BSEP 13-0107, Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard (NFPA) 805 (NRC TAC Nos. ME9623 and ME9624)

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Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard (NFPA) 805 (NRC TAC Nos. ME9623 and ME9624)
ML13277A040
Person / Time
Site: Brunswick  Duke Energy icon.png
Issue date: 09/30/2013
From: Hamrick G
Duke Energy Carolinas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 13-0107, TAC ME9623, TAC ME9624
Download: ML13277A040 (11)
v • d • e


Text

Letter Enclosure 4 Contains Security-Related Information - George T. Hamrick Vice President Withhold in Accordance with 10 CFR 2.390 Brunswick Nuclear Plant

~ENERGY, P.O. Box 10429 Southport, NC 28461 o: 910.457.3698 September 30, 2013 Serial: BSEP 13-0107 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-BasedStandard for Fire Protectionfor Light Water Reactor Electric 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-BasedStandard for 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 Standard 805 (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 Standard 805 (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 National Fire Protection Association Standard 805 (TAC Nos. ME9623 and ME9624), dated July 15, 2013, ADAMS Accession Number ML13205A016 When Enclosure 4 is removed, this document is no longer Security-Related A'frs

U.S. Nuclear Regulatory Commission Page 2 of 4

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
7. Letter from George T. Hamrick (Duke Energy) to U.S. Nuclear Regulatory Commission (Serial: BSEP 13-0097), Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard805 (TAC Nos. ME9623 and ME9624), dated August 29, 2013 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 (RAIs) regarding the license amendment request. This letter divided these RAIs into 60-day, 90-day, and 120-day responses.

A tabulation of the individual RAIs from the May 15, 2013, NRC letter and the response submittal dates is provided in Enclosure 1.

During preparation of the 120-day RAI response, Duke Energy determined that additional time was needed to complete revision of the Fire Probabilistic Risk Assessment (FPRA) model and perform review of the new FPRA sensitivity study results. The need for additional time was discussed and agreed to with the NRC in a telephone call on August 26, 2013. As a result, the following RAI responses were delayed until September 30, 2013:

  • Fire Modeling RAI 1E
  • FPRARAI1D
  • FPRARAI1E
  • FPRARAI1H
  • FPRARAI1L
  • FPRARAP11B Duke Energy's responses to those RAls are provided in Enclosure 2. Duke Energy is providing the sensitivity study results 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 ReactorElectric GeneratingPlants, 2001 Edition Main Report and a revision to NFPA 805 Transition Report, Attachment W, Fire PRA Insights. The revised Transition Report is provided as Enclosure 3. The Revised Attachment W is provided as Enclosure 4 and is considered Security-Related Information.

This document contains no new regulatory commitments.

U.S. Nuclear Regulatory Commission Page 3 of 4 Please refer any questions regarding this submittal to Mr. Lee Grzeck, Manager - Regulatory Affairs, at (910) 457-2487.

I declare, under penalty of perjury, that the foregoing is true and correct. Executed on September 30, 2013.

Sincerely, GeorggT. Hamrick

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 Protection for Light Water Reactor Electric Generating Plants, 2001 Edition, Transition Report, September 19, 2013, Main Report Without Attachments
4. Revised NFPA 805 Transition Report, Attachment W, Fire PRA Insights (Security-Related Information - Withhold from Public Disclosure)

U.S. Nuclear Regulatory Commission Page 4 of 4 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 3):

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

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 Fire Modeling 1A, 1E, 1F, 1G, 1H, 2A, 2B, 5A, 5B June 28, 2013) 60-Day Response - PRA Probabilistic Risk 1A, 1B, 1C, 1D, 1F, 1G, 11, 1N, 10, July 15, 2013 Assessment 1 P, 1Q, 1R, 4, 5, 9, 10, 17, 18 (Complete 1 July 15, 2013) 90 Day Response _____..._.....__

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, 1 K, M, 2, 3, 6, 7, 11, 12, 13, 14, 15, July 31, 2013)

Assessment 16 Fire Modeling 1B, 2C, 5C 120 Day Response Fire Protection Engineering 2 Safe Shutdown Analysis 15 August 30, 2013 Probabilistic Risk 1E, 1 H, 1 L, 8 (Complete Assessment August 29, 2013)

Fire Modeling 1C, 1D, 11, 2D, 3, 4, 6 150 Day Response _ _

PRA Sensitivity Results FPRA 1D, FPRA 1 E, FPRA 1 H, September 30, 2013 FPRA 1L, FPRA 3D, FPRA 11 B, FM 1E

Enclosure 2 Page 1 of 6 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., 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 (RAIs) regarding the license amendment request. This letter divided these RAls into 60-day, 90-day, and 120-day responses.

During preparation of the 120-day RAI response, Duke Energy determined that additional time was needed to complete revision of the Fire Probabilistic Risk Assessment (FPRA) model and perform review of the new FPRA sensitivity study results. The need for additional time was discussed and agreed to with the NRC in a telephone call on August 26, 2013. As a result, the following RAI responses were delayed until September 30, 2013:

  • Fire Modeling RAI 1E
  • FPRARAI1D
  • FPRARAI1E
  • FPRARAI1H
  • FPRARAI1L
  • FPRARAI 3D
  • FPRARA111B Duke Energy's responses to those RAls are provided below. Duke Energy is providing the sensitivity study results via an update to the Brunswick NFPA 805 Transition Report, Transition to 10 CFR 50.48(c) - NFPA 805 Performance-BasedStandard for Fire Protection for Light Water ReactorElectric GeneratingPlants, 2001 Edition Main Report and a revision to NFPA 805 Transition Report, Attachment W, Fire PRA Insights. The revised Transition Report and the Revised Attachment W are provided as Enclosures 3 and 4 to this letter.

Fire Modeling Requests for Additional Information 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

Enclosure 2 Page 2 of 6 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

As described in the 120-day response for Fire Modeling RAI 1E, an additional screening process was completed to identify ignition sources where the effect of the increased heat release rate (HRR) from intervening combustibles may result in additional risk contributing fire scenarios (i.e., reference Evaluation Engineering Change 93644). That screening process excluded scenarios where the time to formation of a hot gas layer was relatively short, scenarios where the conditional core damage probability (i.e., conditional large early release probability, CLERP) for the ZOI was already 1.0, scenarios where the ignition frequency was less than 1 E-8/year, and scenarios where the product of the zone of influence (ZOI) ignition frequency and the hot gas layer (HGL) conditional core damage probability (CCDP) was less than 1 E-8/year. This bounding approach screened out most scenarios and showed only a few scenarios in the cable spreading rooms had the potential to be significant risk contributors due to this effect. Although these scenarios might also be eliminated by a less conservative and more realistic approach, even a more realistic approach is unlikely to yield any new risk insight.

Fires in those areas were already recognized as significant contributors to risk, and certain planned modifications (e.g., area-wide incipient detection, and selected use of Electrical Raceway Fire Barrier System (ERFBS)), along with existing cable tray solid bottoms in most locations, are available to mitigate risk. As such, no additional sensitivities identifying impact on core damage frequency (CDF), ACDF, large early release frequency (LERF) and ALERF are provided for this specific RAI, as was previously mentioned in the 120-day response.

Enclosure 2 Page 3 of 6 Fire Probabilistic Risk Assessment Requests for Additional Information FPRA RAI ID F&O 1-19 against FSS-A1 (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

This response supplements the information provided in Duke Energy's letter dated July 15, 2013 (i.e., ADAMS Accession Number ML13205A016), to FPRA RAI 1D. As described in the previous response to FPRA RAI 1D, the bounding sensitivity study described in the response to Fact and Observation (F&O) 1-19 was performed as an ad-hoc white paper during interaction with the Peer Review Team and was not captured as plant records. In response to FPRA RAI 1D, a new sensitivity study was created, as documented in Section 4.12 of Calculation BNP-PSA-095, BNP Fire PRA - Sensitivities, to provide the following results using the current FPRA model and the bounding approach of the previous sensitivity study. Like the original sensitivity study, the recreated sensitivity study estimated impacts on CDF and LERF, but did not include ACDF and ALERF.

Unit 1 Unit 2 CDF [/yr] LERF [/yr] CDF [/yr] LERF [/yr]

Internal Events plus External Flooding and 14E-05 6.2E-07 14E-05 6.2E-07 High Winds Fire[1 ] 7.2E-05 8.1E-06 6.4E-05 7.6E-06 Fire - Recovery Actions[21 1.OE-06 1.OE-07 1.OE-06 1.OE-07 Total 8.7E-05 8.9E-06 7.9E-05 8.3E-06

[1] Fire results do not credit control room abandonment for loss of control sequences.

[2] Values are for recovery actions associated with control room abandonment due to environmental reasons.

A manual summation may differ from the Total due to rounding in the last digit.

Enclosure 2 Page 4 of 6 FPRA RAI 1E 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.

Response

The impact of transient fires on sensitive electronics is not specifically modeled in the FPRA.

However, the guidance in the draft Frequently Asked Question (FAQ) 13-0004 is considered applicable to BSEP. The requested sensitivity study was performed as documented in Section 4.10 of Calculation BNP-PSA-095, and the estimated impacts on CDF, LERF, ACDF, and ALERF are provided in Section 4.8.3.8 of the Brunswick NFPA 805 Transition Report, Transitionto 10 CFR 50.48(c) - NFPA 805 Performance-BasedStandardfor Fire Protectionfor Light Water Reactor Electric GeneratingPlants, 2001 Edition, Main Report.

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 8th 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

The requested sensitivity study was performed as documented in Section 4.6 of Calculation BNP-PSA-095, and the estimated impacts on CDF, LERF, ACDF, and ALERF are provided in Section 4.8.3.7 of the Brunswick NFPA 805 Transition Report, Transition to 10 CFR 50.48(c) -

NFPA 805 Performance-BasedStandard for FireProtection for Light Water ReactorElectric GeneratingPlants, 2001 Edition, Main Report.

Enclosure 2 Page 5 of 6 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

Section 4.13 of Calculation BNP-PSA-095 evaluated the treatment of smoke damage, as discussed in Appendix T of NUREG/CR-6850, and the possible effects on the quantification results in the BSEP FPRA. The practical implications 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 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.

Section 4.8.3.9 of the Brunswick NFPA 805 Transition Report, Transitionto 10 CFR 50.48(c) -

NFPA 805 Performance-BasedStandardfor Fire Protection for Light Water ReactorElectric GeneratingPlants,2001 Edition, Main Report, provides a summary of the results of the smoke damage evaluation.

FPRA 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.

Enclosure 2 Page 6 of 6

Response

The requested sensitivity study was performed as documented in Section 4.11 of Calculation BNP-PSA-095, and the estimated impacts on CDF, LERF, ACDF, and ALERF are provided in Section 4.8.3.11 of the Brunswick NFPA 805 Transition Report, Transition to 10 CFR 50.48(c) -

NFPA 805 Performance-BasedStandard for FireProtection for Light Water Reactor Electric GeneratingPlants, 2001 Edition, Main Report.

FPRA RAI II B 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

As detailed in Attachment 16 of Calculation BNP-PSA-080, BNP File PRA - Quantification, whether single-bundle or multi-bundle was used for MCR cabinets was determined from the source walkdowns for all cabinets in the Main Control Room (MCR). However, as a special case, the Main Control Boards (MCBs) were treated as a single-bundle, rather than a multi-bundle, because the evaluation of the risk for MCR abandonment is relatively simplistic and a multi-bundle treatment was considered to produce overly conservative results for the MCBs.

A more realistic treatment would include credit for the close and constant proximity of reactor operators, the expected slow growth resulting from low energy ignition sources inside the cabinets, and a high probability of fire detection by the control staff such that early suppression will occur before a large multi-bundle fire could develop. Consequently, to achieve more realistic results within the parameters of the analysis described in Attachment 16, the MCBs were approximated as single-bundles.

A sensitivity study which evaluated the treatment of the MCBs as multi-bundle fires was performed as documented in Section 4.8 of Calculation BNP-PSA-095, and the estimated impacts on CDF, LERF, ACDF, and ALERF are provided in Section 4.8.3.10 of the Brunswick NFPA 805 Transition Report, Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standardfor Fire Protection for Light Water ReactorElectric Generating Plants,2001 Edition Main Report.

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