ML14147A368

Joseph M. Farley, Response to Request for Additional Information Regarding License Amendment Request for Transition to 10 CFR 50.48(c) - NFPA 805 Performance Based Standard for Fire Protection for Light Water Reactor Generating Plants

ML14147A368
Person / Time
Site:	Farley
Issue date:	05/23/2014
From:	Pierce C R Southern Co, Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML14147A366	List: NL-14-0733, Response to Request for Additional Information Regarding License Amendment Request for Transition to 10 CFR 50.48(c) - NFPA 805 Performance Based Standard for Fire Protection for Light Water Reactor Generating Plants
References
NL-14-0733, TAC ME9741, TAC ME9742
Download: ML14147A368 (23)

Text

WITHHOLD FROM PUBLICCharles R. Pierce Southern Nuclear DISCLOSURE UNDERRegulatory Affairs Director Operating Company, Inc. 10 CFR 2.390.40 Inverness Center ParkwayPost Office Box 1295Birmingham, Alabama 35201Tel 205.992.7872Fax 205.992.7601SOUTHERNACOMPANYMay 23, 2014Docket Nos.: 50-348 NL-14-073350-364U. S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, D. C. 20555-0001Joseph M. Farley Nuclear PlantResponse to Request for Additional Information Regarding License AmendmentRequest for Transition to 10 CFR 50.48(c) -NFPA 805 Performance BasedStandard for Fire Protection for Light Water Reactor Generating PlantsLadies and Gentlemen:By letter dated September 25, 2012, the Southern Nuclear Operating Company(SNC) submitted a license amendment request (LAR) for Joseph M. Farley Units1 and 2 (Ref. TAC NOS. ME9741 and ME9742). The proposed amendmentrequests the review and approval for adoption of a new fire protection licensingbasis which complies with the requirements in Sections 50.48(a) and 50.48(c) toTitle 10 to the Code of Federal Regulations (10 CFR), and the guidance inRegulatory Guide (RG) 1.205, Revision 1, Risk-Informed, Performance-BasedFire Protection for Existing Light-Water Nuclear Power Plants.By letter dated December 12, 2012, the Nuclear Regulatory Commission (NRC)Staff requested supplemental information regarding the acceptance of the licenseamendment (Adams Accession No. ML12345A398). SNC provided the requestedinformation by letter dated December 20, 2012. The NRC staff subsequentlycompleted the acceptance review by letter dated January 24, 2013, (AdamsAccession No. ML13022A158).By letter dated July 8, 2013, the NRC Staff formally transmitted a request foradditional information (RAI) related to the referenced license amendment. SNC'sresponses to these RAIs are being provided by three submittals. By letter datedSeptember 16, 2013, SNC provided the first set of responses. By letter datedOctober 30, 2013, SNC provided the second set of responses and by letter datedNovember 12, 2013, SNC provided the remaining set of responses. SNCprovided that supplemental responses would be provided for nine of the RAIs.

U.S. Nuclear Regulatory CommissionNL-14-0733Page 2By letter dated March 28, 2014, the NRC Staff formally transmitted the secondround of requests for additional information related to the referenced licenseamendment request. By letter dated April 23, 2014, SNC provided the 30 dayresponse to the second round of RAIs. Enclosure 1 to this letter provides the 60day response to six of the eight remaining eight RAIs. PRA RAIs 06.a.01 andPRA 35 will be provided when the composite effect of the quantification of thePRA model is completed as agreed during the conference call with ShawnWilliams on May 20, 2014. The supplemental responses identified in enclosure 1of SNC's letter dated April 23, 2014 will be included in the response to PRA RAI35. Enclosure 2 provides a supplemental response to the Safe ShutdownAnalysis RAI 10.01.Attachment G, Recovery Actions Transition provides a revision to page 9 whichadds component numbers. Attachment G contains sensitive information andshould be withheld from public disclosure under 10 CFR 2.390.The No Significant Hazards Consideration determination provided in the originalsubmittal is not altered by the RAI responses provided herein.If you have any questions, please contact Ken McElroy at (205) 992-7369.Mr. C. R. Pierce states he is Regulatory Affairs Director of Southern NuclearOperating Company, is authorized to execute this oath on behalf of SouthernNuclear Operating Company and, to the best of his knowledge and belief, thefacts set forth in this letter are true and correct.Respectfully submitted,ý X. /f.C. R. PierceRegulatory Affairs DirectorCRP/jkb/lacStoandsubscribedbeforeme this Z day of ,2014.Notary Public fMy commission expires:/- 2 0 / 9' U.S. Nuclear Regulatory CommissionNL-14-0733Page 2

Enclosures:

1. Response to Probabilistic Risk Assessment RAIs2. Supplemental Response to Safe Shutdown Analysis RAIAttachments: 1. Revision to Recovery Actions Transition- Attachment Gcc: Southern Nuclear Operating CompanyMr. S. E. Kuczynski, Chairman, President & CEOMr. D. G. Bost, Executive Vice President & Chief Nuclear OfficerMs. C. A. Gayheart, Vice President -FarleyMr. B. L. Ivey, Vice President -Regulatory AffairsMr. D. R. Madison, Vice President -Fleet OperationsMr. B. J. Adams, Vice President -EngineeringRTYPE: CFA04.054U. S. Nuclear Regulatory CommissionMr. V. M. McCree, Regional AdministratorMr. S. A. Williams, NRR Project Manager -FarleyMr. P. K. Niebaum, Senior Resident Inspector -FarleyMr. J. R. Sowa, Resident Inspector -FarleyAlabama Department of Public HealthDr. D. E. Williamson, State Health Officer Joseph M. Farley Nuclear PlantResponse to Request for Additional InformationRegarding License Amendment Request for Transition to 10 CFR 50.48(c)NFPA 805 Performance Based Standard for Fire Protection for Light WaterReactor Generating PlantsEnclosure 1Response to Probabilistic Risk Assessment RAIs Response to Probabilistic Risk Assessment RAIsFarley PRA RAI 01.01LAR Attachment V, Table V.2-2, provides the results of the electrical cabinet fireseverity sensitivity analysis for Unit 1, also indicating similar results for Unit 2.There, the base CDF rose from 5.24E-5/y to 7.05E-5/y, an increase of 1.81 E-5/y.For A CDF, the base value rose from 8.80E-6/y to 1.03E-5/y, an increase of1.50E-6/y. The analogous results for LERF and A LERF were as follows: (1) aLERF increase of 2.59E-6/y from 1.26E-6/y to 3.85E-6/y; (2) a A LERF increaseof 9.90E-8/y from 4.14E-7/y to 5.13E-7/y. Subsequently, the LAR wassupplemented by a sensitivity analysis which included the effect of removingcredit for very early warning fire detection system (VEWFDS) in the main controlroom (MCR) in addition to the electrical cabinet fire severity adjustment. Theresults were as follows: (1) CDF now rose only 1.41 E-5/y (vs. the previous 1.81 E-5/y); (2) A CDF now rose only 1.18E-6/y (vs. the previous 1.50E-6/y); (3) LERFnow rose more by 6.28E-6/y (vs. the previous 2.59E-6/y); (4) A LERF now rosemore by 2.88E-7/y (vs. the previous 9.90E-8/y).In a letter dated September 16, 2013 (ADAMS Accession No. ML14038A019), asjustification for the smaller increase for CDF and A CDF with credit for bothVEWFDS and electrical cabinet severity adjustment removed, the licenseeindicated via Table 1 that, in addition to removing credit for VEWFDS in the MCR,the following additional refinements were now included: (1) refined main controlboard (MCB) fire scenarios (via App. L of NUREG/CR-6850); (2) more realisticprobabilities for HGLs; (3) refined circuit analysis for selected fire scenarios; (4)correction to anomalies in fire ignition frequencies for selected fire scenarios. Asa result, the CDF and A CDF increase for removing both VEWFDS and electricalcabinet factor credit were actually less than prior to removal of the VEWFDScredit alone. While the licensee's explanation is sound for these metrics, itremains unclear as to why the LERF and A LERF increases do not display thesame trend as CDF and ACDF. If the CDF and A CDF showed a smallerincrease with the additional refinements, why did not the LERF and A LERF aswell? Explain why the increases in LERF and A LERF after removal of theVEWFDS credit and addition of the four refinements trended upward vs. thedownward trend for the CDF and A CDF increases.RESPONSE:During the review of the sensitivity and subsequent model refinement, moreattention was placed on the change in risk associated with CDF rather thanLERF. The reason for this is that the LERF risk and delta risk were much lowerthan the acceptable limits stated in RG 1.174. Since CDF was much closer tothese limits, more time was spent reviewing the scenarios that were contributingto the change in CDF risk. Furthermore, as a result of this RAI and others, theelectrical cabinet severity factor and credit for incipient detection in the MCB wereremoved from the baseline model altogether. The updated plant risk and deltarisk results will be provided with the supplemental response for PRA RAI 35.El-1 Response to Probabilistic Risk Assessment RAIsFarley PRA RAI 06.a.01In a letter dated November 12, 2013 (ADAMS Accession No. ML13318A027) thelicensee responded to PRA RAI 06(a) and stated that section V.2.2 of the LARprovides the details of the sensitivity analysis related to the bins that have analpha that is less than or equal to one. Indicate if the acceptance guidelines ofRegulatory Guide (RG) 1.174, "An Approach for Using Probabilistic RiskAssessment in Risk- Informed Decisions on Plant-Specific Changes to theLicensing Basis," may be exceeded when this sensitivity study for those binswith an alpha less than or equal to 1 is applied to the integrated study of PRA RAI35 (see below). If these guidelines may be exceeded, provide a description offire protection or other measures that can be taken to provide additional defensein depth (DID) (see FAQ 08-0048).RESPONSE:The types of scenarios and the location of the rooms that have the highestcontribution of risk as it relates to this sensitivity are located in rooms that containelectrical cabinets. These types of ignition sources are most predominantthroughout the plant and typically have the largest ZOI and therefore impacts themost targets. The types of rooms that contain these types of ignition sources areswitchgear, electrical penetration rooms, and general Auxiliary building hallways.The rooms that have the increased risk and delta risk have defined Defense inDepth actions in place, including safety margin evaluations. These are included inthe Fire Risk Evaluations for the specific fire areas.The updated base risk results includes the composite effect from the RAIresponses referenced in PRA RAI 35, with the exception of the bin 15 firefrequency discussed in our previous response to PRA RAI 28.k, which has beenjudged to have insignificant impact on the risk results. Updated total plant, FirePRA, and delta risk values will be submitted in PRA RAI 35 under separate coveronce completed.El -2 Response to Probabilistic Risk Assessment RAIsFarley PRA RAI 16.a.01In a letter dated October 30, 2013 (ADAMS Accession No. ML1 3305A1 05), thelicensee responded to PRA RAI 16.a and partially addressed some of the criteriafor assuming damage within MCR panels to be limited to the initiating panel,namely the presence of no openings and a double wall with an air gap. However,Appendix S of NUREG/CR-6850 also states that there be no sensitive electricalequipment in the adjacent cabinet (or else such equipment to have already been"qualified" above 82C), even with the double wall with air gap. Otherwisedamage to such equipment should be postulated. Explain whether theseadditional criteria are met or not. If the latter, explain how damage is modeled or,if not, the basis for assuming no damage. (Also see PRA RAI 33.a.01.)RESPONSE:Damage to adjacent panels in the Main Control Room (MCR) area is postulatedto be limited to the initiating panel provided there are no openings and the panelshave a double steel wall and an air gap between them as described in theresponse to PRA RAI 16.a based on the guidelines provided in Appendix S ofNUREG/CR-6850. If there are sensitive electronics within the adjacent panels,Appendix S of NUREG/CR-6850 notes that a damage time of ten minutes shouldbe assumed, based on a comparison of the interior temperatures of the testpanels as compared to the damage criteria for sensitive electronics provided inAppendix H of NUREG/CR-6850. Appendix S also notes that damage tosensitive electronics can be prevented if the fire is extinguished and the cabinet iscooled before ten minutes.Sensitive electronics are present in most electrical panels in the Farley MCRwithin the main control board (MCB) area, as well as the back panel area. Assuch, the additional criteria described in Appendix S of NUREG/CR-6850 relatingto sensitive electronics are applicable.Main Control Board:The MCBs also contain sensitive electronics; however, because there is nointernal separation between panels, a fire is assumed to propagate and damageadjacent targets in accordance with the methodology described in Appendix L ofNUREG/CR-6850 (see response to RAI 33.a.01).Non-Main Control Board Panels in the control room:The Fire PRA does not postulate damage to sensitive electronics in adjacentpanels that are separated by double wall construction with an air gap in the MCR(which includes the control room and the back panel area of the control room).The basis for this treatment is that the MCR is continuously staffed and thecontrol room operators are trained to quickly detect and mitigate the effects of afire that may occur in the electrical equipment. Per Appendix S of NUREG/CR-6850, damage to sensitive electronics in an adjacent panel can be prevented ifthe fire is extinguished and the exposing cabinet is cooled within ten minutes.Operations guidance for control room panel fires will be revised to emphasize theneed to evaluate the initial fire and to open the panel doors if the potential existsfor damage/overheating in an adjacent panel. In this case, the actions of theoperators are credited for interrupting the fire growth (and therefore the peakEl -3 Response to Probabilistic Risk Assessment RAIsinternal temperature) and exposure hazards of the panel ignited, includingopening the panel doors and cooling the adjacent panels. As noted in Appendix Sof NUREG/CR-6850, the full scale fire tests documented in NUREG/CR-4527indicated that open door panel fires did not cause the conditions in an adjacentpanel to exceed the sensitive electronic temperature threshold if the panel door isopen. Further, the actions of the operators are not necessarily postulated toextinguish the fire in the exposing panel; rather, they are assumed to delay orprevent further fire growth and propagation to adjacent panels until the time atwhich the fire brigade arrives and fully extinguishes the fire. A newimplementation item has been added to the LAR Attachment S, Table S-3 toaddress the additional Operations guidance, and is attached to this RAIresponse.This approach is further supported by the control room temperature dataassociated with the control room abandonment analysis. As described in Report0005-0030-003-001, Rev. 1 ("Evaluation of the Control Room AbandonmentTimes at the Farley Nuclear Power Plant"), the abandonment time for the 98'hpercentile heat release rate closed panel non-propagating (separated by doublewall air gap) fire scenarios is between 13 -17 minutes, depending on theventilation configuration. The hot gas layer temperature in the control room, whichis indicative of the hot gas layer temperature in the back panel area, isapproximately 600C (1400F) at this time, which is below the damage threshold of6500 (149°F) for sensitive electronics as described in Appendix H of NUREG/CR-6850. This ensures sufficient time for operator action to control the fire prior tocontrol room abandonment. Ultimately the hot gas layer temperatures reach 80 -90°C (1 76°F -194°F) and the hot gas layer depth descends to the floor;however, the fire model does not credit the actions the operators would take tomitigate the fire prior to abandonment and are thus conservatively biased. Inaddition, the maximum temperature is comparable to the damage temperaturethreshold of 8200 (1 800F) assumed in Appendix S of NUREG/CR-6850 fordamage to sensitive electronics in adjacent panels.As previously noted, the assumption that sensitive electronics are not damaged inadjacent panels separated by a double steel wall with an air gap in the controlroom area is considered applicable to panels in the main control board area aswell as the back panel area. The basis for this assumption is that the controlroom is continuously staffed and that operators are trained to mitigate the effectsof a fire using fire extinguishers and other fire protection equipment located in thegeneral area. Note that the back panel area may not be continuously staffed, butthe electrical panels in this area are within the control room HVAC envelope. Thismeans that smoke generated from fires at these panels will be identified in theirearly stages and early initiation of fire mitigation actions by the operator and thefire brigade, including efforts to limit the fire to the initiating source and to cool anyadjacent panels, will be implemented.E1-4 Response to Probabilistic Risk Assessment RAIsFarley PRA RAI 21.a.01In a letter dated September 16, 2013 (ADAMS Accession No. ML14038A019),the licensee responded to PRA RAI 21 .a and confirmed that the three severityfactors, 5.02E-4, 4.84E-4 and 0.00158, do not derive from Figure L-1 inNUREG/CR-6850 but are specifically calculated based on the type of ignitionsource, scenario location and abandonment time for the MCR abandonmentanalysis. The three severity factors correspond to the abandonment probabilitiesfor transient ignition sources, equipment room fixed ignition sources and MCRfixed ignition sources, respectively. Provide a discussion of the derivation ofthese factors, including their bases, e.g., as given in Section 13.2.1 of the FarleyScenario Development Report, PRA-BC-1 1-014, and Section 6 of Units 1 and 2Control Room Abandonment Times at the Joseph M. Farley Nuclear Plant,Rev 0.RESPONSE:OverviewThe computation of the severity factors (control room abandonment probabilities)is derived from the analysis of the control room abandonment time for each bin ofa heat release rate distribution (defined in Report 0005-0030-003-001 Rev. 1,"Evaluation of Control Room Abandonment Times at the Farley Nuclear PowerPlant", based on the Heat Release Rate probability distributions defined inNUREG/CR-6850, Appendix E). A non-suppression probability is determined foreach bin based on the time to abandonment calculated for that heat release ratebin from NUREG/CR-6850, Supplement 1, Chapter 14. The severity factor foreach heat release rate bin is multiplied by the non-suppression probabilitycorresponding to the time to abandonment for that heat release rate bin to obtainthe probability of abandonment for that heat release rate bin. The probability byheat release rate bin is summed for all heat release rate bins to calculate theprobability of abandonment for the particular control room fire configuration.Details of the Severity Factor Derivation MethodologyThe analysis as described in PRA-BC-F-1 1-014 Rev. 5 has been updated toaccount for the responses to other RAIs. The MCR Abandonment calculation hasbeen updated, Report 0005-0030-003-001, Rev. 1, "Evaluation of Control RoomAbandonment Times at the Farley Nuclear Power Plant". This report provides thetime at which the operators would abandon the control room given theabandonment temperature or visibility conditions defined in NUREG/CR-6850.The following is the updated severity factor derivation for the MCR Abandonmentscenarios at Farley.The analysis evaluates cases for the following Heat Release Rates:* Single cable bundle: closed electrical panel thermoset cable* Multiple cable bundle: closed electrical panel thermoset cable" Multiple cable bundle: closed electrical panel thermoset cable. The firepropagates to two adjacent panels after 10 minutes and two additionalpanels after 20 minutesE1-5 Response to Probabilistic Risk Assessment RAIs* Transient fire in an open location" Transient fire in a wall configuration" Transient fire in a corner configurationIt is assumed that the MCR will be ventilated by opening at least one door within15 minutes. In the development of the analysis for the MCR Abandonment, someof the panels located in the control room area were opened in support of thisanalysis but many of the panels were not. For this reason all panels will beconsidered to consist of multiple cable bundles.The FNP main control room area includes two distinct areas: the MCB panelarea and the back panels or "equipment area." These areas are somewhatisolated by panels, partitions, doors and walls such that gases/smoke transport islimited between the two spaces. While the main control room is continuouslyoccupied the equipment area is not. For this reason, there are three separatenon-suppression probabilities used in the abandonment calculation. A lookup ofthese times in Chapter 14 of Supplement 1 to NUREG/CR-6850, for ControlRoom Fires, electrical and transient fires, provides a non-suppression value.Abandonment cases are prepared for these areas/configurations individually;however the HVAC system is the same for both rooms and will therefore considerboth rooms to be part of the main control room analysis.Given the expected rapid fire brigade response due to continuous manning of thecontrol room, the 15 minute timeframe for opening a door to the control room isconsidered appropriate. The 15 minute time frame is further justified in responseto RAI FM 01a.The abandonment times are documented in Report 0005-0030-003-001, Rev. 1,"Evaluation of Control Room Abandonment Times at the Farley Nuclear PowerPlant". These times are used along with the non-suppression probabilities ofSupplement 1 to NUREG/CR-6850 (ext) to determine a cumulative case NSP forall heat release rates. Note that the MCB electrical cabinet severity factors arethe same as general electrical cabinets consisting of multiple cable bundles withpropagating fires.The computed NSPs are used to calculate a probability of abandonment thatrepresents the probability that a given fire will cause the operators to abandon thecontrol room. There were 13 abandonment severity factors calculated: five for theMCB panel area and eight for the Unit 1 & 2 equipment rooms and the outlyingareas within Fire Area 044. In addition to the HRR cases identified above, thereare also three HVAC configurations analyzed:0 HVAC not operating* HVAC operating normally* HVAC in purge modeFor a given fire scenario, one of the above configurations will be true.The following severity factor types were calculated using the above HRR casesand HVAC configurations:-MCREP.PROP.HVAC-OP: Main Control Room Electrical Panel Propagating, HVACoperating normallyE1-6 Response to Probabilistic Risk Assessment RAIs-MCREp-PROP-HVAC.INOP: Main Control Room Electrical Panel Propagating,HVAC not operating-MCREP.PROP-HVAC-PURGE: Main Control Room Electrical Panel Propagating,HVAC in purge mode-MCRTRAN-HVAC-OP: Main Control Room Transient, HVAC operating normally-MCRTRAN-HVAC-INOP: Main Control Room Transient, HVAC not operating-EREP.NOPROP.HVAC.OP: Equipment Room Electrical Panel Non-Propagating,HVAC operating normally-EREP.NOPROP.HVAC.INOP: Equipment Room Electrical Panel Non-Propagating,HVAC not operating-EREP-NOPROP-HVAC.PURGE: Equipment Room Electrical Panel Non-Propagating,HVAC in purge mode-EREP.PROP-HVAC-OP: Equipment Room Electrical Panel Propagating, HVACoperating normally"EREP-PROP-HVAC-INOP: Equipment Room Electrical Panel Propagating, HVAC notoperating-EREP.PROP.HVAC.PURGE: Equipment Room Electrical Panel Propagating, HVACin purge mode-ERTRAN-HVAC-OP: Equipment Room Transient, HVAC operating normally-ERTRAN-HVAC-INOP: Equipment Room Transient, HVAC not operatingWhere a propagating fire in the equipment room is defined as an ignition sourceinvolving two or more cubicles that are open to each other.E1-7 Response to Probabilistic Risk Assessment RAIsFarley PRA RAI 29.01In a letter dated September 16, 2013 (ADAMS Accession No. ML14038A019),the licensee responded to PRA RAI 29 and indicated that 22 supportingrequirements (SRs) fail to meet Capability Category (CC) II, 17 more than thestaff was able to determine by review of LAR Attachment V, Table V-I. Thelicensee response refers to dispositions in LAR Attachment V, Table V-i, which,while indicating how the licensee addressed the related findings and observations(F&Os), do not specifically explain why failing to meet CC-Il is acceptable fortransition under NFPA 805. Provide Table V-2 which explains the rationale foracceptability of less than CC-Il satisfaction for all 22 SRs.RESPONSE:The response to this RAI was provided in SNC letter dated April 23, 2014.E1-8 Response to Probabilistic Risk Assessment RAIsFarley PRA RAI 33.a.01In a letter dated October 30, 2013 (ADAMS Accession No. ML1 3305A1 05), thelicensee responded to PRA RAI 33.a and referenced PRA RAI 16.a. However,neither of the responses to PRA RAI 16.a or PRA RAI 33 discussed the timing fordetection and manual suppression prior to fire spread to adjacent cabinets.Furthermore, the response to PRA RAI 33.a indicates that all MCB panels arephysically open to one another. Discuss the basis for assuming rapid enoughdetection and manual suppression prior to fire spread into the adjacent cabinet.(See also PRA RAI 16.a.01.)RESPONSE:The MCB panel fires are postulated in a manner in which a fire starts at a givenpoint on the board and then propagates to the adjacent panel sections,regardless of the panel name or section. Damage to all targets in the scenariooccurs at t=0. Appendix L of NUREG/CR-6850 is credited from a manual non-suppression standpoint. The MCB sections are open to each other and the fire isconsidered to propagate to the adjacent sections of the MCB. The analysis doesnot assume rapid enough detection and manual suppression prior to fire spreadinto the adjacent cabinet (or section panels) in MCB. No specific credit fordetection and manual suppression is taken beyond that inherent in theNUREG/CR-6850, Appendix L Probability of Target Damage versus Distancecurves.For discussion of treatment of non-MCB electrical cabinets in the main controlroom, refer to the response to RAI PRA 16.a.01.El -9 Response to Probabilistic Risk Assessment RAIsFarley PRA RAI 33.c.01In a letter dated November 12, 2013 (ADAMS Accession No. ML13318A027), thelicensee responded to PRA RAI 33.c and indicated an intent to revise its MCRabandonment calculation as follows:"The CCDP for the abandonment scenario is based on failure of allactions in the control room. [A] conservative basis was used fordetermining the abandonment CCDP based on the calculated CCDPassociated with panel damage and failure of the MCR actions. The intentof this criteria is to ensure that the abandonment CCDP is an appropriatebounding value given that, shutting down the plant from outside thecontrol room has an inherently higher risk associated with it."These criteria are presented as (1) using conditional core damage probability(CCDP) = 0.1 if FRANC calculates a CCDP < 0.001, (2) using CCDP = 0.2 ifFRANC calculates a CCDP between 0.001 and 0.1, and (3) using 1.0 if FRANCcalculates a CCDP > 0.1. These FRANC-calculated CCDPs are based on bothMCB panel damage and failure of human actions in the MCR. Clarify how thesehuman actions were quantified, including any detrimental effects (increasedfailure probabilities) due to fire effects in the MCR. If screening or other boundingvalues were used, specify their bases, e.g., screening/scoping approach fromNUREG-1921, "Fire Human Reliability Analysis Guidelines" (or equivalent).RESPONSE:The evaluation of the plant risk given abandonment of the main control room(MCR) involves a two-step process for assignment of the conditional coredamage probability (CCDP) values. The first step in this process involves the useof FRANC to determine the characteristic CCDP for each of the fire scenariostreated in the MCR. There are approximately 100 scenarios that are individuallyevaluated for abandonment related risk contribution. Each of these scenarios arefirst quantified in FRANC assuming all MCR actions are 'failed' -set to TRUE.Because the underlying risk model that is used in the FRANC quantification doesnot address the impact of the shift of command and control from the control roomto the Hot Shutdown Panel for an abandonment scenario, this quantificationeffectively provides insights and characterization of the direct fire induced impactsgiven the originating fire. Based on this CCDP result, the second step in thequantification process determines the corresponding abandonment related CCDPfor each of these individual scenarios by manually resetting the CCDP to one ofthe values provided in the original RAI response (0.10, 0.20, or 1.0), to accountfor the shift of command and control to the hot shutdown panel.The assigned CCDP value (0.10, 0.20, or 1.0) is intended to reflect thecombination of human reliability (HEP) and hardware failures. As provided in theoriginal RAI response, a CCDP value of 0.10 is assigned only if thecorresponding CCDP is less than 1 E-3. Given that such a CCDP value wouldhave been determined without any credit for in-control room human actions (allMCR HFEs are set to TRUE), the resultant plant trip would be uncomplicated andnon-fire affected plant systems would continue to operate until they are manuallysecured as part of the prescribed abandonment procedure steps. It is noted that:El-10 Response to Probabilistic Risk Assessment RAIs* The quantification provides no credit for primary bleed and feed since theassociated HFE would have been set to TRUE" Any consequential loss of coolant event would have yielded a CCDP of1.0 as the human action to transition to recirculation from the containmentsump would have also been set to TRUE.* These considerations would infer that the time window for completion ofthe abandonment related action is relatively long.* A resultant CCDP of less than 1 E-3 indicates that offsite power isavailable.In this instance, the resultant assigned CCDP for the abandonment case of 0.10easily bounds the sum of the abandonment HEP and hardware failure probability.It is noted that a comparable result could be produced using the scopingguidance in NUREG-1921 and combining that result with the associatedhardware failures. Although not formally evaluated and included in the analysisdocumentation, Figure 5-5 and Table 5-5 of NUREG-1 921 would suggest that anHEP of 0.04 would be appropriate. Since offsite power is available in this set ofscenarios, the corresponding hardware failures would be expected to yield avalue much less than 0.06 (0.1 -0.04). Based on this comparison, the use of a0.10 value as the characteristic abandonment CCDP for those instances wherethe FRANC quantification yielded a CCDP of less than 1 E-3 is considered to bevery conservative.A CCDP value of 1.0 would be assigned if the FRANC generated CCDP is 0.10or greater. It is noted that any induced or consequential loss of coolant eventwould result in a FRANC generated CCDP of 1.0. This is because the dependentHFE to initiate recirculation from the containment sump will have been set toTRUE. For these cases, no HEP treatment is necessary as the scenario isconservatively assumed to result in core damage.The third possible assigned CCDP of 0.20 is used if the FRANC generated CCDPis 1 E-3 or greater but less than 0.10. As noted earlier, the FRANC calculatedCCDP is performed without any credit for in-control room human actions (all MCRHFEs are set to TRUE) and non-fire affected plant systems would continue tooperate until they are manually secured as part of the prescribed abandonmentprocedure steps. This FRANC quantification provides no credit for primary bleedand feed since the associated HFE would have been set to TRUE. In addition,any consequential loss of coolant event would have yielded a CCDP of 1.0 as thehuman action to transition to recirculation from the containment sump would havealso been set to TRUE. These considerations would infer that the time windowfor completion of the abandonment related action is relatively long.The scoping HEP of 0.04 from NUREG-1921 for the case when a CCDP of 0.10is used, would also be applicable in this case. However, the hardware relatedfailures involves a much wider range of possible values. An upper bound valueof 0.10 would therefore be applicable in this case. When these two terms arecombined, the result of 0.14 is still below the assigned value of 0.20. As such,the treatment remains conservative and bounding.El-11 Response to Probabilistic Risk Assessment RAIsThe process that was used in the Farley fire PRA for the treatment of MCRabandonment appropriately considered the combination of human errorprobability and equipment hardware failure likelihood. Although a formal detailedHRA was not performed nor was NUREG-1 921 specifically applied for the MCRabandonment scenarios, the results are expected to bound the results from adetailed HRA. As such, the resultant risk estimates are considered to have beenadequately characterized for the purposes of this NFPA 805 application.E1-12 Response to Probabilistic Risk Assessment RAIsFarley PRA RAI 35Section 2.4.3.3 of the NFPA 805 standard incorporated by reference into 10 CFR50.48(c) states that the PSA approach, methods, and data shall be acceptable tothe AHJ, which is the NRC. Regulatory Guide (RG) 1.205, "Risk-Informed,Performance-Based Fire Protection for Existing Light-Water Nuclear PowerPlants," identifies NUREG/CR-6850 as documenting a methodology forconducting a Fire PRA (FPRA) and endorses, with exceptions and clarifications,NEI 04-02, "Guidance for Implementing a Risk-Informed, Performance-BasedFire Protection Program Under 10 CFR 50.48(c)," Rev. 2, as providing methodsacceptable to the staff for adopting a fire protection program consistent withNFPA 805. RG 1.200, "An Approach for Determining the Technical Adequacy ofProbabilistic Risk Assessment Results for Risk-Informed Activities," describes apeer review process utilizing an associated ASME/ANS standard (currentlyASME/ANS-RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard forLeveil/Large Early Release Frequency Probabilistic Risk Assessment forNuclear Power Plant Applications") as one acceptable approach for determiningthe technical adequacy of the PRA once acceptable consensus approaches ormodels have been established. In a letter dated July 12, 2006 to NEI (ADAMSAccession No. ML061660105), the NRC established the ongoing FAQ processwhere official agency positions regarding acceptable methods can bedocumented until they can be included in revisions to RG 1.205 or NEI 04-02.Section 2.4.4.1 of NFPA 805 states that the change in public health risk arisingfrom transition from the current fire protection program to an NFPA 805 basedprogram, and all future plant changes to the program, shall be acceptable to theAHJ, which is the NRC. RG 1.174, "An Approach for Using Probabilistic RiskAssessment in Risk-Informed Decisions on Plant-Specific Changes to theLicensing Basis," provides quantitative guidelines on CDF and LERF andidentifies acceptable changes to these frequencies that result from proposedchanges to the plant's licensing basis and describes a general framework todetermine the acceptability of risk-informed changes.As stated on page B-1 of Appendix B of PRA-BC-F-1 1-004, "Fire PRA LogicModel," the new Westinghouse Shutdown Shield (SDS) was installed in fall 2010.The internal events PRA (IEPRA), upon which the FPRA is based, takes credit forthe SDS (failure rate of 0.0271/demand), limiting the leakage rate to 2 gpm wherethe faces of the SDS seal components remain in contact. The assumed leakagerate is increased to 19 gpm if the SDS actuates but the pump shaft continues torotate if not tripped in a timely manner. Finally, if the SDS does not actuate at all,"existing" (Westinghouse Owners Group (WOG) 2000 or Rhodes Model) sealmodel leakage rates are applied. Given the July 26, 2013, 10 CFR Part 21notification by Westinghouse concerning defects with the SDS performance,provide a sensitivity evaluation that removes all credit for the SDS package,including both probability and consequences as appropriate. Provide revisedestimates of CDF, LERF, A CDF and A LERF, including non-fire hazards for CDFand LERF, as a result of removal of this credit. Should this result in any changesto conclusions regarding the transition satisfying RG 1.174 risk/A risk guidelines,address any changes that will be made to accommodate this.When performing this analysis, include the composite effect from all previous re-evaluations, including any synergistic effects, specifically including the following:E1-13 Response to Probabilistic Risk Assessment RAIsa. From the LAR and the December 20, 2012 LAR Supplement,sensitivities related to the electrical cabinet fire severity method(Section V.2.1) and use of control power transformer (CPT)(Section V.2.3; also response to PRA RAI 08.a).b. From the RAI Responses dated September 16, 2013 (ADAMSAccession No. ML14038A019):i. PRA RAI 01 .a -Removal of credit for VEWFDS in theMCR (also PRA RAI 01.01)ii. PRA RAI 15.a -Revised seismic CDF based on 2008USGS dataiii. PRA RAI 28.k -Validity of current Ignition Bin 15 firefrequenciesc. From the RAI Responses dated November 12, 2013 (ADAMSAccession No. ML13318A027):i. PRA RAI 07.e -Use of 0.1 CCDP for MCR Abandonmentii. PRA RAI 17.d- Turbine Building Collapseiii. PRA RAI 33.c -Revised MCR Abandonment analysis (alsoRAI PRA 33.c.01)RESPONSE:Reactor Coolant Pump (RCP) Seal ModelThe Farley Internal Events PRA model and Fire PRA model for the NFPA 805License Amendment Request (LAR) includes credit for the Westinghouseshutdown seal (SDS) as outlined in Revision 1 to Pressurized Water ReactorOwners Group Topical Report WCAP-1 7100-P/NP, "PRA Model for theWestinghouse Shut Down Seal." SNC is aware of the operating experienceissues with the Westinghouse shutdown seals which resulted in the 10 CFR Part21 notification, and how that experience does not match the SDS performancecredited in the Farley model which is based on the topical report.In response to the issues which resulted in the 10 CFR Part 21 notification,Westinghouse has developed a new version of the SDS which has undergone arigorous testing program. SNC plans to install this next generation shutdown sealfrom Westinghouse as a corrective action resolution for the issues identified inthe Part 21 notification. Installation of this next generation shutdown seal in eachFNP RCP will be added as an installation item to Attachment S.The current modeling of the SDS performance in the PRA for NFPA 805 reflectsthe original PRA modeling guidance for the Westinghouse shutdown seal(WCAP-171 00-P/NP Rev. 1). PRA modeling guidance specific to this nextgeneration (Generation Ill) SDS is in the development and review process forsubmittal to the NRC. SNC expects the RCP SDS risk benefit using this model tobe consistent with the benefit currently being determined using the previousWCAP-1 71 00-P/NP, Rev. 1 model. SNC will monitor Westinghouse and PWROwners Group efforts relative to issuance of this modeling guidance and NRCreview.As stated in the License Amendment Request Section 4.8.2, "Plant Modificationsand Items to be Completed During the Implementation," following the installationof modifications and the as-built installation details, additional refinementsEl-14 Response to Probabilistic Risk Assessment RAIssurrounding the modification may need to be incorporated into the Fire PRAmodel, and the Fire PRA will verify the validity of the reported change-in-risk onas-built conditions after the modification is completed. This is captured underimplementation item 30 in Table S-3, "Implementation Items," and is applicable tothe installation of this next generation shutdown seal in each FNP RCP. Themodel will be revised to incorporate any new data or modeling requirements andfurther refinements or modifications will be addressed as necessary.Composite Effect from all Previous Re-Evaluationsa) From LAR and the December 20, 2012 LAR Supplement:i) NFPA 805 LAR Submittal Section V.2.1 -The electrical cabinet fireseverity method is no longer used in the Farley Fire PRA.ii) NFPA 805 LAR Submittal Section V.2.3 and RAI PRA 08(a) (ML12359A051) -The hot short-induced spurious operation conditionallikelihood estimates used in the development of the Fire PRA hasbeen updated to reflect the guidance provided in ML14086A1 65,"Supplemental Interim Technical Guidance on Fire-Induced CircuitFailure Mode Likelihood Analysis". This guidance has beenincorporated into the base model. The original response to RAI PRA08(a) indicated that this issue would be addressed via a sensitivityanalysis. This response supersedes the previous response and willincorporate this change into the baseline Fire PRA.b) From RAI Responses dated September 16, 2013:i) PRA RAI 01 .a -Credit for VEWFDS has been eliminated from thecontrol room main control board area (fire zone 044). Credit forVEWFDS modifications is taken for electrical panels located behindthe main control room vertical boards within the control roomenvelope. (See RAI 01.a response, including reference to supportingdocument RBA 13-006-F, Sensitivity Study of VEWFDS credited in theMCR, dated August 9, 2013)ii) PRA RAI 15.a -The updated seismic contribution to the total plantrisk is being used.iii) PRA RAI 28.k -The number of vertical sections of 1 F Switchgear inRoom 0335 was corrected to 15 from 16 and sensitivity analysis wasperformed for the discrepancy of total plant count of electricalcabinets. The sensitivity analysis provided in the RAI 28.k submittaldemonstrated that the associated discrepancy had an insignificantimpact on the Fire PRA. Therefore, this change is not included in theupdated fire PRA risk evaluation in support of the NFPA 805 LAR.This change will be incorporated in conjunction with other futurechanges as appropriate.c) For RAI Responses dated November 12, 2013:i) PRA RAI 07.e -The methodology of applying a 0.1 CCDP for allcontrol room abandonment scenarios is no longer credited in the FirePRA. See also PRA RAI 33.c response in the November 12, 2013submittal as well as the response to PRA RAI 33.c.01 accompanyingthis RAI response submittal.El-15 Response to Probabilistic Risk Assessment RAIsii) PRA RAI 17.d -The ignition frequency and suppression terms for theTurbine Building Collapse scenarios have been updated.iii) PRA RAI 33.c -See response to PRA RAI 33.c.01 for a discussionand clarification of the methodology used for the MCR Abandonmentanalysis.The updated base risk results includes the composite effect from the RAIresponses referenced above, with the exception of PRA RAI 28.k above, whichhas been judged to have insignificant impact on the risk results.Updated total plant, Fire PRA, and delta risk values will be submitted underseparate cover once completed. This will include a revised NFPA 805 LAR,Attachment W as well as updates to Attachments C, G, S, and V.El-16 Joseph M. Farley Nuclear PlantResponse to Request for Additional InformationRegarding License Amendment Request for Transition to 10 CFR 50.48(c)NFPA 805 Performance Based Standard for Fire Protection for Light WaterReactor Generating PlantsEnclosure 2Supplemental Response to Safe Shutdown Analysis RAI Supplemental Response to Safe Shutdown Analysis RAIFarley Safe Shutdown Analysis (SSA) Request for Additional Information(RAI) 10.01In a letter dated October 30, 2013 (Agencywide Documents Access andManagement System (ADAMS) Accession No. ML1 3305A1 05), the licenseeresponded to SSA RAI 10 and indicated that no recovery actions (RAs) wereomitted from license amendment request (LAR) Attachment G, and that thecorrelation of RAs to variances from deterministic requirements (VFDRs) wereprovided in LAR Attachment G, Table G-1, of the LAR supplement datedDecember 20, 2012 (ADAMS Accession No. ML12359A050).However, the NRC staff is providing the following examples of inconsistencesbetween: LAR Attachment C, Table C-1 and LAR Attachment G, Table G-1; withthe LAR supplement and the RAI responses:a. The LAR Attachment G, Table G-1 submitted on December 20,2012, is missing many fire areas compared to the original LARAttachment G, Table G-1 provided in September, 2012 (e.g., mostof U2-040, U2 2-040, U2 2-041, U2 2-075, and U2 2-076... up toU2-2-021).b. Components identified in the new LAR Attachment G, Table G-1don't correspond to VFDRs in LAR Attachment C, Table C-1 (e.g.,Q1P16V0530 and Q1P16V0593).c. Components corresponding to VFDRs in the LAR Attachment C,Table C-1 are not identified in the new LAR Attachment G, TableG-1 (e.g., VFDRs: U1-044-PCS-040 and U1-1-040-PCS-186), andthere are potential duplicate VFDRs (e.g., U1-1-040-PCS-145 &146, U1-044-PCS-127 & 128 and U2-044-PCS-079 & 155)d. A partial LAR Attachment G, Table G-1 was submitted with theRAI responses on November 12, 2013 (ADAMS Accession No.ML13318A027), which includes LAR pages G-9 through G-26. This table has corrections and multiple entries needing to beremoved as duplicates. However, this LAR Attachment G, TableG-1 still includes several components (e.g., OP-RECOV-XXXX)that are not in LAR Attachment C, Table C-1.Provide LAR Attachment G, Table G-1 and LAR Attachment C, Table C-1 that areup-to-date and correlate accordingly.SUPPLEMENTAL RESPONSE:b. The component(s) identified in LAR Attachment G, Table G-1 arecomponent(s) made available by application of recovery action(s). Incases where a VFDR requires a recovery action (to satisfy the risk ordefense-in-depth criteria of NFPA 805, Ch. 4.2.4) alternate or redundantcomponents are restored to meet the performance criteria. Theserecovered components are identified in Attachment G, Table G-1. Thecomponents associated with the VFDR are documented in Attachment C,Table C-1. The previous LAR Att. G tables that were submitted did notinclude the recovery action type to VFDR to recovered componentE2-1 Supplemental Response to Safe Shutdown Analysis RAIcorrelation; however, to facilitate more complete mapping this has beenrevised with the updated Attachment G included in this response.To specifically address the examples in the RAI the following detail isprovided:SE(I.0t51 W601-008 Atijacti,mntl FRt foF Unit I (Aitoewate Sbutdowni) ý ue Af t*a 044Repred MU Recovery ActionC91om1ome"t1 Cogin of t Omoti. Wt-R Reovera cavi er Adcaw OIPi6smis Train a OW to Diese &i"V 114511511 Remrove power ftm senic wate velessQ1P116W2201PI1W83001PI6VW23OIPIOVU3I101IP16v063601P16V038GO1PVW54UWi 2 SW $upl 10o EDO 18Uri 2 SW Return ftm EDOUnit 1SW SUPp to EDO 18WUi tSWR ftoMr EDOG 1BTran 8 SW Rfetm from DieselTrain 8 sW Discierge to PoNdTrain 8 SW DischaMe to RiverasOC ted V cooi fat EDO lB by openirgtie supply dra*i tlu at MCC IT. Powerto these vae Is ramoved Pio to starln BEDO to pmtu aPan of lIe vaveswon EDG 18 Is p nacd i serwce.nSW to EDO Is not avaal" foM Unt 1t.manwsW opw lhe Vto 2 SW supply ad runvatv to EDO 18 Thes vaves so VOWed becaMe tI"e a nomalny dDsdThese ac s ov-ue ta SW is povded to theEDG tom t lest one of ft urit SW systemAssociated VFORs Reference(Components Affcted by FireListed m At& C)Table G-1 (Componeft Recovered -Nlot Necessarly WOUR Cofiponent)Fire AreaQIP16VO530 UI 044 145-151 SE-C051326701-O08(FRE)Table 6-1 (Component Recovered -Fire Area Associated VFORs ReferenceNot Necessarily VFDR Component) (Components Affected by FireListed in AtLC)Q1P16V0593 U2 044 7,51-55,145,146 SE-C051326701-.08(FRE)St 0~r3A6?0 008r- Atnhm [RE $m Unwit ? (Afternate MShuldown) i mi A~a 0)44Rer*4ud OR ReCOMeY AcuonecompoOent U) componen t Oescn1 VFOR Roe"" "Con 086"I1116QP6V051 Train sW to Diese auk** 7.51- Remove pmwr krn sv wate valvesQ1P16VU9W thit 1SW SuPply to EDO 28 55, asmodoad wlbh cooiq tor EOG 28 by opening IteQ0P16W" 3 Un I 9W RPeltur tm EDG 2B 145,146 spl dmcdi brears A MCC IP, 2DO, 2V md02P16VU59 Wli 2 SW 91pl to EDG 28 2T. Power to th"ese vetoe is removed rIor to02PI8V03 Wi 2 SW Redun trom EDO 28 staring the EDO to prevent spurious operaon ofC2P16VO536 Train 8 SW Retu tom Desel "th vamves one EDG 29 1 piaced in seviceQ2P06O38 BDWOQ2P16V-545 TraMi 8 W Obdise to Pond If SW to EDO 2B is not avaal e eom Uit 2,Train S8W DOidwe to River ,nar"Aly open oh Ut sw Upply mad end vaalesto EDO 28. The vaes me i uni rsysoembecause thwey onarmaly dosedThese acbon enans tha SW Is provkWe to theEDO tOM at Weast one of fte WWls SW system,E2-2