NLS2012068, 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: | Dori Willis Nebraska Public Power District (NPPD) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NLS2012068, TAC ME8551 | |
| Download: ML12202A042 (9) | |
Text
N Nebraska Public Power District Always there when you need us 50.90 NLS2012068 July 12, 2012 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001
Subject:
Response to Acceptance Review of Cooper Nuclear Station License Amendment Request to Adopt NFPA-805 Cooper Nuclear Station, Docket No. 50-298, DPR-46
Reference:
- 1. E-mail from Lynnea Wilkins, U.S. Nuclear Regulatory Commission, to Edward 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 Amendment Request to Revise the Fire Protection Licensing Basis to NFPA 805 Per 10 CFR 50.48(c)" (NLS2012006)
Dear Sir or Madam:
The purpose of this letter is for the Nebraska Public Power District (NPPD) to respond to the Nuclear 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 License Amendment Request are provided in Attachment 2. NPPD has determined that the No Significant Hazards Consideration determination provided in Reference 2 remains bounding, and that this change therefore does not involve a significant hazard.
There are no commitments made in this submittal. Should you have any questions concerning this matter, please contact Todd Stevens, NFPA 805 Transition Project Manager, at (402) 825-5159.
COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 www.nppd.com
NLS2012068 Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct.
Executed on: :: I Vo\%
(Date)
Sincerely, Demetrius4WiL. is General Manager of Plant Operations DLW/wv : Response to Acceptance Review of Cooper Nuclear Station License Amendment Request to Adopt NFPA-805 : Revisions to the License Amendment Request to Revise the Fire Protection Licensing Basis to NFPA 805 Per 10 CFR 50.48(c) cc: Regional Administrator w/Attachments USNRC - Region IV Cooper Project Manager w/Attachments USNRC - NRR Project Directorate IV-1 Senior Resident Inspector w/Attachments USNRC - CNS Nebraska Health and Human Services w/Attachments Department of Regulation and Licensure NPG Distribution w/o Attachments CNS Records w/Attachments
NLS2012068 Page 1 of 4 Attachment I Response to Acceptance Review of Cooper Nuclear Station License Amendment Request to Adopt NFPA-805 The Nuclear Regulatory Commission (NRC) comments regarding the Acceptance Review of the Cooper Nuclear Station (CNS) License Amendment Request (LAR) to adopt NFPA 805 is shown in italics. The Nebraska Public Power District's (NPPD) response to each comment is shown in block font.
NRC Comment 1 LAR, Attachment G states that a majority of recover, actions (RA) currently creditedhave been assessed under the existing Fire ProtectionProgram,which includedfield validation. However, according to LAR Attachment S, Implementation Item S-3.6, a confirmatory demonstration(field validation walk-through) of the feasibility for the creditedNFPA 805 recovetrv actions will be performed and documented aspart of LAR implementation. It appears that the extent to which the RA feasibility evaluation remains to be completed. Pleaseprovide the results of the completed evaluation and clarify the potential impact this remaining work may have on the results presented in the LAR.
NPPD Response CNS Calculation NEDC 10-041 documents the results of the NFPA 805 RA feasibility assessment that evaluated the RA in Attachment G against the criteria outlined in Frequently Asked 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 high level of confidence. All RA, except those associated with manually opening the 4160VAC feeder breakers from the start-up transformer to Bus C and D, are part of the existing Appendix R program, and are documented in existing CNS procedures. These procedures have been field validated per the requirements of the CNS Procedure Change Process. The new RA to manually open the 4160VAC feeder breakers from the start-up transformer to Bus C and D involve removing control power fuses and tripping the breaker. These new actions are identical to existing actions to manually open other 4160VAC feeder breakers that exist in current CNS procedures. Therefore, it is concluded that there is a high level of confidence the new actions can be performed.
With the implementation of the Nuclear Safety Capability Assessment into the fire response procedures, there will be significant changes to the procedures due to the large reduction of required manual actions. The intent of Implementation Item S-3.6 is to do a confirmatory demonstration of the final procedures to assure all RA for a given fire area are in concert with each other, and to demonstrate coordination between operators performing the actions. The results of this confirmatory demonstration will be used as inputs for Implementing Item S-3.7.
NLS2012068 Attachment I Page 2 of 4 NRC Comment 2 The additionalrisk of recovery actions is reportedin LAR Table W-2 to be negative in several fire areas. This appearsto be due to combining the risk reductionsfrom plant modifications with the additionalrisk of recovery actions in the going forwardplant, but not in the baseline (compliant)plant. To correct the calculation of the additionalrisk of recovery actions,plant modifications must be creditedin the goingforwardplant as well as the baseline; in other words, the only difference in the goingforwardplant PRA and the baseline PRA is the creditfor recovery actions. Please provide correctedresultsfor the additionalcredit for recovery actions identified in Table W-2 on afire area basis.
NPPD Response Table W-2 has been revised, as provided in Attachment 2, to detail the remaining risk of RA on a fire area basis for the going forward plant probabilistic risk assessment (PRA). These risk values are positive numbers, and were calculated in conformance with FAQ 07-0030. There is no impact on results or conclusions due to the changes made to Table W-2.
NRC Comment 3 LAR section V2 identifies deviationsfrom the guidance in NUREG/CR-6850. For "Transient Fire Frequency," influencefactors less than one were used. Also, a factor of 0. 1 was used to modify both fire frequencyfrom transients (bins 7, 25, and 3 7) and transientfires caused by welding and cutting (bins 6, 24, and 36) for certainfire zones. Please provide a simultaneous or composite sensitivity analysis of the impact on CDF,LERF, AJCDF, andALERF using NUREG/CR-6850 methods. It is expected that, concurrently, the creditfor the factor of 0. 1 which modifies the firefrequency would be removed, i.e. 1 would replace the 0. 1, and that influencefactors less than 1 would be replaced by those integers documented in Table 6-3 of NUREG/CR-6850.
NPPD Response The 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 based on proposed modifications/enhancements to the combustible and hot work controls for Fire Zone 8A (Auxiliary Relay Room) and Fire Zone 9A (Cable Spreading Room). The fire frequency was initially 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 enhanced administrative controls. NPPD identified that fractional values less than I were appropriate for the enhanced administrative controls instead of a value of zero (0) for the influencing factors to account for any unlikely violations of the new enhanced controls. The fractional values were utilized for only two of the influencing factors, therefore, the total transient influencing factor values for these two fire zones were greater than I (i.e., 1.15 and 3.15), which does not represent orders of magnitude differences in the ranges of influencing factors. The ignition frequency calculation weighting factors for other areas of the plant were analyzed based on the existing
NLS2012068 Attachment I Page 3 of 4 occupancy, storage, and maintenance factors, and as such, it was not appropriate to increase the influencing factors for the other areas to account for the proposed enhanced controls in these two fire zones.
The sensitivity analysis was performed for the fire frequency results for transient fire scenarios in Control/Auxiliary/Reactor Building Plant Locations based on using only NUREG/CR-6850 methods by replacing the "Very Low" influencing factor values with maintenance and storage weighting factors identified in Table 6-3 of NUREG/CR-6850 for Fire Zone 9A (Cable Spreading Room) and Fire Zone 8A (Auxiliary Relay Room). Additionally, for three specific transient locations within fire zones (Fire Zone 3C and 3D, the area located above the TIP Room on the 903'-6" Elevation of the Reactor Building, and Fire Zone 2C, the floor areas immediately located around Instrument Racks 25-5 and 25-6 on the 931 '-6" Elevation of the Reactor Building), 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 other fire zones in the Generic Plant Location - Control/Auxiliary/Reactor Building, fire zone transient frequencies also change by decreasing a very small amount. This very small decrease in other transient scenario frequencies was not included in sensitivity analysis, thus providing a worse case sensitivity result. There is no change to conditional core damage probability (CCDP) or conditional 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 risk increases are 6.1 OE-08/year for CDF (difference between base delta CDF and sensitivity delta CDF) and 1.1 E-08/year for LERF (difference between base delta LERF and sensitivity delta LERF).
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 the following reason. Several scenarios in fire zones 8A and 9A result in the use of alternate shutdown. The Fire PRA model assumes that CLERP given core damage is 1.0. This conservatism more than compensates for the noted changes above.
NRC Comment 4 LAR Attachment U uses the term "vendor and utility" when describing the makeup of the team that performed the R. G. 1.200 Rev 1 peer review of the internalevents PRA. This is confusing terminology potentially implying that the review was not an independentpeer review as defined
NLS2012068 Attachment I Page 4 of 4 in the ASME/ANS PRA Standard. Provide clarificationthat the peer review of the internal events PRA discussedin Attachment U met the requirementsof Sections 1-6 and 2-3 of the ASME/ANS PRA Standardfor a peer review. If not provide the F&Os and resolutionsfrom the most recentfidl-scope peer review.
NPPD Response The 2008 CNS internal events PRA peer review referenced in Attachment U of the LAR meets the 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 internal events, at-power PRA. The peer review utilized and was conducted in accordance with the following standards and references:
" NEI 05-04, "Process for Performing Follow-on PRA Peer Reviews Using the ASME PRA Standard", Nuclear Energy Institute, Rev. 1, November 2007.
- ASME RA-Sc-2007, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications", August 2007.
" NRC Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities", Rev. 1, January 2007.
" NRC memorandum, "Notice of Clarification to Rev. I of Regulatory Guide 1.200", FRN July 27, 2007, NRC ADAMS Accession number: ML071170054.
The peer review team was comprised of seven reviewers, of which, two were contractors and five were BWROG utility personnel. Reviewer qualifications and independence were confirmed and documented as part of the 2008 CNS peer review. The peer review team meets the current peer review team composition and personnel qualifications requirements of Sections 1-6 and 2-3 of the combined PRA standard (ASME/ANS RA-Sa-2009).
NLS2012068 Page 1 of 3 Attachment 2 Revisions to the License Amendment Request to Revise the Fire 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 in , 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 Summary Fire NFA NFPA Fire Area Fire Area VFDR(s)
VDs) RA(s)
Rs) Fire Risk FrRik Fire Risk ieRsk Additional Risk of RAs AreaFieAeDecito80 Area Description 805 CF LEFCDan (Fire (Yes/ (Yes/ Eval Delta Eval Delta RiDk An Basis CDF LERF No) No) CDF LERF LERFa RHR SW Booster Pump and Service Air Compressor Areas 1.48E-07 CB-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-08 RPS Room 1A Seal Water Pump Area and Corridor 4.2.4.2 3.43E-06 1.14E-07 Yes Yes 9.49E-08 4.68E-08 9.49E-08 CB-A-1 DC Room1AI1A SwgrRoom Battery 4.68E-08 CB-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-07 DC Swgr Room 1B 6-.947.7 1E-08 CB-C RPS Room lB 4.2.4.2 1.74E-07 7.83E-09 Yes YesNo E E NA Computer Room Control Room and SAS Corridor 0-54.74E-07 CB-D Aux Relay Room 4.2.4.2 1.37E-05 4.20E-06 Yes Yes -1.53E-05 -1.44E-05 Cable Spreading Room 058.91E-08 Cable Expansion Room DG-A Div. 1 Diesel Generator 4.2.3.2 5.09E-06 9.18E-08 No No NA NA NA DG-B Div. 2 Diesel Generator 4.2.3.2 1.16E-06 9.OOE-08 No No NA NA NA
NLS2012068 Page 2 of 3 Table W-2 CNS Fire Area Risk Summary Additional Fire NFPA VFDR(s) RA(s) Fire Risk Fire Risk Risk of RAs re Fire Area Description 805 Area Fire Area (Yes/ (Yes/ Eval Delta Eval Delta Risk As Basis CDF LERF No) No) CDF LERF LERFa SW Pump Area 2.36E-08 IS-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 443E4ONA Area 7.90E-08 4.2.4.2 2.52E-07 1.04E-08 Yes Yes 7.90E-08 2.29E-09 7.9E-08 RB-A RCIC and CS A Pump Room 2.29E-09 4.2.4.2 1.25E-07 1.15E-08 Yes Yes 1.25E-07 1.15E-08 1.25E-07 RB-B Core Spray B Pump Room Hydraulic Drive Pump Area 1.15E-08 RHR Pump Rm 1A and IC 4.2.4.2 1.67E-06 7.15E-08 Yes Yes 7.90E-07 5.12E-08 7-908.02E-07 RB-CF CRD Units-North 903' 6" South Corridor 5-t-25.22E-08 RHR HX-1A RHR Pump Room lB and ID HPCI Pump Room 071.20E-06 RB-DI 903' 6" South Corridor 4.2.4.2 2.86E-06 5.05E-07 Yes Yes 9.50E-07 2.OOE-08 CRD Units South 091.05E-07 RHR HX-1B 2.18E-09 2.18E-09 4.2.4.2 2.15E-07 5.14E-09 Yes Yes 2.18E-09 2.44E-09 RB-E Suppression Pool Area 2.44E-09 4.241.60E-07 RB-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
-09 4.2.4.2 1.25E-06 2.58E-07 Yes Yes 1.43E-07 5.65E-08 1.43E-07 RB-J SWGR Room IF 5.65E-08 066.09E-07 4.2.4.2 1.64E-06 2.64E-07 Yes Yes -2.08E-06 -2.15E-06 RB-K SWGR Room IG 061.09E-07
NLS2012068 Page 3 of 3 Table W-2 CNS Fire Area Risk Summary Fire NFA Fire Area Fire Area VD SVFDR(s)s) R RA(s)s) FrRik Fire Risk Fire Risk ieRsk Additional Risk of RAs Fire Area Description Nrea 805 (Yes/ (Yes/ Eval Delta Eval Delta Risk As Basis CDF LERF No) No) CDF LERF (CDF and LERF)
-2.76E-09 0-7.4 1E-08 RB-M RWCU Recirc Pumps and Corridor 4.2.4.2 4.46E-07 2.2 1E-08 Yes Yes -6.32E-07 091.60E-08 RHR HX-IB RB-N Regenerative HX Areas 4.2.4.2 2.63E-07 2.48E-08 Yes YesNo 6.43E-08 5.70E-10 RWCU Recirc Pumps and Corridor RB Elevator and accessway Area RB HVAC Areas 4.63E-08 RB-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-08 Raw Water Cleanup Areas Reactor MG Set Oil Pump Area SBLC Pump Tank and Accessway RB-T (Zone 5A) and Refueling Floor (Zone 4.2.3.2 3.80E-08 6.53E-09 No No NA NA NA 6) 1.88E-08 4.2.4.2 3.15E-08 7.19E-09 Yes Yes 1.88E-08 6.3 IE-09 1.88E-08 RB-V Rx MG Set Area 6.3 1E-09 6.77E-06 4.2.4.2 9.56E-06 3.87E-06 Yes Yes 6.77E-06 3.33E-06 6.77E-06 TB-A Turbine Building 3.33E-06 TB-C Steam Tunnel 4.2.4.2 1.09E-08 1.23E-09 Yes YesNo E NA Yard Yard Outside of Buildings 4.2.3.2 1.35E-06 6.30E-07 No No NA NA NA 8.71E 061.12E-05 5.07E-05 1.05E-05 -8.71E-06 -1.29E-05 Total 0-53.97E-06
Reference:
Response to Comment 2 from Attachment 1. Additionally, fire areas CB-C, IS-A, RB-N, and TB-C contain corrections. 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.