IR 05000285/2006018

EA-07-047 Omaha Public Power District (Fort Calhoun Station FC-2-4 Adm.), Final Significance Determination for a White Finding and Notice of Violation; NRC IR05000285/2006018 - Fort Calhoun Station), 05/29/2007

ML071500074
Person / Time
Site:	Fort Calhoun
Issue date:	05/29/2007
From:	Mallett B Region 4 Administrator
To:	Ridenoure R Omaha Public Power District
References
EA-07-047, IR-06-018
Download: ML071500074 (13)

Text

May 29, 2007EA-07-047R. T. RidenoureVice PresidentOmaha Public Power DistrictFort Calhoun Station FC-2-4 Adm.P.O. Box 550Fort Calhoun, NE 68023-0550SUBJECT:FINAL SIGNIFICANCE DETERMINATION FOR A WHITE FINDING ANDNOTICE OF VIOLATION; NRC INSPECTION REPORT 05000285/2006018 -FORT CALHOUN STATION

Dear Mr. Ridenoure:

The purpose of this letter is to provide you the final results of our significance determination ofthe preliminary White finding identified in the subject inspection report. The NRC's finalrisk-informed conclusion is that the violation of an NRC requirement discussed in this letter isbest characterized as a White finding. Our rationale for this conclusion is discussed below, aswell as in Enclosure 2.Our preliminary finding was discussed with your staff during an exit meeting on February 13,2007. The inspection finding was addressed using the Significance Determination Process andwas preliminarily characterized as White, a finding with low to moderate increased importanceto safety that may require additional NRC inspections. This finding involved the improperinstallation of the valve disk of Containment Spray Header Isolation Valve HCV-345. Theimproper installation resulted in a condition in which the actual position of the valve was nearlyopposite of the indicated position.This finding also represented a violation of NRC requirements. The violation involved theconduct of maintenance activities on valve HCV-345 without procedures or work instructionsappropriate to the circumstances and the failure to have appropriate work instructions toconduct adequate post-maintenance testing prior to returning the valve to service. Thisviolation resulted in an inoperable train of the containment spray system for an entire operatingcycle and also provided a reactor coolant system diversion flow path if shutdown cooling (SDC)was initiated following certain postulated accident conditions.The NRC's preliminary assessment of the safety significance of this inspection finding, which isdocumented in Attachment 2 of NRC Inspection Report 05000285/2006018 (ML070640155)resulted in an increase in core damage frequency (CDF) for internal events of 5.7E-6/year, orWhite for safety significance. The NRC's assessment of this issue was conducted using the 2NRC's Risk-Informed Inspection Notebook for Fort Calhoun Station, the NRC's probabilistic riskassessment (PRA) Standardized Plant Analysis Risk (SPAR) model for Fort Calhoun Station,and the SPAR-H Human Reliability Analysis Method (NUREG/

CR-6883). Our preliminaryassessment included, but was not limited to, the following assumptions: Containment SprayTrain B was functional with the internals of Valve HCV-345 installed incorrectly; the exposuretime for the condition was 454 days; and consideration that if an accident occurred whichinvolved operators initiating SDC prior to a recirculation actuation signal, then reactor coolantwould be diverted from the reactor coolant system. The NRC and your staff agreed upon therisk models and the accident sequences. However, there were differences between yourevaluation and the NRC's preliminary significance determination analysis in regard toperformance shaping factors (PSFs) used in the PRA of this issue. Additionally, your data andanalysis of the external initiators/events were not available for our evaluation at the time of theissuance of the report.At the request of the Omaha Public Power District (OPPD), a Regulatory Conference was heldon April 16, 2007, to discuss OPPD's position on the safety significance of the finding andcorrective actions taken in response to the improperly installed valve disk. During theRegulatory Conference, OPPD agreed with the apparent violation as characterized in NRCInspection Report 05000285/2006018, and your staff described the corrective actions taken inresponse to the finding. However, your staff asserted that the safety significance was very low,or at a Green level. Your staff's analysis and conclusions are included as an enclosure to theRegulatory Conference Meeting Summary (ML071160393), issued on April 26, 2007.In response to questions from NRC staff, OPPD provided additional information on April 23,2007. This information included: "HRA Review Comments and Recommendations on OPPDHCV-345 Mispositioning SDP, Revision 1," (Scientech letter SEA-JFG-07-005) and, "AdditionalFire Perspectives for HCV-345 Containment Spray Valve" (ML071320002). The NRC staffconsidered the additional information provided by your staff in performing the final significancedetermination for this issue.After careful consideration of the information developed during the inspection, the informationyour staff provided at the Regulatory Conference, and the revised information you provided onApril 23, 2007, the NRC has concluded that the inspection finding is appropriately characterizedas White. We estimate the change in core damage frequency associated with this condition tobe 4.6E-6/year, as discussed in Enclosure 2 to this letter. The NRC staff agrees with most ofyour assumptions and analyses of the applicable accident scenarios. However, severaldifferences account for the variances between OPPD's significance evaluation and the NRC'sfinal analyses. These differences include the assumptions in the PSFs of three human error probabilities (HEPs), the method of performing a dependency analysis for the HEPs, and thetreatment of the fire-induced loss of offsite power (LOOP) events multiplication factor.Regarding the PSFs utilized in your analysis, the NRC determined that your evaluation givesmore credit for successful operator actions than is considered reasonable based on guidance inNUREG/CR-6883. The NRC concluded that "Nominal" credit should be given to these PSFsbased upon the complexity of the modeled event and the level of education and training ofoperators. In giving "Nominal" credit, the NRC acknowledges that recognizing a postulatedreactor coolant system leak would not be difficult to perform and would have little ambiguity, andthat your staff had an adequate amount of formal schooling and instruction to ensure they wereproficient in day-to-day operations and had been exposed to abnormal conditions. Detailed 3descriptions of the NRC's conclusions regarding the PSFs are provided in Enclosure 2. For theHEP dependency analysis for adding additional high-pressure safety injection flow, your staffdetermined that there was "low" dependency for this HEP because of "additional cues" thatwere available. However, the NRC concluded there is a "moderate" dependency becausecrediting the reactor vessel level monitoring system as an additional cue may provideinsufficient time to prevent core damage. For the external contributor of fire-induced LOOPevents, your staff stated that a multiplication factor is not required. However, based on theguidance in NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear PowerFacilities," the NRC staff concluded that the use of a multiplication factor of 10 is appropriate. The NRC concluded that these fire scenarios could involve events in which significant amountsof smoke would be generated, offsite fire fighting capability coul d be needed, and multipleconcurrent activities affecting plant operators would be occurring. The NRC considered that thisscenario is consistent with the NUREG/CR-6850 philosophy of increasing the HEP by a factor of10 to account for a "minor" increase in operator workload. You have 30 calendar days from the date of this letter to appeal the staff's determination ofsignificance for the identified White finding. Such appeals will be considered to have merit onlyif they meet the criteria given in the NRC Inspection Manual Chapter 0609, "SignificanceDetermination Process," Attachment 2, "Process for Appealing NRC Characterization ofInspection Findings (SDP Appeal Process)."The NRC has also determined that the improper installation of the Valve HCV-345 is a violationof 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," as citedin the enclosed Notice of Violation (Notice). In accordance with the NRC Enforcement Policy,the Notice is considered escalated enforcement action because it is associated with a Whitefinding. You are required to respond to this letter and should follow the instructions specified inthe enclosed Notice when preparing your response.In addition, we will use the NRC Action Matrix, as described in NRC Inspection ManualChapter 0305, "Operating Reactor Assessment Program," to determine the most appropriateNRC response and any increase in NRC oversight. We will notify you by separatecorrespondence of that determination.In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, itsenclosures, and your response will be made available electronically for pub lic inspecti on in theNRC Public Document Room or from the NRC's document system (ADAMS), accessible fromthe NRC Web site at www.nrc.gov/reading-rm/adams.html. To the extent possible, yourresponse should not include any personal privacy, proprietary, or safeguards information so thatit can be made available to the public without redaction.

Sincerely,

/RA TPGwynn for/

Bruce S. MallettRegional AdministratorEnclosures: (see next page)

4Docket No. 50-285License No. DPR-40Enclosures:1. Notice of Violation2. Final Significance Determinationcc w/enclosures:Joe l. McManis, Manager - LicensingOmaha Public Power DistrictFort Calhoun Station FC-2-4 Adm.P.O. Box 550Fort Calhoun, NE 68023-0550David J. BannisterManager - Fort Calhoun StationOmaha Public Power DistrictFort Calhoun Station FC-1-1 PlantP.O. Box 550Fort Calhoun, NE 68023-0550James R. CurtissWinston & Strawn1700 K Street NWWashington, DC 20006-3817ChairmanWashington County Board of SupervisorsP.O. Box 466Blair, NE 68008Julia Schmitt, ManagerRadiation Control ProgramNebraska Health & Human ServicesDept. of Regulation & LicensingDivision of Public Health Assurance301 Centennial Mall, SouthP.O. Box 95007Lincoln, NE 68509-5007Daniel K. McGheeBureau of Radiological HealthIowa Department of Public HealthLucas State Office Building, 5th Floor321 East 12th StreetDes Moines, IA 50319 5DISTRIBUTION:RIDSSECYMAILCENTERRIDSOCAMAILCENTERRIDSEDOMAILCENTERRIDSOEMAILCENTERRIDSOGCMAILCENTERRIDSNRRODRIDSNRRADIPRIDSOPAMAILRIDSOIMAILCENTERRIDSOIGMAILCENTERRIDSOCFOMAILCENTERRIDSRGN1MAILCENTERRIDSRGN2MAILCENTERRIDSRGN3MAILCENTEROEWEBRIDSNRRDIPMIIPBROPreportsvia e-mail:Gwynn - TPGMallett - BSM1Fuller - KSFR4ALLEGEMaier - WAMHowell - ATHVegel - AXVChamberlain - DDCCaniano - RJC1Clark - JACPowers - DAPKirkland - JCK3Paulk - CJPDricks - VLDHanna - JDH1Solorio - DLS2Carpenter - CACOEMAIL Starkey - DRSAshley - MABHaire - MSH2Vasquez - GMVTrocine - LXTFCS Site Secretary - BMMSUNSI Review Completed: _JAC__ADAMS: Yes G No Initials: _JAC

___ Publicly Available G Non-Publicly Available G Sensitive Non-SensitiveC:\FileNet\ML071500074.wpdRIV:RI:DRP/ESRA:DRSD:DRSD:DRP JAClarkRLBywaterDDChamberlainATHowell

/RA//MRunyan for RA//RA//RA/

5/10/075/09/075/09/075/11/07ACESD:ACESOENRRDRARAMHaireKFullerDSolorioMAshleyPGwynnBMallett gmv for /RA//RA/ /RA//RA//RA TPGwynn for/5/9 /075/14 /075/22 /075/18/075/29/075/29/07OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax NOTICE OF VIOLATIONOmaha Public Power DistrictDocket No. 50-285Fort Calhoun StationLicense No. DPR-40EA-07-047During an NRC inspection completed on February 13, 2007, a violation of NRC requirementswas identified. In accordance with the NRC Enforcement Policy, this violation is listed below:10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, andDrawings," states, in part, that activities affecting quality shall be prescribed bydocumented instructions, procedures, or drawings of a type appropriate to thecircumstances and shall be accomplished in accordance with these instructions,procedures, or drawings. Contrary to the above, in May of 2005, Fort Calhoun Station personnelaccomplished maintenance activities without procedures appropriate to thecircumstances. Specifically, the licensee performed maintenance and post-maintenance activities on Containment Spray Header Isolation Valve HCV-345using procedures that were not appropriate to the circumstances because theprocedures did not require actions to verify the correct orientation of the valve. As a result, the valve was installed in the wrong orientation during maintenance,and post-maintenance testing did not detect the improper reassembly prior toreturning the valve to service. This failure caused one train of the ContainmentSpray system to be inoperable from May 11, 2005 to September 9, 2006.This violation is associated with a White SDP finding.Pursuant to the provisions of 10 CFR 2.201, Omaha Public Power District is hereby required tosubmit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN:Document Control Desk, Washington, DC 20555-0001 with a copy to the RegionalAdministrator, Region IV, and a copy to the NRC Resident Inspector at Fort Calhoun Station,within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This replyshould be clearly marked as a "Reply to a Notice of Violation; EA-07-047" and should include:(1) the reason for the violation, or, if contested, the basis for disputing the violation or severitylevel, (2) the corrective steps that have been taken and the results achieved, (3) the correctivesteps that will be taken to avoid further violations, and (4) the date when full compliance will beachieved. Your response may reference or include previous docketed correspondence, if thecorrespondence adequately addresses the required response. If an adequate reply is notreceived within the time specified in this Notice, an order or a Demand for Information may beissued as to why the license should not be modified, suspended, or revoked, or why such otheraction as may be proper should not be taken. Where good cause is shown, consideration willbe given to extending the response time.ENCLOSURE - 1 2 ENCLOSURE - 1If you contest this enforcement action, you should also provide a copy of your response, withthe basis for your denial, to the Director, Office of Enforcement, United States NuclearRegulatory Commission, Washington, DC 20555-0001.Because your response will be made available electronically for public inspection in the NRCPublic Document Room or from the NRC's document system (ADAMS), accessible from theNRC Web site at www.nrc.gov/reading-rm/adams.html, to the extent possible, it should notinclude any personal privacy, proprietary, or safeguards information so that it can be madeavailable to the public without redaction. If personal privacy or proprietary information isnecessary to provide an acceptable response, then please provide a bracketed copy of yourresponse that identifies the information that should be protected and a redacted copy of yourresponse that deletes such information. If you request withholding of such material, you mustspecifically identify the portions of your response that you seek to have withheld and provide indetail the bases for your claim of withholding (e.g., explain why the disclosure of information willcreate an unwarranted invasion of personal privacy or provide the information required by10 CFR 2.390(b) to support a request for withholding confidential commercial or financialinformation). If safeguards information is necessary to provide an acceptable response, pleaseprovide the level of protection described in 10 CFR 73.21.Dated this 29 th day of May 2007 Final Significance DeterminationFort Calhoun StationContainment Spray Header Isolation Valve HCV-345The NRC reviewed the information provided by the licensee during the Regulatory Conferenceheld on April 16, 2007. In response to questions from the staff, the licensee provided additionalinformation on April 23, 2007. This information included: "HRA Review Comments andRecommendations on OPPD HCV-345 Mispositioning SDP, Revision 1," (Scientech letter SEA-JFG-07-005); and "Additional Fire Perspectives for HCV-345 Containment Spray Valve." Usingthe additional information provided by the licensee, the NRC staff performed a final significancedetermination by modifying the original evaluation, as appropriate. The documentation thatfollows is not a stand-alone eval uation; the reader must be familiar wi th the preliminarysignificance determination documented in NRC Inspection Report 05000285/2006018 (ADAMSML070640155).The NRC concluded that the preliminary significance determination (that the finding was oflow-to-moderate safety significance) remained unchanged.The following summarizes the NRC staff's review of the licensee's information provided at theRegulatory Conference and afterwards.I.Internal Events:a.The NRC staff reviewed the licensee's approach of binning the initiating events ofinterest into four categories. Category 3 (dry containment sump at time ofshutdown cooling initiation) and Category 4 (wet containment sump at time ofshutdown cooling initiation) were potential contributors to core damage.The NRC staff concluded the licensee's frequency estimates were reasonable. The frequency of Category 3 events was approximately 1.69E-2/year and thefrequency of Category 4 events was approximately 1.06E-2/year.b.The NRC staff reviewed the licensee's approach of developing event trees toevaluate the Category 3 and Category 4 events using the four top events:COGEARLY, COMCOG, HPSIFLOW, and XSPRAYVALVE. These top eventswere reasonable representations of operator response to indications of a flowdiversion through the incorrectly positioned containment spray valve. Theyprovided a framework for evaluation using the NRC's SPAR-H method of estimating human erro r probabilities (HEPs).c.The NRC staff reviewed and performed independent evaluations of the licensee'sHEP estimates of the four top events. A summary of this evaluation is includedbelow.ENCLOSURE - 2 2ENCLOSURE - 2 COGEARLYFor top event COGEARLY (representing operator failure to detect and diagnosethat a loss of coolant accident (LOCA) has been caused by flow diversion given adry containment sump initially), the licensee's estimate of the the total HEP ofCOGEARLY was 1.1E-3. The NRC staff concluded this was a reasonableestimate.COMCOGFor top event COMCOG (representing operator failure to detect and diagnosethat a LOCA has been caused by the flow diversion given a wet containmentsump) the licensee's estimate of the HEP was 5E-6. The NRC staff disagreedwith the licensee's performance shaping factor (PSF) levels for Complexity andExperience/Training.For Complexity, the licensee concluded that the PSF level was "ObviousDiagnosis" but the NRC staff concluded the PSF level was best represented as"Nominal." Using the guidance in NUREG/CR-6883, "The SPAR-H HumanReliability Analysis Met hod," the NRC concl uded that det ermining a reactorcoolant system leak had occurred was "not difficult to perform and had littleambiguity." These are the criteria for a Nominal PSF level result. The NRCconcluded the cues for a LOCA would not be so compelling in this context towarrant an order of magnitude reduction that a PSF level of "Obvious Diagnosis"would provide.For Experience/Training, the licensee concluded that the PSF level was "High"but the NRC concluded the PSF level was best represented as "Nominal." TheNRC concluded that operating crews had an adequate amount of formal trainingand instruction to ensure they were proficient in day-to-day operations and hadbeen exposed to abnormal conditions. These are the criteria for a Nominal PSFlevel result. The "High" level was considered but required demonstratedmaster-level experience. Because of the extended duration of the postulatedevent, and the need to consider the performance capabilities of an average crew,Nominal was chosen.The NRC concluded the HEP estimate for COMCOG was 1E-4, as shown below:COMCOGPSFDiagnosis Multiplier(Base = 0.01)Action Multiplier (Base = 0.001)Available TimeExpansive Time (0.01)N/AStressHigh (2.0)N/AComplexityNominal (1.0)N/AExperience/TrainingNominal (1.0)N/A COMCOGPSFDiagnosis Multiplier(Base = 0.01)Action Multiplier (Base = 0.001)

3ENCLOSURE - 2ProceduresDiagnostic/symptom oriented (0.5)N/AErgonomics/HMINominal (1.0)N/AFitness for DutyNominal (1.0)N/AWork ProcessesNominal (1.0)N/ASUBTOTAL1.0E-4N/ATOTAL1E-4HPSIFLOWFor top event HPSIFLOW (representing operator failure to increasehigh-pressure safety injection flow on a loss of reactor coolant system inventory),the licensee's estimate of the HEP was 1.5E-5. For this HEP, the NRC staffagreed with the licensee's PSF level evaluation for Complexity in the Diagnosiscomponent. In this context, the decision to add additional safety injection flowgiven that HPSI was already inservice was considered an "Obvious Diagnosis"level. But, the NRC disagreed with the licensee's PSF level forExperience/Training. In the NRC's view this was best represented as "Nominal"in both the Diagnosis and Action components of the HEP. The basis for thischange was the same as described previously for COMCOG.The NRC concluded the HEP estimate for HPSIFLOW was 3E-5, as shownbelow:HPSIFLOWPSFDiagnosis Multiplier(Base = 0.01)Action Multiplier (Base = 0.001)Available TimeExpansive Time (0.01)>=50x time required (0.01)StressHigh (2.0)High (2.0)ComplexityObvious Diagnosis (0.1)Nominal (1.0)Experience/TrainingNominal (1.0)Nominal (1.0)ProceduresDiagnostic/symptom oriented (0.5)Nominal (1.0)Ergonomics/HMINominal (1.0)Nominal (1.0)Fitness for DutyNominal (1.0)Nominal (1.0)Work ProcessesNominal (1.0)Nominal (1.0)SUBTOTAL1.0E-52.0E-5 HPSIFLOWPSFDiagnosis Multiplier(Base = 0.01)Action Multiplier (Base = 0.001)

4ENCLOSURE - 2TOTAL3E-5XSPRAYVALVEFor top event XSPRAYVALVE (representing operator failure to isolate the flowdiversion path), the licensee's estimate of the HEP was 4.1E-4. The NRCdisagreed with the licensee's PSF level for Complexity in the Diagnosiscomponent of the HEP. The NRC concluded the licensee's selected PSF level of"Obvious Diagnosis" was overly optimistic and was best represented as"Nominal."The NRC concluded the HEP estimate for XSPRAYVALVE was 5E-4, as shownbelow:XSPRAYVALVEPSFDiagnosis Multiplier(Base = 0.01)Action Multiplier (Base = 0.001)Available TimeExpansive Time (0.01)>=50x time required (0.01)StressHigh (2.0)High (2.0)ComplexityNominal (1.0)Nominal (1.0)Experience/TrainingNominal (1.0)Nominal (1.0)ProceduresDiagnostic/symptom oriented (0.5)Incomplete (20.0)Ergonomics/HMINominal (1.0)Nominal (1.0)Fitness for DutyNominal (1.0)Nominal (1.0)Work ProcessesNominal (1.0)Nominal (1.0)SUBTOTAL1.0E-44.0E-4TOTAL5E-4d.The NRC staff reviewed the licensee's method of performing a dependencyanalysis for the HEPs and disagreed with the licensee's dependency decisiontree evaluation for HPSIFLOW. The NRC determined that "Additional Cues"should not be credited. The licensee credited the shift technical advisorperiodically checking the reactor vessel level monitoring system (RVLMS) as anadditional cue; but, the NRC concluded that using the RVLMS may not allowsufficient time for action between indication of a lowering water level and coredamage. Therefore, the NRC's dependency decision tree evaluation resulted in"moderate" dependency rather than "low" dependency. This resulted in a 5ENCLOSURE - 2conditional HEP of 0.14 for HPSIFLOW following failure to isolate the shutdowncooling flow diversion rather than 0.05.Based on the above review, the NRC staff estimated that the Joint HEP for Category 3events was 1.51E-7 and that the Joint HEP for Category 4 events was 1.37E-4.The increase in core damage frequency for each category of initiating event of concerncan be expressed as:Category 3: 1.69E-2/year * 1.51E-7 = 2.55E-9/year Category 4: 1.06E-2/year * 1.37E-4 = 1.45E-6/yearTherefore, Category 3 events had negligible contribution and the increase in coredamage frequency due to internal events was estimated as that from Category 4 eventsalone.II.External Initiatorsa.The NRC staff reviewed the licensee's approach for evaluating the contribution torisk significance of the finding due to external initiating events and concluded thatonly fire events were potentially significant.b.The licensee referenced use of NUREG/CR-6850, "EPRI/NRC-RES Fire PRAMethodology for Nuclear Power Facilities," for estimating HEPs following fireinitiating events. With respect to the events involving a fire-induced loss ofcomponent cooling water, the licensee identified four fire scenarios of interest. Two of these involved fires in the auxiliary building and two involved fires in thecontrol room. For each of the fire scenarios, the licensee multiplied the joint HEPfor these events by a factor of 10. NUREG/CR-6850 recommends applying thisfactor to HEPs from an internal events PRA when a specific set of conditions aremet. The multiplication factor is to account for the effects of potential fire brigadeinteraction and other minor increased workload and/or distraction issues. TheNRC noted that using this factor for control room fires was a nonconservativeapplication. The NUREG states that the HEP should be set to 0.1 or multipliedby a factor of 10, whichever is greater, based upon the complexity. Control roomfires may result in control room evacuation or other significant additionalcomplications that may require additional detailed analysis, and would indicatesetting the HEP valve to 0.1 is correct. However, the frequency of loss of reactorcoolant pump seal cooling events resulting from a control room fire-induced lossof component cooling water was very low (3.93E-6/year). Even if the 0.1 valuewere used, then the change in risk would be less than 3.93E-7/year. Therefore,the NRC concluded that the contribution of these fire scenarios to the overall risksignificance would be very small. As a result, the NRC still agreed with thelicensee's overall conclusion that the significance contribution from loss ofcomponent cooling water events resulting from fire was still very small.c.With respect to the events involving a fire-induced loss of offsite power (LOOP),the licensee stated that use of the NUREG/CR-6850 multiplication factor of 10 6ENCLOSURE - 2was not required. The fire scenarios of concern that would cause a LOOP eventare all fires involving transformers outside the plant in the transformer yard. Thelicensee stated that because these initiating events occurred outside the plantand several hours would elapse before operators would initiate shutdowncooling, a factor of 10 increase in the HEP was not necessary. The NRC staffdisagreed with this conclusion and believed that a factor of 10 increase in theHEP was still appropriate. These fire scenarios would typically be major eventsinvolving significant amounts of smoke, possibly involving the response of offsitefire fighters, and a coincident LOOP with operators performing a plant cooldownand depressurization using emergency diesel generators. The NRC consideredthat this scenario was consistent, at a minimum, with the NUREG/CR-6850philosophy of increasing the HEP by a factor of 10 to account for a minorincrease in operator workload. Therefore, using the licensee's frequency of a fire-induced LOOP of 2.31E-3/year, and the NRC's revised joint HEP multipliedby a factor of 10 (1.37E-4 * 10 = 1.37E-3), the increase in core damagefrequency resulting from fire-induced LOOP scenarios was:2.31E-3/year * 1.37E-3 = 3.16E-6/yearIII.Final Significance DeterminationThe overall safety significance of a performance deficiency with respect to core damagefrequency is expressed as the summation of the increase in core damage frequencyfrom internal and external initiating events. Therefore, the total increase in core damagefrequency is estimated as: 1.45E-6/year (Internal Events) + 3.16E-6/year (ExternalEvents) = 4.61E-6/year. This result is of low-to-moderate safety significance (White).