ML12202A042

Cooper - Response to Acceptance Review of Cooper Nuclear Station License Amendment Request to Adopt NFPA-805

ML12202A042
Person / Time
Site:	Cooper
Issue date:	07/12/2012
From:	Willis D L Nebraska Public Power District (NPPD)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NLS2012068, TAC ME8551
Download: ML12202A042 (9)

Text

NNebraska Public Power DistrictAlways there when you need us50.90NLS2012068July 12, 2012U.S. Nuclear Regulatory CommissionAttn: Document Control DeskWashington, D.C. 20555-0001Subject: Response to Acceptance Review of Cooper Nuclear Station License AmendmentRequest to Adopt NFPA-805Cooper Nuclear Station, Docket No. 50-298, DPR-46Reference: 1. E-mail from Lynnea Wilkins, U.S. Nuclear Regulatory Commission, toEdward L. McCutchen, Nebraska Public Power District, dated June 21, 2012,"Acceptance Review of Cooper Nuclear Station LAR to Adopt NFPA-805(ME8551)"2. Letter from Brian J. O'Grady, Nebraska Public Power District, to U.S.Nuclear Regulatory Commission, dated April 24, 2012, "License AmendmentRequest to Revise the Fire Protection Licensing Basis to NFPA 805 Per 10CFR 50.48(c)" (NLS2012006)Dear Sir or Madam:The purpose of this letter is for the Nebraska Public Power District (NPPD) to respond to theNuclear Regulatory Commission e-mail that provided LIC- 109 acceptance review observations(Reference 1) on the NPPD NFPA 805 Transition License Amendment Request (Reference 2).The responses are provided in Attachment 1. Corresponding changes to the NFPA 805 LicenseAmendment Request are provided in Attachment 2. NPPD has determined that the NoSignificant Hazards Consideration determination provided in Reference 2 remains bounding, andthat this change therefore does not involve a significant hazard.There are no commitments made in this submittal. Should you have any questions concerningthis matter, please contact Todd Stevens, NFPA 805 Transition Project Manager, at (402) 825-5159.COOPER NUCLEAR STATIONP.O. Box 98 / Brownville, NE 68321-0098Telephone: (402) 825-3811 / Fax: (402) 825-5211www.nppd.com NLS2012068Page 2 of 2I declare under penalty of perjury that the foregoing is true and correct.Executed on: :: I Vo\%(Date)Sincerely,Demetrius L. 4Wi isGeneral Manager of Plant OperationsDLW/wvAttachment 1: Response to Acceptance Review of Cooper Nuclear Station License AmendmentRequest to Adopt NFPA-805Attachment 2: Revisions to the License Amendment Request to Revise the Fire ProtectionLicensing Basis to NFPA 805 Per 10 CFR 50.48(c)cc: Regional Administrator w/AttachmentsUSNRC -Region IVCooper Project Manager w/AttachmentsUSNRC -NRR Project Directorate IV-1Senior Resident Inspector w/AttachmentsUSNRC -CNSNebraska Health and Human Services w/AttachmentsDepartment of Regulation and LicensureNPG Distribution w/o AttachmentsCNS Records w/Attachments NLS2012068Attachment 1Page 1 of 4Attachment IResponse to Acceptance Review of Cooper Nuclear StationLicense Amendment Request to Adopt NFPA-805The Nuclear Regulatory Commission (NRC) comments regarding the Acceptance Review of theCooper Nuclear Station (CNS) License Amendment Request (LAR) to adopt NFPA 805 isshown in italics. The Nebraska Public Power District's (NPPD) response to each comment isshown in block font.NRC Comment 1LAR, Attachment G states that a majority of recover, actions (RA) currently credited have beenassessed under the existing Fire Protection Program, which included field validation. However,according to LAR Attachment S, Implementation Item S-3.6, a confirmatory demonstration (fieldvalidation walk-through) of the feasibility for the credited NFPA 805 recovetrv actions will beperformed and documented as part of LAR implementation. It appears that the extent to whichthe RA feasibility evaluation remains to be completed. Please provide the results of thecompleted evaluation and clarify the potential impact this remaining work may have on theresults presented in the LAR.NPPD ResponseCNS Calculation NEDC 10-041 documents the results of the NFPA 805 RA feasibilityassessment that evaluated the RA in Attachment G against the criteria outlined in FrequentlyAsked Question (FAQ) 07-0030, "Establishing Recovery Actions." Based on this assessment,the credited RA have been reviewed, and subsequently determined to be feasible with a highlevel of confidence. All RA, except those associated with manually opening the 4160VACfeeder breakers from the start-up transformer to Bus C and D, are part of the existing AppendixR program, and are documented in existing CNS procedures. These procedures have been fieldvalidated per the requirements of the CNS Procedure Change Process. The new RA to manuallyopen the 4160VAC feeder breakers from the start-up transformer to Bus C and D involveremoving control power fuses and tripping the breaker. These new actions are identical toexisting actions to manually open other 4160VAC feeder breakers that exist in current CNSprocedures. Therefore, it is concluded that there is a high level of confidence the new actionscan be performed.With the implementation of the Nuclear Safety Capability Assessment into the fire responseprocedures, there will be significant changes to the procedures due to the large reduction ofrequired manual actions. The intent of Implementation Item S-3.6 is to do a confirmatorydemonstration of the final procedures to assure all RA for a given fire area are in concert witheach other, and to demonstrate coordination between operators performing the actions. Theresults of this confirmatory demonstration will be used as inputs for Implementing Item S-3.7.

NLS2012068Attachment IPage 2 of 4NRC Comment 2The additional risk of recovery actions is reported in LAR Table W-2 to be negative in severalfire areas. This appears to be due to combining the risk reductions from plant modificationswith the additional risk of recovery actions in the going forward plant, but not in the baseline(compliant) plant. To correct the calculation of the additional risk of recovery actions, plantmodifications must be credited in the going forward plant as well as the baseline; in otherwords, the only difference in the going forward plant PRA and the baseline PRA is the credit forrecovery actions. Please provide corrected results for the additional credit for recovery actionsidentified in Table W-2 on afire area basis.NPPD ResponseTable W-2 has been revised, as provided in Attachment 2, to detail the remaining risk of RA on afire area basis for the going forward plant probabilistic risk assessment (PRA). These risk valuesare positive numbers, and were calculated in conformance with FAQ 07-0030. There is noimpact on results or conclusions due to the changes made to Table W-2.NRC Comment 3LAR section V2 identifies deviations from the guidance in NUREG/CR-6850. For "TransientFire Frequency," influence factors less than one were used. Also, a factor of 0. 1 was used tomodify both fire frequencyfrom transients (bins 7, 25, and 3 7) and transient fires caused bywelding and cutting (bins 6, 24, and 36) for certain fire zones. Please provide a simultaneous orcomposite sensitivity analysis of the impact on CDF, LERF, AJCDF, and ALERF usingNUREG/CR-6850 methods. It is expected that, concurrently, the credit for the factor of 0. 1which modifies the fire frequency would be removed, i.e. 1 would replace the 0. 1, and thatinfluence factors less than 1 would be replaced by those integers documented in Table 6-3 ofNUREG/CR-6850.NPPD ResponseThe transient fire frequency influencing factors utilized to generate the core damage frequency(CDF), large early release frequency (LERF), ACDF, and ALERF in the CNS LAR were basedon proposed modifications/enhancements to the combustible and hot work controls for Fire Zone8A (Auxiliary Relay Room) and Fire Zone 9A (Cable Spreading Room). The fire frequency wasinitially calculated using transient influencing factors identified in Table 6-3 of NUREG/CR-6850; however, it was determined that these two fire zones would receive enhancedadministrative controls. NPPD identified that fractional values less than I were appropriate forthe enhanced administrative controls instead of a value of zero (0) for the influencing factors toaccount for any unlikely violations of the new enhanced controls. The fractional values wereutilized for only two of the influencing factors, therefore, the total transient influencing factorvalues for these two fire zones were greater than I (i.e., 1.15 and 3.15), which does not representorders of magnitude differences in the ranges of influencing factors. The ignition frequencycalculation weighting factors for other areas of the plant were analyzed based on the existing NLS2012068Attachment IPage 3 of 4occupancy, storage, and maintenance factors, and as such, it was not appropriate to increase theinfluencing factors for the other areas to account for the proposed enhanced controls in these twofire zones.The sensitivity analysis was performed for the fire frequency results for transient fire scenarios inControl/Auxiliary/Reactor Building Plant Locations based on using only NUREG/CR-6850methods by replacing the "Very Low" influencing factor values with maintenance and storageweighting factors identified in Table 6-3 of NUREG/CR-6850 for Fire Zone 9A (CableSpreading Room) and Fire Zone 8A (Auxiliary Relay Room). Additionally, for three specifictransient locations within fire zones (Fire Zone 3C and 3D, the area located above the TIP Roomon the 903'-6" Elevation of the Reactor Building, and Fire Zone 2C, the floor areas immediatelylocated around Instrument Racks 25-5 and 25-6 on the 931 '-6" Elevation of the ReactorBuilding), the 0.1 factor was eliminated (changed to 1.0).Since any adjustment in the transient influencing factors for a specific fire zone impacts all otherfire zones in the Generic Plant Location -Control/Auxiliary/Reactor Building, fire zone transientfrequencies also change by decreasing a very small amount. This very small decrease in othertransient scenario frequencies was not included in sensitivity analysis, thus providing a worsecase sensitivity result. There is no change to conditional core damage probability (CCDP) orconditional large early release frequency (CLERP).The risk increases by 2.54E-06/year for CDF and 1.51 E-06/year for LERF. The delta riskincreases are 6.1 OE-08/year for CDF (difference between base delta CDF and sensitivity deltaCDF) and 1.1 E-08/year for LERF (difference between base delta LERF and sensitivity deltaLERF).The results are summarized as follows:1. CDF increased by about 5%2. LERF increased by about 12%3. Delta CDF increased by less than 1%4. Delta LERF increased by less than 1%Conclusions did not change with respect to Regulatory Guide 1.174 acceptance guidelines.Further, there is considerable conservatism in the large early release frequency change for thefollowing reason. Several scenarios in fire zones 8A and 9A result in the use of alternateshutdown. The Fire PRA model assumes that CLERP given core damage is 1.0. Thisconservatism more than compensates for the noted changes above.NRC Comment 4LAR Attachment U uses the term "vendor and utility" when describing the makeup of the teamthat performed the R. G. 1.200 Rev 1 peer review of the internal events PRA. This is confusingterminology potentially implying that the review was not an independent peer review as defined NLS2012068Attachment IPage 4 of 4in the ASME/ANS PRA Standard. Provide clarification that the peer review of the internalevents PRA discussed in Attachment U met the requirements of Sections 1-6 and 2-3 of theASME/ANS PRA Standard for a peer review. If not provide the F&Os and resolutions from themost recent fidl-scope peer review.NPPD ResponseThe 2008 CNS internal events PRA peer review referenced in Attachment U of the LAR meetsthe requirements for an independent peer review as defined in the ASME/ANS PRA Standard.The 2008 CNS PRA peer review is a full-scope review of the Technical Elements of the internalevents, at-power PRA. The peer review utilized and was conducted in accordance with thefollowing standards and references:" NEI 05-04, "Process for Performing Follow-on PRA Peer Reviews Using the ASME PRAStandard", Nuclear Energy Institute, Rev. 1, November 2007.* ASME RA-Sc-2007, "Standard for Probabilistic Risk Assessment for Nuclear Power PlantApplications", August 2007." NRC Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy ofProbabilistic Risk Assessment Results for Risk-Informed Activities", Rev. 1, January 2007." NRC memorandum, "Notice of Clarification to Rev. I of Regulatory Guide 1.200", FRNJuly 27, 2007, NRC ADAMS Accession number: ML071170054.The peer review team was comprised of seven reviewers, of which, two were contractors andfive were BWROG utility personnel. Reviewer qualifications and independence were confirmedand documented as part of the 2008 CNS peer review. The peer review team meets the currentpeer review team composition and personnel qualifications requirements of Sections 1-6 and 2-3of the combined PRA standard (ASME/ANS RA-Sa-2009).

NLS2012068Attachment 2Page 1 of 3Attachment 2Revisions to the License Amendment Request to Revise theFire Protection Licensing Basis to NFPA 805 Per 10 CFR 50.48(c)This attachment provides changes to the NFPA 805 License Amendment Request based on the responses to the Comments provided inAttachment 1, as well as for other clarifications. The changes are presented in underline/strikeout format.1. Table W-2, CNS Fire Area Risk Summary, is revised as follows.Table W-2 CNS Fire Area Risk SummaryNFPA VFDR(s) RA(s) Fire Risk Fire Risk AdditionalFire NFA Fire Area Fire Area VDs) Rs) FrRik ieRsk Risk of RAs(Fire Area Description 805 (Yes/ (Yes/ Eval Delta Eval Delta RiDk AnAreaFieAeDecito80 CF LEFCDanBasis CDF LERF No) No) CDF LERF LERFaRHR SW Booster Pump and ServiceAir Compressor Areas 1.48E-07CB-A Emerg Condensate Storage Tank Area 4.2.4.2 6.89E-07 7.03E-08 Yes Yes 1.48E-07 4.62E-08 4.62E-08RPS Room 1ASeal Water Pump Area and CorridorCB-A-1 Battery Room 1A 4.2.4.2 3.43E-06 1.14E-07 Yes Yes 9.49E-08 4.68E-08 9.49E-08DC Swgr Room I1A 4.68E-08CB-B Battery Room lB 4.2.4.2 4.61E-06 1.85E-07 Yes Yes 1.54E-07 6.94E-08 4--14.34E-07DC Swgr Room 1B 6-.947.7 1 E-08CB-C RPS Room lB 4.2.4.2 1.74E-07 7.83E-09 Yes YesNo E E NAComputer RoomControl Room and SAS Corridor 0-54.74E-07CB-D Aux Relay Room 4.2.4.2 1.37E-05 4.20E-06 Yes Yes -1.53E-05 -1.44E-05Cable Spreading Room 058.91E-08Cable Expansion RoomDG-A Div. 1 Diesel Generator 4.2.3.2 5.09E-06 9.18E-08 No No NA NA NADG-B Div. 2 Diesel Generator 4.2.3.2 1.16E-06 9.OOE-08 No No NA NA NA NLS2012068Attachment 2Page 2 of 3Table W-2 CNS Fire Area Risk SummaryAdditionalFire NFPA VFDR(s) RA(s) Fire Risk Fire Risk Risk of RAsre Fire Area Description 805 Area Fire Area (Yes/ (Yes/ Eval Delta Eval Delta Risk AsBasis CDF LERF No) No) CDF LERF LERFaSW Pump Area 2.36E-08IS-A Circ Water Pump and Traveling Screen 4.2.4.2 2.14E-07 2.61E-08 Yes Ye-sNo 2.36E-08 4.43E-10 443E4ONAArea7.90E-08RB-A RCIC and CS A Pump Room 4.2.4.2 2.52E-07 1.04E-08 Yes Yes 7.90E-08 2.29E-09 7.9E-082.29E-09RB-B Core Spray B Pump Room 4.2.4.2 1.25E-07 1.15E-08 Yes Yes 1.25E-07 1.15E-08 1.25E-07Hydraulic Drive Pump Area 1.15E-08RHR Pump Rm 1A and ICRB-CF CRD Units-North 4.2.4.2 1.67E-06 7.15E-08 Yes Yes 7.90E-07 5.12E-08 7-908.02E-07903' 6" South Corridor 5-t-25.22E-08RHR HX-1ARHR Pump Room lB and IDHPCI Pump Room 071.20E-06RB-DI 903' 6" South Corridor 4.2.4.2 2.86E-06 5.05E-07 Yes Yes 9.50E-07 2.OOE-08CRD Units South 091.05E-07RHR HX-1B2.18E-09RB-E Suppression Pool Area 4.2.4.2 2.15E-07 5.14E-09 Yes Yes 2.18E-09 2.44E-09 2.18E-092.44E-094.241.60E-07RB-FN Rx Bldg 903' 6" NE Comer 4.2.4.2 1.84E-06 1.83E-08 Yes Yes -1.24E-07 -7.03E-10 -703-E4-05.12E-09RB-J SWGR Room IF 4.2.4.2 1.25E-06 2.58E-07 Yes Yes 1.43E-07 5.65E-08 1.43E-075.65E-08066.09E-07RB-K SWGR Room IG 4.2.4.2 1.64E-06 2.64E-07 Yes Yes -2.08E-06 -2.15E-06061.09E-07 NLS2012068Attachment 2Page 3 of 3Table W-2 CNS Fire Area Risk SummarySVFDR(s) RA(s) Fire Risk Fire Risk AdditionalFire NFA Fire Area Fire Area VD s) R s) FrRik ieRsk Risk of RAsNrea Fire Area Description 805 (Yes/ (Yes/ Eval Delta Eval Delta Risk AsBasis CDF LERF No) No) CDF LERF (CDF andLERF)0-7.4 1E-08RB-M RWCU Recirc Pumps and Corridor 4.2.4.2 4.46E-07 2.2 1E-08 Yes Yes -6.32E-07 -2.76E-09091.60E-08RHR HX-IBRB-N Regenerative HX Areas 4.2.4.2 2.63E-07 2.48E-08 Yes YesNo 6.43E-08 5.70E-10RWCU Recirc Pumps and CorridorRB Elevator and accessway AreaRB HVAC Areas 4.63E-08RB-P Fuel Pool, HX, CRD Repair Room, and 4.2.4.2 9.02E-08 1.97E-08 Yes Yes 4.63E-08 1.72E-08 1.72E-08Raw Water Cleanup AreasReactor MG Set Oil Pump AreaSBLC Pump Tank and AccesswayRB-T (Zone 5A) and Refueling Floor (Zone 4.2.3.2 3.80E-08 6.53E-09 No No NA NA NA6)1.88E-08RB-V Rx MG Set Area 4.2.4.2 3.15E-08 7.19E-09 Yes Yes 1.88E-08 6.3 IE-09 1.88E-086.3 1E-096.77E-06TB-A Turbine Building 4.2.4.2 9.56E-06 3.87E-06 Yes Yes 6.77E-06 3.33E-06 6.77E-063.33E-06TB-C Steam Tunnel 4.2.4.2 1.09E-08 1.23E-09 Yes YesNo E NAYard Yard Outside of Buildings 4.2.3.2 1.35E-06 6.30E-07 No No NA NA NA8.71E061.12E-05Total 5.07E-05 1.05E-05 -8.71E-06 -1.29E-050-53.97E-06Reference:Response to Comment 2 from Attachment 1. Additionally, fire areas CB-C, IS-A, RB-N, and TB-C containcorrections. No Recovery Actions (RA) are credited with the VFDR for these fire areas; thus, column "RA(s)(Yes/No)" is revised to "No," and the associated additional risk of RA for these fire areas is not applicable.