fire hazard calculations Section 4.5.1.2, "FPRA" of the LAR states that fire modeling was performed as part of the FPRA development (NFPA 805 Section 4.2.4.2). Reference is made to Attachment J, "Fire Modeling V&V [verification and validation]," for a discussion of the acceptability of the fire models that were used.

Regarding the acceptability of the PSA approach, methods, and data in general:

L Attachment 7 to BNP-PSA-080, Revision 2, discusses the multi-compartment analysis (MCA). Section 1.1.2 discusses some of the underlying assumptions in the MCA, which include that (1) an open surface area of approximately 9 ft 2 is a general rule of thumb for the minimum area between compartments to transmit a HGL, and (2) the zone of influence is defined as approximately 5 ft vertical and 2 ft horizontal. The licensee stated that this is appropriate, based on discussions with industry experts, which concluded that "the heat diffusion in the adjacent room would limit the HGL to a local area around the failed barrier." Provide additional information about how these two sets of criteria were specifically used in the MCA.

Response

This response will address the two parts to the question.

Part 1:

(1) "an open surface area of approximately 9 ft2 is a general rule of thumb for the minimum area between compartments to transmit a HGL" This criterion is not used in the analysis and will be removed from the documentation in to Calculation BNP-PSA-080, Revision 4. Drawing review identified compartment 2

combinations to be used for the multi-compartment analysis (MCA) without the 9 ft consideration. It was assumed that any adjacent compartment to the exposing room is potentially able to transmit a hot gas layer, and each combination was analyzed in the MCA.

Additionally, the HGL calculation does not consider a 9 ft 2 opening between exposing and exposed compartments. The volumes of both compartments are simply summed assuming total barrier failure.

Part 2:

(2) the zone of influence is defined as approximately 5 ft vertical and 2 ft horizontal. The licensee stated that this is appropriate,based on discussions with industry experts, which concluded that "the heat diffusion in the adjacent room would limit the HGL to a local area around the failed barrier.

A 5 foot vertical and 2 foot horizontal ZOI was used to identify local targets in the exposed compartment that could potentially fail due to barrier failure. Walk downs were conducted to verify local targets above potentially failed barriers and are documented in Attachment 7 to Calculation BNP-PSA-080, Revision 4.

Enclosure 2 Page 15 of 20 Fire Modeling RAI 2D NFPA 805, Section 2.5, requires damage thresholds be established to support the performance-based approach. Thermal impact(s) must be considered in determining the potential for thermal damage of structures, systems, or components. Appropriate temperature and critical heat flux criteria must be used in the analysis.

Section 3.1.1.b of the HGL Calculation (BNP-MECH-HGL-001, Rev. 1), states that "BNP predominantly has thermoset cables so the damage criteria associated with thermoset cables has been used in this analysis."

d. Describe the damage criteria that were used for exposed temperature-sensitive equipment. Explain how temperature-sensitive equipment inside an enclosure was treated, and provide a technical justification for these damage criteria.

Response

Calculation BNP-PSA-080, Revision 3, Section 3.6.6, Table 9, "Sources of Uncertainty,"

indicates that sensitive electronics are typically contained in cabinets/panels, and enclosure will provide some amount of protection from external fires. For fires inside MCR cabinets or Remote Shutdown panels, failure of the components is already assumed for fires originating within the electrical cabinet or panel. For Main Control Boards, the failure of sensitive electronics is mitigated by incipient detection. For sensitive electronics not contained in enclosures, it is very likely that the cables to the components are already failed in the scenarios, even though they are not the limiting failure for the component.

Fire Modeling RAI 3 NFPA 805, Section 2.7.3.2, "Verification and Validation," states: "Each calculational model or numerical method used shall be verified and validated through comparison to test results or comparison to other acceptable models."

Section 4.5.1.2, "Fire FPRA" of the LAR states that fire modeling was performed as part of the FPRA development (NFPA 805 Section 4.2.4.2). Reference is made to Attachment J, "Fire Modeling V&V," for a discussion of the V&V of the fire models that were used.

Furthermore Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805" of the LAR states "Calculational models and numerical methods used in support of compliance with 10 CFR 50.48(c) were verified and validated as required by Section 2.7.3.2 of NFPA 805."

Regarding the V&V of fire models:

a. Attachment J of the LAR does not provide the V&V basis for the following fire models and correlations that were used in the transition:
1. All models and correlations that were used in the development of the GFMTs.
2. Method to determine the "heat soak" time in the method of McCaffrey, Quintiere, and Harkleroad hot gas layer calculations for naturally vented compartments, described in Section 3.3.1.1 of BNP-MECH-HGL-0001, Revision 1.

Enclosure 2 Page 16 of 20

3. Method to determine the "heat soak" time in Beyler's HGL calculations for closed compartments, described in Section 3.3.2.1 of BNP-MECH-HGL-0001, Revision 1.
4. Method to account for the "cable thermal endurance duration" in the calculation of the thermal damage time of cables above a burning electrical cabinet, described in HNP-M/MECH-1 194.
5. Method to determine the horizontal ZOI described in HNP-M/MECH-1 129, Revision 0.
6. Method to determine the radiant energy target damage profile described in NED-M/MECH-1007, Revision 0.
7. FDS, in particular the version(s) that were used in the analyses in support of the NFPA 805 transition.

Revise Attachment J to the LAR to include a discussion of the V&V basis of these models and correlations, and describe their application in the transition at BNP.

b. Provide technical details to demonstrate that fire models that were used in the NFPA 805 transition have been applied within the validated range of input parameters, or to justify the application of the model outside the validated range reported in NUREG-1 824 or other V&V basis documents.
c. The discussion for the flame height calculation in Table J-1 states incorrectly that it "is used in both the CFAST and FDS." Revise the discussion to correct this error.

Response

a. The fire modeling V&V analysis is documented in Calculation OFP-1212, Verification and Validation of Fire Models Supporting the Brunswick Nuclear Plant (BNP) Fire PRA. This analysis covers all the fire modeling, including Fire Dynamics Simulator (FDS), CFAST in the appropriate versions, and hand calculations used in the BSEP FPRA. The V&V study has been developed following the guidance available in NUREG-1824, Verification and Validation of Selected Fire Models for Nuclear PowerPlant Applications, and NUREG-1 934, NuclearPowerPlant Fire Modeling Application Guide (NPPFIRE MAG) -

FinalReport. Attachment J to the NFPA 805 LAR has been revised to summarize the technical material contained in Calculation OFP-1212. A copy of the revised Attachment J is provided as Enclosure 8 of this letter.

b. The fire modeling V&V analysis is documented in Calculation OFP-1212, Verification and Validation of Fire Models Supporting the Brunswick Nuclear Plant (BNP) Fire PRA. This analysis covers all the fire modeling, including FDS, CFAST in the appropriate versions and, hand calculations used in the BSEP FPRA. The V&V study has been developed following the guidance available in NUREG-1 824 and NUREG-1 934. Consequently, models that are used outside the validation range reported in Table 2-5 of NUREG-1934 have been identified. Technical justifications have been provided in Calculation OFP-1212 for fire modeling analysis applied in fire scenarios where the results are outside the validated range.
c. The fire modeling V&V analysis is documented in Calculation OFP-1212, Verification and Validation of Fire Models Supporting the Brunswick Nuclear Plant (BNP) Fire PRA.

Attachment J to the NFPA 805 LAR has been revised to summarize the technical material contained in calculation OFP-1212. The revision corrects the error identified in

Enclosure 2 Page 17 of 20 the request for additional information Fire Modeling RAI 3C. A copy of the revised Attachment J is provided as Enclosure 8 of this letter.

Fire Modeling RAI 4 NFPA 805, Section 2.7.3.3, "Limitations of Use," states: "Acceptable engineering methods and numerical models shall only be used for applications to the extent these methods have been subject to verification and validation. These engineering methods shall only be applied within the scope, limitations, and assumptions prescribed for that method."

Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805," of the Transition Report states that "Engineering methods and numerical models used in support of compliance with 10 CFR 50.48(c) were applied appropriately as required by Section 2.7.3.3 of NFPA 805."

Regarding the limitations of use:

a. The parts of the GFMTs that were used in the transition at BSEP have an applicability section, which discusses the limitations of the treatments described in each part. Explain how it was ensured that the GFMTs were used within their limits of applicability, or that uses of the GFMTs outside their limits of applicability were justified.
b. The following documents describe the ZOI, HGL and MCA and related calculations performed in support of the transition at BSEP.
1. BNP-MECH-HGL-001, Rev. 1, "Hot Gas Layer Calculation"
2. BNP-PSA-080, Attachment 7, Rev. 1, "Multi-Compartment Analysis"
3. BNP-PSA-080, Attachment 19, Rev. 1, "Cable Tray Fire Propagation"
4. HNP-M/MECH-1 129, Rev. 0, "Fire Zone of Influence Calculation"
5. HNP-M/MECH-1 194, Rev. 0, "Thermal Damage Time of Cables above a Burning Electrical Cabinet"
6. NED-M/MECH-1006, Rev. 0, "Generic Fire Modeling Treatments"
7. NED-M/MECH-1007, Rev. 0, "Radiant Energy Target Damage Profile" Each of these documents has a section that describes the assumptions and limitations of the calculations presented in the document. Explain how it was ensured that the results of these calculations were used within their limits of applicability, or that uses of these results outside their limits of applicability were justified.
c. Demonstrate that FDS was used within its scope, assumptions and limitations to evaluate the effect of fire-induced flow in an MCC (see HNP-M/MECH-1207, "Fire Induced Flow within a Motor Control Center," Rev. 0).

Response

a. The fire modeling V&V analysis documented in calculation OFP-1212, Verification and Validation of Fire Models Supporting the Brunswick Nuclear Plant (BNP) Fire PRA, includes the generic fire modeling treatments. The V&V study has been developed following the guidance available in NUREG-1824 and NUREG-1934. Technical justifications have been provided for fire modeling analysis applied in fire scenarios

Enclosure 2 Page 18 of 20 where the results are outside the validated range. Attachment J to the NFPA 805 LAR has been revised to summarize the technical material contained in calculation OFP-1212, which includes the V&V for the generic fire modeling material.

b. The fire modeling V&V analysis documented in calculation OFP-1212, Verification and Validation of Fire Models Supporting the Brunswick Nuclear Plant (BNP) Fire PRA, includes the calculations and reports listed in the request for additional information. The V&V study has been developed following the guidance available in NUREG-1824 and NUREG-1 934. Technical justifications have been provided for fire modeling analysis applied in fire scenarios where the results are outside the validated range. Attachment J to the NFPA 805 LAR has been revised to summarize the technical material contained in calculation OFP-1212.
c. The fire modeling analysis documented in HNP-M/MECH-1207, Fire Induced Flow within a Motor Control Center,Revision 0, is not used in support of the BSEP FPRA.

Therefore, the V&V of the fire modeling documented in HNP-M/MECH-1207 is outside of the scope of the V&V analysis in Calculation OFP-1212. In addition, since the fire modeling application documented HNP-M/MECH-1207 is not used in the FPRA, it is not listed and discussed in the revised Attachment J to the BSEP NFPA 805 LAR.

Fire Modeling RAI 6 NFPA 805, Section 2.7.3.5, "Uncertainty Analysis," states: "An uncertainty analysis shall be performed to provide reasonable assurance that the performance criteria have been met."

Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805," of the LAR states that "Uncertainty analyses were performed as required by 2.7.3.5 of NFPA 805 and the results were considered in the context of the application. This is of particular interest in fire modeling and FPRA development."

Regarding the uncertainty analysis for fire modeling:

a. Describe how the uncertainty associated with the fire model input parameters was accounted for in the fire modeling analyses.
b. Describe how the "model" uncertainty was accounted for in the fire modeling analyses.
c. Describe how the "completeness" uncertainty was accounted for in the fire modeling analyses.

Response

a. Parameter uncertainty is addressed by: (1) using conservative inputs, or (2) varying inputs in sensitivity cases.

(1) The fire model input parameters were chosen to be conservative, and each instance is addressed in the individual fire modeling calculations.

(2) The Main Control Room Abandonment Study describes specific treatments of parameter uncertainty throughout the analysis when varying inputs in sensitivity cases. In Report 1RCS04042.103.007-01, Revision 2, the Main Control Room

Enclosure 2 Page 19 of 20 (MCR) abandonment study, heat release rates are varied over the range of the probability distribution described in Appendix E of NUREG/CR-6850. In addition, in the MCR abandonment study, uncertainties are addressed through the following sensitivity analyses documented in Appendix B of 1RCS04042.103.007-01, Revision 2: 1) sensitivity analysis on the use of one room model, 2) on the assumed boundary leakage area, 3) on the assumed burning regime, 4) on the assumed fire base height, 5) on the assumed fire radiant fraction, 6) on the assumed heat of combustion, 7) on the assumed initial ambient conditions, 8) on the main control room window failure, 9) on the assumed main control room door openings, 10) on the HVAC recirculation mode,

11) on the main control room acoustic ceiling failure, and 12) on the door opening time in manual purge mode.
b. The fire modeling V&V analysis is calculation OFP-1212, Verification and Validation of Fire Models Supporting the Brunswick Nuclear Plant (BNP) Fire PRA. This analysis covers all the fire modeling, including FDS, CFAST, and hand calculations used in the BSEP FPRA. The V&V analysis determines whether models are used within their validated range. If the models are found to be used outside of the range, justification is provided describing how the model is utilized conservatively as suggested in NUREG-1934. In summary, the concept of model uncertainty is addressed in the BSEP FPRA through the V&V studies documented in calculation OFP-1212 and the sensitivity analysis supporting the FPRA.
c. Completeness associated with fire models is addressed in the BSEP FPRA within the overall quantification process, as the FPRA is an integrated analysis. Fire Modeling provides inputs to a broad comprehensive Fire PRA which includes modeling of electrical systems, operator actions, and the plant systems and components needed to shut down the plan. One of the first steps in the fire modeling process is to identify the fire scenarios that will be analyzed. In general, three situations can be encountered:
  • Fire scenarios requiring no fire modeling analysis, in which full target sets assigned to physical analysis units are assumed failed by fire at the time of ignition. This is not typically used.
  • Fire scenarios requiring detailed analysis for which fire modeling tools specifically designed for the application are available, and
  • Fire scenarios requiring fire modeling capabilities that are not currently available (i.e., outside the state of the art).

The last two cases listed above require some level of fire modeling analysis. For the case in which fire models are fully capable for addressing the scenario condition (i.e.,

second case listed above), parameter and model uncertainty considerations should address the modeling requirements.

The third case listed above generates the completeness uncertainty situation described in the question. When the fire modeling does not provide a full answer or an answer with sufficient resolution, the scenario definition and target mapping within the FPRA conservatively compensates for the lack of information. The FPRA allows the analyst to conservatively compensate for the lack of fire modeling capabilities outside the fire modeling analysis so that the scenario is properly modeled in the FPRA. Some examples are listed below:

Enclosure 2 Page 20 of 20

  • The determination of time-to-automatic suppression. The detection or suppression activation models may not be fully applicable to some of the postulated scenarios; therefore, as part of the scenario definition, targets are failed intentionally before the automatic suppression is credited.

" Both zones of multi-compartment combinations are failed conservatively when fire modeling propagation calculations from one compartment to another are not conducted.

  • Full main control board panels are failed due to the lack of analytical fire modeling methods, with appropriate V&V studies, to predict flame propagation within a panel.

The examples above illustrate show how the completeness uncertainty associated with fire modeling calculations is addressed "outside of the fire modeling" by conservatively failing targets in the fire scenarios so that the risk contribution is bounding.

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