BSEP 14-0035, Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard 805

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Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard 805
ML14118A105
Person / Time
Site: Brunswick  Duke Energy icon.png
Issue date: 04/10/2014
From: Hamrick G
Duke Energy Carolinas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML14118A104 List:
References
BSEP 14-0035, TAC ME9623, TAC ME9624
Download: ML14118A105 (18)
v • d • e


Text

George T. Hamrick

  • DUKEENERGY, Letter Enclosure 2 Contains Security-Related Information -

Withhold in Accordance with 10 CFR 2.390 Vice President Brunswick Nuclear Plant P.O. Box 10429 Southport, NC 28461 o: 910.457.3698 April 10, 2014 Serial: BSEP 14-0035 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 Operating 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 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 GeneratingPlants (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 Protection for Light Water Reactor Electric GeneratingPlants (2001 Edition), dated December 17, 2012, ADAMS Accession Number ML12362A284

3. Letter from Farideh Saba (USNRC) to George T. Hamrick (Duke Energy Progress, Inc.), Second Request for Additional Information Regarding Voluntary Risk Initiative National Fire ProtectionAssociation Standard 805 (TAC Nos. ME9623 and ME9624), dated February 12, 2014, ADAMS Accession Number ML14028A178
4. Letter from George T. Hamrick (Duke Energy Progress, Inc.) to U.S. Nuclear Regulatory Commission (Serial: BSEP 14-0029), Response to Additional Information Regarding Voluntary Risk Initiative National Fireprotection Association Standard805 (NRC TAC Nos. ME9623 and ME9624), dated March 14, 2014, ADAMS Accession Number ML14079A233 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., submitted a license Ct' When Enclosure 2 is removed, this document is no longer Security-Related

U.S. Nuclear Regulatory Commission Page 2 of 3 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.

On February 12, 2014 (i.e., Reference 3), the NRC provided a request for additional information (RAI) regarding the license amendment request. During a telephone call conducted with the NRC staff on January 14, 2014, Duke Energy agreed to submit responses to the RAIs on the following schedule:

RAI Planned Set RAI Number Response Date 1 1.d.01, 1.f.ii.01, 1.f.iii.01, 14.01, Submitted by letter 15.01, 16.01, 18.g.01, 22, and 24 dated March 14, 2014 (i.e.,

Reference 4) 2 1.1.01, 6.01, 8.01, and 23 April 11, 2014 Duke Energy's responses to the second set of RAIs are provided in Enclosure 1 of this letter. of this letter provides an updated copy of the License Amendment Request (LAR)

Attachment S. This supersedes, in its entirety, the Attachment S previously provided as part of Duke Energy's letter dated March 14, 2014.

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.

I declare, under penalty of perjury, that the foregoing is true and correct. Executed on April 10, 2014.

Sincerely, George T. Hamrick

Enclosures:

1. Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard 805
2. Updated License Amendment Request Attachment S, Modifications and Implementation Items (Security-Related Information - Withhold from Public Disclosure)

U.S. Nuclear Regulatory Commission Page 3 of 3 cc (with Enclosures 1 and 2):

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. Siva P. Lingam (Mail Stop OWFN 8G9A) (Electronic Copy Only) 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 Enclosure 1 only):

Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 Mr. W. Lee Cox, Ill, Section Chief (Electronic Copy Only)

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 15 Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard 805 By letter dated September 25, 2012, as supplemented by letter dated December 17, 2012, Duke Energy Progress, Inc., 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.

On February 12, 2014, the NRC provided a request for additional information (RAI) regarding the license amendment request. During a telephone call conducted with the NRC staff on January 14, 2014, Duke Energy agreed to submit responses to the RAIs on the following schedule:

RAI Planned Set RAI Number Response Date 1 1.d.01, 1.f.ii.01, 1.f.iii.01, 14.01, 15.01, Submitted by 16.01, 18.g.01, 22, and 24 letter dated March 14, 2014 2 1.1.01,6.01, 8.01, and 23 April 11, 2014 Duke Energy's responses to the second set of RAIs are provided below.

Probabilistic Risk Assessment (PRA) Request for Additional Information (RAI) 1.1.01 By letter dated July 15, 2013, the licensee responded to PRA RAI 1.i and explained that the state of knowledge correlations (SOKC) were omitted for fire ignition frequencies and nonsuppression probabilities, and implied that SOKC was only applied within cutsets. This response also explains that an SOKC multiplier was calculated using the standard deviation for hot short probabilities. Provide an estimate of the impact of considering SOKC for fire ignition frequency, nonsuppression probability, and circuit failure probabilities on the risk estimates (i.e.,

CDF, LERF, delta (A)CDF, ALERF) across cutsets.

Response

The impact of considering SOKC for fire ignition frequency, non-suppression probability, and circuit failure probabilities on the risk estimates (i.e., Core Damage Frequency (CDF), Large Early Release Frequency (LERF), ACDF, and ALERF) was estimated within and across cutsets (i.e., Engineering Change (EC) 96040). This SOKC analysis used the cutsets where no credit for control power transformer (CPT) circuit failure probabilities was applied. The increase in CDF and LERF where the CPT credit is removed is only slightly higher (i.e., <1E-7 for CDF,

<1 E-8 for LERF) than the base case where CPTs were credited. Use of the no CPT credit cutsets permitted a simultaneous analysis of the combined effects of no CPT credit and SOKC to identify potential synergisms.

The individual merged cutset files for CDF and LERF were modified to correlate the circuit failure data, the fire ignition source bin data, and the non-suppression type data across cutsets.

The Electric Power Research Institute (EPRI) UNCERT program was used to perform an uncertainty analysis on the modified cutset files using a sampling approach (i.e., Monte Carlo

Enclosure 1 Page 2 of 15 Method) to determine the effect of SOKC on the mean values for CDF, LERF, ACDF, and ALERF. The overall cutset data analysis results were:

SOKC Results for Delta CDF and Delta LERF Results from Mean Value - No CPT Credit Unit I Unit 2 ACDF [lyr.] ALERF [/yr.] ACDF [/yr.] ALERF [/yr.]

Variances From Deterministic 1.2E-06 2.E-08 3.OE-06 2.5E-08 Requirements (VFDRs)

Recovery Actions[1 ] 9.1 E-07 9.1 E-08 9.1 E-07 9.1 E-08 Total 2.1 E-06 1.2E-7 3.9E-06 1.2E-07

[1] Values are for recovery actions associated with control room abandonment due to environmental reasons and address those actions away from the remote shutdown panel.

SOKC Results for Total CDF and LERF from Mean Value - No CPT Credit Unit I Unit 2 CDF [/yr.] LERF [/yr.] CDF [Iyr.] LERF [/yr.]

Internal Events plus External 14E-05 6.2E-07 14E-05 6.2E-07 Flooding and High Winds Mean Firel1 ] 2.1E-05 4.2E-06 2.1E-05 4.1E-06 Fire - Recovery Actions[21 1.OE-06 1.OE-07 1.OE-06 1.OE-07 Total 3.6E-05 4.9E-06 3.6E-05 4.8E-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.

Compared to the corresponding risk metrics reported in LAR Section W.2.2 of the September 30, 2013, letter (i.e., ADAMS Accession Number ML13277A040), the combined effects of the SOKC and no CPT credit were <1 E-6 and <1 E-7, for CDF and LERF, respectively, and likewise for ACDF and ALERF, respectively. The changes in CDF and LERF are the impacts of the sensitivity analysis on the base model. The changes in ACDF and ALERF are the impacts of the sensitivity analysis on the risk associated with VFPRs from the compliant case.

For Unit 1, the changes in CDF and LERF were 3.OE-7 and -1.OE-8, respectively, while the changes in ACDF and ALERF were 2.OE-8 and 1.1 E-8, respectively. For Unit 2, the changes in CDF and LERF were 8.OE-7 and 3.OE-8, respectively, while the changes in ACDF and ALERF were 3.OE-7 and 2.OE-8, respectively. The small negative change for Unit 1 LERF resulted from the UNCERT use of the Monte Carlo simulation because the removal of CPT credit had no change in Unit 1 LERF.

In addition to the above tabulated effects of considering the SOKC for fire ignition frequency, non-suppression probability, and circuit failure probabilities, the effects were examined

Enclosure 1 Page 3 of 15 separately for these parameters. When examined separately for the circuit failure probability events, with multiple correlated events within the same cutsets, the effect of SOKC was measurable. However, the effect only accounted for 0.2% increase in the total Unit 2 LERF and was smaller for Unit 1 CDF, Unit 2 CDF, and Unit 1 LERF. The effect was too small to be observable on the results tabulated above. When examined separately for fire ignition source bins, the SOKC had no-measurable effect seen in the simulation mean values from within and across cutsets. Likewise, for the non-suppression probability bins, the effects of SOKC within and across cutsets were not measurable.

The overall results of the uncertainty analysis (i.e., simulation means) showed that the effect of SOKC on the Brunswick Fire PRA are minimal to non-measurable with regard to the impact on the risk estimates (i.e., CDF, LERF, ACDF, and ALERF). The following Figures 1 through 4 from EC 96040 provide the data simulation results of the of parametric uncertainty analysis.

Enclosure 1 Page 4 of 15 BNP UNIT 1 CDF UNCERT ANALYSIS FOR SOKC IN AND ACROSS CUTSETS - NO CPT CREDIT hIdudes Addiin of Fre Iitnlon Soame Bin, Manual Non-Suppression Bin, and Not Short Probabiirty CoreL Dat Reported Fire CDF: 2.06E-05 Moan o :0 OE.05 Merged CSED Fire CDF: 2.08E-05

  • 5.[(

50*- X : 596E-6 1.43E-05 Uncert Pt. Est Fire CDF: 2.08E-0S u ,-I !4E-M MnerWanFire CDI:: 2-07E-(D5 11sampiu :2C00W NC Ab~andonmentActions CD:: L03E-406 ITotal Wean FPRA CD: Z.IE-OS Statistical Results 5% Conf. mean 95%Conf, Probability Density Function Point Est 2.08E-OS and Mean 2.07E-05 2.07E-05 2.08E-05 Cumulative Density Function 5% 5.96E-06 5.98E-06 6.O1E-06 Median 1.42E-05 1.43E-05 1.43E-05 95% 5JOE-05 5.74E-05 5.79E-OS StdDev 1.92E-05 Skewness 3.29E+00 Kurtosis L87E+01 BNP UNIT 1 COMPUANT CDF UNCERT ANALYSIS FOR SOKC IN AND ACROSS CUTSETS - NO CPT CREDIT Whndes Addition of Fre %ntion Suuom Min, I an aotShort PrbablltyvCaweled Duta Reported Fire CDF: 1.94E-05 Merged CSED Fire CDF: 1.95E-05 LUncert Pt. Est. Fire CDF: 1.95E-O UnMtean Fire CDI:: LB5E-0SJ Figure 1 - Brunswick Unit 1 CDF Base and Compliant Uncertainty Analyses

Enclosure 1 Page 5 of 15 BNP UNIT 2 CDF UNCERT ANALYSIS FOR SOKC IN AND ACROSS CUTSETS - NO CPT CREDIT Indudes Additon of Fire Ilpiion Source Bin, Manual Non-Suppression Bin, and Hot Short Proba Wty Correlat*d Data Reported Fire CO:: 2.02E-05 Mean-o :Z07E05 Merged CSED Fire CDF: 2.08E-05t 5. [ :4.20E6 50 x :1.24E05 Uncert Pt. Est. Fire CDF: 2.0SE-05 95%.  : &4?E-UIcert MWan Fire COP: Z07E-05 # samgim :20000 CR AbntdomnAions COP: L03E.C0j

~Total Weain FPRA COP: ZVE-IB Statistical Results 5% Conf. mean 95% Conf. Probability Density Function Point Est 2.08E-05 and Mean 2.06E-05 2.07E-05 2.OSE-05 Cumulative Density Function 5% 4.18E-06 4.20E-06 4.21E-06 Median 1.24E-05 1.24E-05 1.25E-05 95% 6L41E-05 6.47E-O5 6.52E-05 StdDev 2.27E-05 Skewness 3.25E+00 Kurtosis 1.74E+01 1.E15 114 Fmwuenoy i Pwbabf BNP UNIT 2 COMPLIANT CDF UNCERT ANALYSIS FOR SOKC IN AND ACROSS CUTSETS - NO CPT CREDIT Indudes Addition of Fire Iinition Source Bin, Manual Non-SupWession Bin, ad Hot Short Probability Corenatd Dot.

Reported Fire CDF: 1.7SE-05 Men- o : 1.771E Merged CSED Fire CDF: 1.78E-05 Sf-  : Z45E.06 Uncert Pt. Est. Fire CDF: L.78E-05 50%-x 1: 453 U#cert Wan FlrpCO: L77E405 I tales :200000 KOCA Non-FaoviwywActonsCOP: L20E-0 j Trotal anComplimnt CDP: 17SE-05i Statistical Results 5% Conf. mean 95% Conf Probability Density Function Point Est 1.78E-05 and ean 76E-05 1.77E-05 1.78E-05 Cumulative Density Function 5% e644E-06 1.47E-05 2.47E-05 Median 9.97E-06 1.00-05 1.01E-05 95% 5.78E-05 5.83E-05 5.88E-05 StdDev 2.1OE-05 Skewness 3.21E+00 Kurtosis 1.65E+01 1.E.6 E-J5 1.E 1.E-4 FE4an,/Poa~

Figure 2 - Brunswick Unit 2 CDF Base and Compliant Uncertainty Analyses

Enclosure 1 Page 6 of 15 BNP UNIT 1 LERF UNCERT ANALYSIS FOR SOKC IN AND ACROSS CUTSETS - NO CPT CREDIT IndudesAdimde efi lniRel n Summ lbn Manm idNon-Suppresdni Bin. and Hot Short Proabahfty Coweht~d Deb Reported Fire LERF: 4,26E-06 IMon-*

SX. [ .24E.0

%7%£4)7 Merged CSED Fire LERF: 427E-06 AL50%-x :2 908E-f0 Uncert Pt. Est Fire LERF: 4,27E-06 5%-] :1.26E.05 Uncert Won Fire LEWF: 424E-061 MICR Ahmndorwnt Acins 1W:F L03EO0 Total MWan FPRA LEW: 434E-06 Statistical Results S% Conf. mean 95% Conf. Probablity Density Function Point Est 4t27E-06 and Mean C22E-06 4.24E06 426E-06 Cumulative Density Function 5% &65E-07 6.70E-07 A7SE-07 Median 2.7KE-06 28E-06 ZSlE-06 95% L2SE-O L26E-0S L27E-OS Stdlev 4&58E-06 Skewness 4.99E4*0 Kurtosis 1,02E+02 1.E-7 1.gE4 1.E-5 1.E4 Fwwwmnq / Ftobakily BNP UNIT 1 COMPUANT LERF UNCERT ANALYSIS FOR SOKC IN AND ACROSS CUTSETS - NO CPT CREDIT lndudinsAddmcmof Rn k.Un Sourm lab. mmmd Nio.-SUpmmdm 111n, md Not Short Probabilty Conerohd Dab Reported Fire LERF: 4L24E-06 Memo 4.22E4 Merled CSED Fire LERF: 42SE-06 ,'h Unrert Pt. Est Fire LERF: 426E-06 50 279E.C6 Jnce t Mean Fire LWR: 0.22E44 KiMR Nom-isteame Acdlns LEW: 12"U4 Total Mean Compliant LEW: 423E.U Statistical Results S%Conf. mean 9S% Conft rProbabilityiDensity Functlon Point Est 4.26E-06 and Mean &20E-06 422E-O6 424E-06 Cumulative Density Function S% &6SE47 &70E-07 67SE-07 Median 2.78E-06 2.79E-06 ZILE-06 9S% L24E-0E l.2SE-M 126E-C StdDev 4.SSE-06 Skewness 4.42E4,0 Kurtosis 5.2E401 BNP 1 DELTA MEAN FIRE LERF RISK It"I 11E-7 1.E-4 Fr..,acy I Robabi 1, Figure 3 - Brunswick Unit 1 LERF Base and Compliant Uncertainty Analyses

Enclosure 1 Page 7 of 15 BNP UNIT 2 LERF UNCERT ANALYSIS FOR SOKC IN AND ACROSS CUTSETS - NO CPT CREDIT Includes Addition of Fire Illnition Source Bin, Manual Non-SuppressIon Sin, and Hot Short Proabiliat Corelated Data Reported Fire LERF: 4.04E-06 Men.o :4.07E-06 Merged CSED Fire LERF: 4.08E-06 5.[  : &42E-07 Uncert Pt. Est Fire LERF: 4.08E-05 50U-x- Z76E-06 nr.-95%- ... 1.16E*5

  1. I uflikt 20OCM MswR Abndn :n Acios 71:-07 Statistical Results 5% Conf. mean 95% Conf. Probability Density Function Point Est 4.0SE-06 and Mean 4.0SE-06 4.07E-06 4.,0E-06 Cumulative Density Function 5% 6.37E-07 6.42E-07 6.48E-07 Median 2.74E-06 2.76E-06 2.77E-06 95% LI5E-05 1. 1M-05 L17E-05 StdDev 4.90E-06 Skewness 177E401 Kurtosis L39E403 1.E7 1E-G11E-5 11E4 Fmqasnc I Pmbaby BNP UNIT 2 COMPLIANT LERF UNCERT ANALYSIS FOR SOKC IN AND ACROSS CUTSETS - NO CPT CREDIT kKludes Addition of nrklpinion Source E 111,n-Summsion n ul M Bind Hot Short P ty C Oda Reported Fire LERF: 4.03-06 MGM-* :k40Ea Merged CSED Fire LERF: 4.07E-06 5%. :6.34E47 Uncert Pt. Est Fire LERF: 4.07E-06 95-]  : 1.165 Statistical Results 5% Conf. mean 95% Conf. Probability Density Function Point Est 4.07E-06 and Mean 4.03E-06 4.04E-06 4.07E-06 Cumulative Density Function 5% 6.28E-07 6.34E-O7 6.39E-07 Median 2.73E-06 2.74E-06 2.76E-06 95% LUSE-O5 1.16E-05 1.17E-0S StdDev 4,62E-06 Skewness 7.40E400 Kurtosis 162EO2, 1.E-7 1.E6 11-.5 11E4 FoQgnc" / Pwbalft Figure 4 - Brunswick Unit 2 LERF Base and Compliant Uncertainty Analyses

Enclosure 1 Page 8 of 15 PRA RAI 6.01 By letter dated July 31, 2013, the licensee responded to PRA RAI 6 and explained that for main control boards (MCBs) with incipient detection that the risk from "self' scenarios was assumed to be negligible and not modeled in the FPRA. By letter dated September 30, 2013, the licensee presented the results of a sensitivity study that credited the currently installed in-panel ion smoke detection system rather than incipient detection, and applied the NUREG/CR 6850, "EPRI [Electric Power Research Institute]/NRC-RES [Nuclear Regulatory Research] Fire PRA Methodology for Nuclear Power Facilities," approach to determine the frequency of "self-fires" in the MCB. Use of the NUREG/CR-6850 Appendix L approach appears not to have been previously used in the FPRA (i.e., the baseline approach excludes in-cabinet MCB fires based on installation of incipient detection), and use of this approach produces significant differences in accident sequence risk estimates (e.g., the CDF increases from 3.5E-5/yr for Units 1 and 2 to 6.3E-5/yr and 6.6E-5/yr, respectively for each unit).

The licensee's application of the Appendix L method appears to deviate from the NUREG/CR-6850 guidance, based on the description presented in the FPRA Sensitivities report (i.e., BNP-PSA-095, Revision 1) that was issued after the audit. The initiating event fire frequencies for 27 MCB "self' scenarios presented in Table 20 of the Fire PRA Sensitivities report were determined by dividing the EPRI MCB Bin 4 fire frequency by 27, leading to a frequency of 6.104E-5/yr per MCB. This conflicts with the counting guidance presented in Section 6.5.6 of NUREG/CR-6850, which presumes that counting will yield only one or two MCBs. Furthermore, when applying the Appendix L method, the frequency of a scenario involving specific target damage in the MCB should be determined by multiplying the probability of target damage, as defined by Figure L-1 of NUREG/CR-6850, by the entire MCB fire frequency. When using this approach partitions or segmentation cannot be used to justify subdividing the MCB fire frequency, although they might be used to preclude certain scenarios involving targets separated by partitions.

Use of the NUREG/CR-6850 Appendix L modeling approach meets the definition of a model upgrade as defined by American Society of Mechanical Engineers (ASME) and American Nuclear Society (ANS) ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications." In light of the fact that use of incipient detection to preclude internal MCB cabinet fires is not consistent with Frequently Asked Question (FAQ) 08-0046 and implementation of Appendix L appears to deviate from NUREG/CR-6850 guidance, the Appendix L analysis of the MCB needs to be re-performed to meet guidance in NUREG/CR-6850 by applying the whole MCB fire frequency to each MCB scenario. In addition, a focused scope peer review on the use of the NUREG/CR-6850 Appendix L modeling approach is needed prior to using the FPRA for self-approval.

Provide the reanalysis of the MCBs as part of PRA RAI 23, and provide a proposed implementation item, to address that a focused-scope peer review will be performed, and any findings will be resolved, before self-approval of post transition changes.

Response

As described in EC 95972, the Appendix L analysis of the MCB was re-performed as a sensitivity analysis to meet the guidance in NUREG/CR-6850. Target sets were identified on the MCB. Damage probabilities were developed using Figure L-1 of Appendix L. The calculated

Enclosure 1 Page 9 of 15 Conditional Core Damage Probability (CCDP)/Conditional Large Early Release Probability (CLERP) for each target set was multiplied by the whole MCB Bin 4 ignition frequency to calculate a CDF/LERF for each target set. No credit was taken for incipient detection. The resulting CDF, LERF, ACDF, and ALERF were: ,

Delta CDF and Delta LERF from Appendix L for MCBs Unit I Unit 2 ACDF I/yr.] ALERF [/yr.] ACDF [Iyr.] ALERF [Iyr.]

VFDRs 1.2E-06 1.3E-08 2.7E-06 5.2E-09 Recovery Actions[1 ]. 9.1 E-07 9.1 E-08 9.1 E-07 9.1 E-08 Total 2.1 E-06 1.OE-07 3.6E-06 9.6E-08 1] Values are for recovery actions associated with control room abandonment due to environmental reasons and address those actions away from the remote shutdown panel.

CDF and LERF from Appendix L for MCBs Unit I Unit 2 CDF [/yr.] LERF [/yr.] CDF [Iyr.] LERF [I/yr.]

Internal Events plus External 1AE-05 6.2E-07 I4E-05 6.2E-07 Flooding and High Winds Fire[1 ] 2.5E-05 3.9E-06 2.5E-05 3.6E-06 Fire - Recovery Actions[21 1.OE-06 1.OE-07 1.OE-06 1.OE-07 Total 4.OE-05 4.6E-06 4.OE-05 4.3E-06

[1]

'ir I r1 i ,Ire Hr% nI t,-rg

'1" r 'if Ir-nfrrl rrIrJm h '.-nJnnnm,-mnf fnr IeJo 1f -,r1nIrrdl 09M9 HrnJn*n ,

[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.

Compared to the corresponding risk metrics reported in LAR Section W.2.2 by letter dated September 30, 2013 (i.e., ADAMS Accession Number ML13277A040), the effects of the Appendix L analysis were <1E-5 and <1E-6, for CDF and LERF, respectively. The changes in CDF and LERF are the impacts of the sensitivity analysis on the base model. There was no change in ACDF or ALERF because the sensitivity analysis did not impact the treatment of VFDRs in the compliant case. Delta CDF and delta LERF were not affected by the Appendix L analysis of the MCBs because the only VFDRs associated with the control room involved MCR abandonment as part of the alternate safe shutdown strategy. The statistical modeling of Appendix L was not applicable because the MCR abandonment was only credited in the Fire PRA for habitability concerns, which was based on fire modeling, and not for loss of control. For

Enclosure 1 Page 10 of 15 Unit 1, the changes in CDF and LERF were 4.9E-6 and -4.1E-7, respectively. For Unit 2, the changes in CDF and LERF were 4.7E-6 and -3.9E-7, respectively.

The increase in CDF with a decrease in LERF is attributed to differences in the scenarios evaluated with a fire modeling approach for the base model compared to those evaluated with the statistical modeling approach for Appendix L. Small fire scenarios (i.e., where the fire is contained within the cabinet) were not quantified in the base model based on early detection and suppression of small fires due to in-cabinet incipient detection. However, where early detection or suppression failed, large fire scenarios (i.e., with the potential to damage external targets within the zone of influence (ZOI) in addition to MCB internal targets) were quantified.

These large MCB fires tended to have both high CCDP and high CLERP values due to a large number of components failing. In the Appendix L sensitivity, no credit was taken for the incipient detectors, and small fires were postulated for the identified internal target sets but were contained within the MCB. While some of these small fires had high CCDPs, the components in the target sets often did not directly cause a large early release. To cause a LERF, these cases usually required either additional random failures or an expanded target set distance (i.e.,

reduced likelihood), both of which contributed to LERF being lower, sometimes significantly lower, than CDF.

Implementation Item #15 will be added to Table S-2 in the LAR to read:

Item Unit Description LAR Section / Source Prior to use of the FPRA to support self-approval of post-transition changes to the main control board, 15 1 2 where the change has been demonstrated to have RAI PRA 6.01 more than a minimal risk impact, a focused scope peer review will be performed for the Appendix L sensitivity analysis and any findings will be resolved.

An updated copy of LAR Attachment S is provided in Enclosure 2 of this letter. This supersedes, in its entirety, the version of Attachment S provided as part of Duke Energy's letter dated March 14, 2014.

PRA RAI 8.01 By letter dated August 29, 2013, the licensee responded to PRA RAI 8 and explained that the conditions under which a fire has the potential to fail additional equipment due to increased room temperature is "very limited," but acknowledged that such a condition might exist in an enclosed room with an ignition source. The response further argues that cables for equipment in such an enclosed space may already be in the zone of influence of the ignition source. The heat-up analysis report (Attachment 21 to BNP-PSA-067), in support of the success criteria for the internal events PRA, identifies the following maximum ambient operating temperatures for temperature limited equipment:

  • 131°F for the limiting component in the Battery Room

" 235 0 F for the limiting component in the Core Spray Room

" 165°F for the limiting component in the High Pressure Coolant Injection Room

Enclosure 1 Page 11 of 15 Discuss whether these components can be affected by fire producing ambient operating temperatures above the maximum temperature for the component. If so, provide the impact on the risk estimates (i.e., CDF, LERF, ACDF, and ALERF).

Response

In general, additional component failures due to fires producing ambient operating temperatures above the maximum temperatures, for the components located outside the ZOI of the fire do not contribute significantly to risk for several reasons.

1) The growth of the fire scenarios is limited to a particular range of sizes. The fire must grow sufficiently large to release the heat necessary to raise the room temperature to the applicable component limit but not so large as to impact the component by a previously modeled fire effect. Small fires do not release sufficient heat, while the hot gas layer (HGL) for a large fire is already assumed to fail everything in the room.
2) The room must be able to sustain a buildup of heat. Rooms having partitioning elements with "open" features (e.g., stairways, grates, pipe chases, etc.) do not facilitate room heatup and would instead promote a natural convection of heat upward and away from the equipment, which would generally be located closer to the floor.
3) The function of the impacted component must be different in some way from the functions of all equipment already affected by the fire. For example, since failure of a pump to start is functionally equivalent in the Fire PRA to failure of the associated discharge valve to open, there would be no additional contribution to risk for the fire failing one by an increase in room ambient temperature, if the fire was already modeled as directly failing the other.

Equipment is generally located in the battery rooms and the Emergency Core Cooling System room by common function and divisionally separated.

However, within conservative assumptions consistent with the Fire PRA, EC 95919 considered some limiting components, which were identified in the room heat-up analyses described in BNP-PSA-070, to be potentially affected by a fire producing ambient operating temperatures above the maximum temperature for the limiting component. For these scenarios, there was little or no measureable impact on the risk estimates (i.e., CDF, LERF, ACDF, and ALERF), as follows:

Enclosure 1 Page 12 of 15 Delta CDF and Delta LERF Unit I Unit 2 ALERF ALERF ACDF [lyr] [/yr] ACDF [lyr] [lyr]

VFDRs 1.2E-06 1.3E-08 2.7E-06 5.2E-09 Recovery Actions 11 9.1 E-07 9.1E-08 9.1E-07 9.1 E-08 Total 2.1E-06 1.OE-07 3.6E-06 9.6E-08

[1] Values are for recovery actions associated with control room abandonment due to environmental reasons and address those actions away from the remote shutdown panel.

CDF and LERF Unit I Unit 2 CDF [Iyr] LERF [/yr] CDF [/yr] LERF [/yr]

Internal Events plus External Flooding and 1.4E-05 6.2E-07 1.4E-05 6.2E-07 High Winds Firet1 ] 2.OE-05 4.3E-06 2.OE-05 4.OE-06 Fire - Recovery Actions[21 1.0E-06 1.0E-07 1.0E-06 1.0E-07 Total 3.6E-05 5.OE-06 3.5E-05 4.8E-06 Il2 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.

For each unit, the risk increases were <1 E-7 and <1 E-9, for CDF and LERF, respectively, and below truncation of 1 E-9 and 1 E-1 0, for ACDF and ALERF. The changes in CDF and LERF are the impacts of the sensitivity analysis on the base model. The changes in ACDF and ALERF are the impacts of the sensitivity analysis on the treatment of VFDRs in the compliant case. For Unit 1, the changes in risk were 7.9E-8 and 2.9E-1 0, for CDF and LERF, respectively. For Unit 2, the changes in risk were 3.5E-8 and 0.OOE+00, for CDF and LERF, respectively. Since this conservatively represents less than 0.5% of the total risk, the risk impacts are considered negligible.

The limiting components and the associated maximum ambient operating temperatures were determined as part of the room heat-up analyses that were performed in BNP-PSA-070 to

Enclosure 1 Page 13 of 15 support Heating, Ventilation, and Air Conditioning modeling for the Internal Event PRA. Those limiting components and the associated maximum ambient operating temperatures were:

  • 131 OF for the battery chargers in the Battery Rooms;

The basis for the limiting components and the associated maximum ambient operating temperatures were reevaluated for consideration of fire effects. Except for the battery chargers, these limits were based on time-dependent temperature profiles from qualification testing over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time. For the High Pressure Coolant Injection solenoid valve and the RCIC pressure switches, the above limits were determined to be applicable to the Fire PRA. For the Core Spray pumps, a review of the qualification testing covering the assumed one-hour duration of the fire determined the appropriate maximum ambient operating temperature to be 3000 F. For the battery chargers, the limiting temperature represented conservative use of vendor-supplied information based on speculation about the possible types of battery charger that might be present. Further review of the associated technical manual clarified the particular type of equipment and permitted the use of other vendor-supplied information which indicated 140°F to be a more appropriate but still conservative maximum ambient operating temperature for the Fire PRA.

With the maximum ambient operating temperature as the limiting criteria, the required heat releases were determined by the same Fire Dynamic Tools (FDTs) that were used to model the hot gas layer. For those fire scenarios determined to be capable of releasing sufficient heat, the risk was quantified with additional failures for the temperature-limited equipment.

The analysis included multiple conservatisms. The Fire PRA conservatively assumed the same room temperature profiles that were used in the room heatup analysis for the Internal Events PRA. However, those temperature profiles were selected to represent the worst case accidents and would be more limiting than the temperature profiles associated with a fire-induced transient. The fire was assumed to be out after one hour, but firefighting activities (e.g., opening doors and spraying water) were not credited with limiting the heat release or the room temperature increase during that hour. In calculating the room temperature increase, the "open" features (e.g., stairways, grates, pipe chases, etc.) in the room partitioning elements were neglected. This would tend to postulate a room temperature rise where one would be physically implausible. The "soak time" factor normally provided for cables to reach ambient temperature was removed for this analysis, and equipment failure was assumed when the ambient temperatures reached the limiting temperatures. Temperature profiles established through qualification testing demonstrate equipment qualification for continued operation with no immediate adverse effects rather than an actual point of failure.

PRA RAI 23 Section 2.4.3.3 of National Fire Protection Association Standard 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition, (NFPA 805) incorporated by reference into Title 10, Code of FederalRegulations (10 CFR)

Section 50.48(c) states that the probabilistic safety assessment (PSA) (PSA is also referred to

Enclosure 1 Page 14 of 15 as PRA) approach, methods, and data shall be acceptable to the authority having jurisdiction, which is the NRC. Regulatory Guide (RG) 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," identifies NUREG/CR-6850, Nuclear Energy Institute 04-02, Revision 2, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program under 10 CFR 50.48(c), Revision 2," and the ongoing FAQ process as documenting acceptable methods to the staff for adopting a fire protection program consistent with NFPA-805.

The RAI and sensitivity study that are listed below address PRA methods that have not been accepted by the NRC staff. Although the licensee demonstrated that the individual effect of removing a specific method did not result in exceeding the guidelines in the RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis." The licensee neither showed the impact of the combined effect nor provided adequate justification to demonstrate acceptability of these methods.

Therefore, unless a method is eventually found to be acceptable by the NRC, that method needs to be replaced by a previously acceptable method, or the FPRA of record for the transition needs to be based on the sensitivity studies listed below, which would limit the ability to exercise "self-approval."

  • PRA RAI 1.d-01 regarding use of transient fire HRRs less than 317 kW (unless the alternative response described in follow-up RAI 1.d-01 is provided);
  • Sensitivity analysis reported in Section 4.8.3.2 of the LAR removing credit for Control Power Transformers in circuit failure analysis;
  • PRA RAI 6-01 regarding the sensitivity analysis reported in Section 4.8.3.6 of the LAR removing credit for incipient detection in MCBs;

" PRA RAI 8-01 regarding evaluation of temperature limited components (unless the response to PRA RAI 8-01 shows the risk to be negligible);

  • PRA RAI 1.i-01 regarding SOKC for FPRA specific parameters.
a. Provide the results of an aggregate analysis that provides the integrated impact on the fire risk (i.e., the total transition CDF, LERF, ACDF, ALERF) of replacing the above methods with methods that are acceptable to the NRC. In this aggregate analysis, for those cases where the individual issues have a synergistic impact on the results, a simultaneous analysis must be performed. For those cases where no synergy exists, a one-at-a-time analysis may be done. For those cases that have a negligible impact, a qualitative evaluation may be done.

Response

For the integrated impact on the fire risk, a simultaneous analysis was performed for the removal of the CPTs in circuit failure analysis and the SOKC for Fire PRA specific parameters to capture possible synergisms. The results of this analysis were provided in the response to PRA RAI 1 .i.01. The sensitivity analysis for removing credit for incipient detection in the MCBs was performed as a one-at-a-time analysis because no synergy is expected between the statistical modeling for MCBs, in accordance with Appendix L of NUREG/CR-6850, and other fire modeling approaches. The results of this analysis were provided in the response to PRA RAI 6.01. Likewise, no synergy is expected for the temperature limited components. As described in the response to PRA RAI 8.01, the evaluation of temperature limited components was determined, based on a conservative analysis, to have negligible impact on risk. The resulting aggregate risk and aggregate delta risk for these analyses are as follows:

Enclosure 1 Page 15 of 15 Unit I Unit 2 Aggregate Delta Risk ALERF ALERF ACDF [lyr.] [/yr.] ACDF [Iyr.] [lyr.]

Baseline (Compliant VFDR) Delta Risk (from LAR Section W.2.2 in letter BSEP 13-0107, dated September 30, 2013)

Changes for CPT and SOKC (from response for PRA RAI 1.i.01)

Changes for MCB incipient detection N/A1 11 N/A111 N/Al1] N/A[11 (from response for PRA RAI 6.01)

Changes for ambient temperature <1E_9[21 <1E-10[2] <1E-9121 <1E-10[2]

(from response for PRA RAI 8.01)

Total 2.1E-6 1.1E-7 3.9E-6 1.2E-7 Note: 'Alternate shutdown was not credited as a recovery for control room abandonment due to loss of control.

2 No cutset was generated at truncation of 1E-9 for CDF and 1E-1 0 for LERF.

Unit I Unit 2 Aggregate Risk CDF [/yr.] LERF [/yr.] CDF [/yr.] LERF [/yr.]

Baseline Risk (from LAR Section W.2.2 in letter BSEP 3.5E-5 5.OE-6 3.5E-5 4.8E-6 13-0107, dated September 30, 2013)

Changes for CPT and SOKC (from response for PRA RAI 1.i.01)

Changes for MCB incipient detection 4.9E-6 -4.1E-7 4.7E-6 -3.9E-7 (from response for PRA RAI 6.01)

Changes for ambient temperature 7.9E-8 2.9E-10 3.5E-8 0.0E+00 (from response for PRA RAI 8.01) 8 Total 4.OE-5 4.6E-6 4.OE-5 4.4E-6 The changes in CDF and LERF are the impacts of the sensitivity analyses on the base model.

The changes in ACDF and ALERF are the impacts of the sensitivity analyses on the treatment of VFDRs in the compliant case.

The risk impact of transient fire Heat Release Rates (HRRs) less than 317 kW was not included in an aggregate analysis, because the use of a lower HRR for transient ignition sources in specific areas where supported by evaluation is consistent with the clarified guidance in Section G.5 of NUREG/CR-6850, as endorsed by the NRC in a letter to the Nuclear Energy Institute (NEI) dated June 21, 2012 (i.e., ADAMS Accession Number ML12171A583). For additional information, see the response to PRA RAI 1.d.01 in letter dated March 14, 2014 (i.e.,

ADAMS Accession Number ML14079A233).

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