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| {{#Wiki_filter:CATEGORY1~REGULATOR INFORMATION DISTRIBUTIO?PKYSTEM (RIDS)ACCESSION NBR:9802120308 DOC.DATE: | | {{#Wiki_filter:CATEGORY 1~REGULATOR INFORMATION DISTRIBUTIO?PKYSTEM (RIDS)ACCESSION NBR:9802120308 DOC.DATE: 98/01/29 NOTARIZED: |
| 98/01/29NOTARIZED: | | NO FACIL'50-315 Donald C.Cook Nuclear Power Plant, Unit 1, Indiana M 650-316'onald C.Cook Nuclear Power Plant, Unit 2, Indiana,M AUTH.NAME AUTHOR AFFILIATION FITZPATRICK,E.E Indiana Michigan Power Co.RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk) |
| NOFACIL'50-315 DonaldC.CookNuclearPowerPlant,Unit1,IndianaM650-316'onald C.CookNuclearPowerPlant,Unit2,Indiana,M AUTH.NAMEAUTHORAFFILIATION FITZPATRICK,E.E IndianaMichiganPowerCo.RECIP.NAME RECIPIENT AFFILIATION DocumentControlBranch(Document ControlDesk)
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| ==SUBJECT:== | | ==SUBJECT:== |
| Forwardsconfirmatory actionresporise validation inspforRAIreitem3,"36HourCooldown.
| | Forwards confirmatory action resporise validation insp for RAI re item 3,"36 Hour Cooldown.DOCKET 05000315 05000316 DISTRIBUTION CODE: IE36D COPIES RECEIVED:LTR i ENCL J STEE-.TITLE: Immediate/Confirmatory Action Ltr (50 Dkt-Other Than Emergency Prepar NOTES: RECIPIENT ID CODE/NAME PD3-3 PD COPIES RECIPIENT LTTR ENCL ID CODE/NAME' 1 T HICKMAN<J COPIES LTTR ENCL 1 1 E G 0 INTERNA: FILE CE T EIB EXTERNAL: NOAC 01 1 1 1 1 1 1 NRR/DRPM/PECB NRC PDR 1 1 R D E N NOTE TO ALL"RIDS" RECIPIENTS: |
| DOCKET0500031505000316DISTRIBUTION CODE:IE36DCOPIESRECEIVED:LTR iENCLJSTEE-.TITLE:Immediate/Confirmatory ActionLtr(50Dkt-Other ThanEmergency PreparNOTES:RECIPIENT IDCODE/NAME PD3-3PDCOPIESRECIPIENT LTTRENCLIDCODE/NAME' 1THICKMAN<J COPIESLTTRENCL11EG0INTERNA:FILECETEIBEXTERNAL:
| | PLEASE HELP US TO REDUCE WASTE.TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD)ON EXTENSION 415-2083 TOTAL NUMBER OF COPIES REQUIRED: LTTR 7 ENCL 7 t*>L$t~ |
| NOAC01111111NRR/DRPM/PECB NRCPDR11RDENNOTETOALL"RIDS"RECIPIENTS:
| | Indiana Michigan Power Company 500 Circle Drive Buchanan, Ml 49107 1395 INSlANA N1CNESSl N PQWM January 29, 1998 AEP:NRC:1260G8 Docket Nos.: 50-315 50-316 U.S.Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop 0-Pl-17 Washington, D.C.20555-0001 Gentlemen: |
| PLEASEHELPUSTOREDUCEWASTE.TOHAVEYOURNAMEORORGANIZATION REMOVEDFROMDISTRIBUTION LISTSORREDUCETHENUMBEROFCOPIESRECEIVEDBYYOUORYOURORGANIZATION, CONTACTTHEDOCUMENTCONTROLDESK(DCD)ONEXTENSION 415-2083TOTALNUMBEROFCOPIESREQUIRED:
| | Donald C.Cook Nuclear Plant Units 1 and 2 CONFIRMATORY ACTION RESPONSE VALIDATION INSPECTION RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING ITEM 3,"36 HOUR COOLDOWN" During the time period January 9, 1998, through January 23, 1998, the NRC conducted a confirmatory action letter (CAL)followup inspection. |
| LTTR7ENCL7 t*>L$t~
| | During the exit meeting for the inspection, on January 26, 1998, we were requested to docket additional information regarding CAL item 3, involving the ability of the plant to be cooled down in 36 hours.Attachment 1 to this letter contains the information requested. |
| IndianaMichiganPowerCompany500CircleDriveBuchanan, Ml491071395INSlANAN1CNESSlNPQWMJanuary29,1998AEP:NRC:1260G8 DocketNos.:50-31550-316U.S.NuclearRegulatory Commission ATTN:DocumentControlDeskMailStop0-Pl-17Washington, D.C.20555-0001 Gentlemen:
| | Attachment 2 contains a flow chart depicting the information in attachment 1.Sincerely, EF~pm'.E.Fitzpatrick Vice President/vlb Attachments J.A.Abramson A.B.Beach MDEQ-DW fc RPD NRC Resident Inspector J.R.Sampson 9802i20308 980i29 PDR ADQCK 050003i5)PDR J j.M~~W+IIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIII |
| DonaldC.CookNuclearPlantUnits1and2CONFIRMATORY ACTIONRESPONSEVALIDATION INSPECTION RESPONSETOREQUESTFORADDITIONAL INFORMATION REGARDING ITEM3,"36HOURCOOLDOWN" DuringthetimeperiodJanuary9,1998,throughJanuary23,1998,theNRCconducted aconfirmatory actionletter(CAL)followupinspection.
| |
| Duringtheexitmeetingfortheinspection, onJanuary26,1998,wewererequested todocketadditional information regarding CALitem3,involving theabilityoftheplanttobecooleddownin36hours.Attachment 1tothislettercontainstheinformation requested.
| |
| Attachment 2containsaflowchartdepicting theinformation inattachment 1.Sincerely, EF~pm'.E.Fitzpatrick VicePresident | |
| /vlbAttachments J.A.AbramsonA.B.BeachMDEQ-DWfcRPDNRCResidentInspector J.R.Sampson9802i20308 980i29PDRADQCK050003i5)PDRJj.M~~W+IIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIII | |
| ' | | ' |
| ATTACHMENT 1TOAEP:NRC:1260G8 ADDITIONAL INFORMATION RELATEDTO36HOURCOOLDOHNCONFIRMATORY ACTIONLETTERITEM3 Attachment 1toAEP:NRC:1260G8 Page1Introduction Item3oftheconfirmatory actionletter(CAL)dealtspecifically withthe36hourcooldown. | | ATTACHMENT 1 TO AEP:NRC:1260G8 ADDITIONAL INFORMATION RELATED TO 36 HOUR COOLDOHN CONFIRMATORY ACTION LETTER ITEM 3 Attachment 1 to AEP:NRC:1260G8 Page 1 Introduction Item 3 of the confirmatory action letter (CAL)dealt specifically with the 36 hour cooldown.During the CAL validation inspection, several concerns were raised by the NRC relating to this issue.This letter is provided to assist with the characterization of'hose concerns.From August 4, 1997, through September 12, 1997, the NRC conducted an architect engineering (AE)inspection at Cook Nuclear Plant.During that inspection, several deficiencies were identified related to the component cooling water (CCW)system.Two such deficiencies were: the 36 hour cooldown analysis completed by Westinghouse for Cook Nuclear Plant contained errors, and no maximum cooling flow limits existed to protect the components cooled by CCW from high, potentially damaging flow rates.Subsequently, concurrent efforts were undertaken to correct the cooldown analysis and establish maximum flow limits.On August 29, 1997, condition report (CR)97-2378 was written to capture the AE inspector's concern that maximum flow limits for the CCW system did not exist.The condition report investigation was completed on December 20, 1997.As part of that investigation, technical reviews were completed that established an upper flow limit for each of the components cooled by the CCW system.A commitment was made in that CR response to incorporate the upper flow limits into the updated final safety analysis report (UFSAR).The due date for that commitment is April 1999.The April 1999 date corresponds to the next planned revision of the UFSAR.As part of the process of updating the UFSAR, a 10 CFR 50.59 evaluation is performed. |
| DuringtheCALvalidation inspection, severalconcernswereraisedbytheNRCrelatingtothisissue.Thisletterisprovidedtoassistwiththecharacterization of'hoseconcerns.
| | On September 19, 1997, the NRC issued a CAL to Cook Nuclear Plant.The CAL detailed eight specific items that require resolution prior to restart, one of which is the 36 hour cooldown issue.In our letter AEP:NRC:1260G3, dated December 2, 1997, we provided our CAL response to the NRC.From January 9, 1998, to January 23, 1998, the NRC conducted a CAL validation inspection to review and verify the CAL response.Discussion During the validation inspection, the NRC reviewed CAL Item 3, 36 hour cooldown.The review was two-fold.First, the NRC reviewed the actual plant cooldown calculation completed by Westinghouse, as well as the supporting documentation that had been compiled by both us and Westinghouse. |
| FromAugust4,1997,throughSeptember 12,1997,theNRCconducted anarchitect engineering (AE)inspection atCookNuclearPlant.Duringthatinspection, severaldeficiencies wereidentified relatedtothecomponent coolingwater(CCW)system.Twosuchdeficiencies were:the36hourcooldownanalysiscompleted byWestinghouse forCookNuclearPlantcontained errors,andnomaximumcoolingflowlimitsexistedtoprotectthecomponents cooledbyCCWfromhigh,potentially damagingflowrates.Subsequently, concurrent effortswereundertaken tocorrectthecooldownanalysisandestablish maximumflowlimits.OnAugust29,1997,condition report(CR)97-2378waswrittentocapturetheAEinspector's concernthatmaximumflowlimitsfortheCCWsystemdidnotexist.Thecondition reportinvestigation wascompleted onDecember20,1997.Aspartofthatinvestigation, technical reviewswerecompleted thatestablished anupperflowlimitforeachofthecomponents cooledbytheCCWsystem.Acommitment wasmadeinthatCRresponsetoincorporate theupperflowlimitsintotheupdatedfinalsafetyanalysisreport(UFSAR).Theduedateforthatcommitment isApril1999.TheApril1999datecorresponds tothenextplannedrevisionoftheUFSAR.AspartoftheprocessofupdatingtheUFSAR,a10CFR50.59evaluation isperformed.
| | That documentation included Westinghouse's safety evaluation check list (SECL)97-189, and our 10 CFR 50.59 evaluation completed for the 36 hour cooldown.It was determined that, during a 36 hour cooldown, the CCW supply temperature may exceed the previously analyzed 95~F and reach a maximum of 120~F.As part of design change 12-DCP-855, the necessary technical reviews and plant modifications were completed to allow the CCW Attachment 1 to AEP:NRC:1260GB Page 2 supply temperature to reach 1204 F.In this letter, the 10 CFR 50.59 evaluation that was completed for 12-DCP-855 is referred to as the 36 hour cooldown safety evaluation. |
| OnSeptember 19,1997,theNRCissuedaCALtoCookNuclearPlant.TheCALdetailedeightspecificitemsthatrequireresolution priortorestart,oneofwhichisthe36hourcooldownissue.InourletterAEP:NRC:1260G3, datedDecember2,1997,weprovidedourCALresponsetotheNRC.FromJanuary9,1998,toJanuary23,1998,theNRCconducted aCALvalidation inspection toreviewandverifytheCALresponse.
| | Upon reviewing the Westinghouse cooldown calculation, the NRC questioned two of the inputs used in the analysis.One question dealt with the total CCW flow.In the analysis, the total CCW flow rate was 8,000 gpm (i.e., 4.0E6 lb./hr.).This value was taken directly from table 9.5-3 of the UFSAR.The NRC expressed concern that there was no margin between the analysis input value and the UFSAR value.To address the concern, a 10 CFR 50.59 evaluation was expedited to revise the UFSAR to allow 9,000 gpm of total CCW flow through a single CCW heat exchanger. |
| Discussion Duringthevalidation inspection, theNRCreviewedCALItem3,36hourcooldown. | | As discussed above, the need for this 10 CFR 50.59 review had already been recognized and committed to as part of the investigation of CR 97-2378.It is important to recognize that the analysis input of 8,000 gpm was taken directly from the UFSAR and that the 9,000 gpm value is to be added to the UFSAR to show that margin exists in the cooldown analysis.Another question dealt with the CCW cooling flow to the residual heat removal (RHR)heat exchangers. |
| Thereviewwastwo-fold.
| | This value was 5,000 gpm in the cooldown analysis.The NRC inquired whether instrument uncertainty had been applied to the flow value.The NRC commented that at other plants it was not uncommon to see a flow uncertainty on the order of 1,000 gpm.To address the concern, we provided the NRC with information showing that the uncertainty in the flow indication loop was only on the order of p200 gpm.Additionally, the CCW flow to the RHR heat exchanger is on our critical parameters list.Per a previous commitment made to the NRC in our letter AEP:NRC:1260G3, dated December 2, 1997, instrument uncertainty will be incorporated into the operating procedures for each value on the critical parameters list.This commitment has a due date of December 31, 1998.Thus, the issue about uncertainty in the CCW cooling flow rate to the RHR heat exchangers is already scheduled to be addressed under a pxevious commitment. |
| First,theNRCreviewedtheactualplantcooldowncalculation completed byWestinghouse, aswellasthesupporting documentation thathadbeencompiledbybothusandWestinghouse. | | to the NRC.In addition to reviewing the actual cooldown calculation, the NRC also reviewed the supporting documentation. |
| Thatdocumentation includedWestinghouse's safetyevaluation checklist(SECL)97-189,andour10CFR50.59evaluation completed forthe36hourcooldown.
| | Included in that documentation was SECL 97-189 completed by Westinghouse to support operation of the CCW system at.the elevated supply temperature of 1204 F.In their SECL, Westinghouse stated the control valve.regulating the CCW flow to the letdown heat exchanger will go full open during a 36 hour cooldown.The NRC noted that Westinghouse assumed, with the valve full open, a total CCW flow of 1,000 gpm through the letdown heat exchanger. |
| Itwasdetermined that,duringa36hourcooldown, theCCWsupplytemperature mayexceedthepreviously analyzed95~Fandreachamaximumof120~F.Aspartofdesignchange12-DCP-855, thenecessary technical reviewsandplantmodifications werecompleted toallowtheCCW Attachment 1toAEP:NRC:1260GB Page2supplytemperature toreach1204F.Inthisletter,the10CFR50.59evaluation thatwascompleted for12-DCP-855 isreferredtoasthe36hourcooldownsafetyevaluation.
| | The inspector pointed out that table 9.5-2 of the UFSAR lists the design CCW flow to the letdown heat exchanger as 984 gpm and questioned why the apparent discrepancy was not called out in the safety review.The following explanation was provided.The actual design cooling flow rate for the letdown heat exchanger is 492,000 lb./hr.-a mass flow rate, not a volumetric flow rate.This is shown in table 9.2-3 of the UFSAR.The flow rate of 984 gpm, as shown in table 9.5-2 of the UFSAR, is the volumetric flow rate that corresponds to 492,000 lb./hr.at standard conditions. |
| Uponreviewing theWestinghouse cooldowncalculation, theNRCquestioned twooftheinputsusedintheanalysis.
| | The flow rate of 1,000 gpm, as shown in the SECL, is the volumetric flow rate that corresponds to 492,000 lb./hr.at the CCW return temperature expected during a 36 hour cooldown.Thus, the volumetric flow rates of 984 gpm and 1,000 gpm are both correct as they are both derived from the design mass flow rate specified for the letdown heat exchanger. |
| OnequestiondealtwiththetotalCCWflow.Intheanalysis, thetotalCCWflowratewas8,000gpm(i.e.,4.0E6lb./hr.).
| | Attachment 1 to AEP:NRC:1260G8 Page 3 As a follow up question to the one above, the NRC asked how much CCW flow is expected to pass through the letdown heat exchanger with the control valve full open.A preliminary calculation determined with the control valve full open, approximately 1,400 gpm would pass through the letdown heat exchanger. |
| Thisvaluewastakendirectlyfromtable9.5-3oftheUFSAR.TheNRCexpressed concernthattherewasnomarginbetweentheanalysisinputvalueandtheUFSARvalue.Toaddresstheconcern,a10CFR50.59evaluation wasexpedited torevisetheUFSARtoallow9,000gpmoftotalCCWflowthroughasingleCCWheatexchanger.
| | The NRC further questioned why the flow rate, which is greater than the value listed in the UFSAR, was not evaluated in the 36 hour cooldown safety evaluation. |
| Asdiscussed above,theneedforthis10CFR50.59reviewhadalreadybeenrecognized andcommitted toaspartoftheinvestigation ofCR97-2378.Itisimportant torecognize thattheanalysisinputof8,000gpmwastakendirectlyfromtheUFSARandthatthe9,000gpmvalueistobeaddedtotheUFSARtoshowthatmarginexistsinthecooldownanalysis.
| | To address the immediate concern, a 10 CFR 50.59 evaluation was expedited to evaluate twice the design cooling flow rate through the letdown heat exchanger (approximately 2,000 gpm).As discussed above, the need for this 10 CFR 50.59 review had already been recognized and committed to as part of the investigation of CR 97-2378.Although the letdown heat exchanger cooling flow rate of 1,400 gpm was not specifically addressed in the 36 hour cooldown safety review, a bounding analysis had been completed and a commitment to revise the UFSAR already existed.It is important to recognize that, although the increased CCW flow rate to the letdown heat exchanger is a consequence of the higher CCW supply temperature, it is not an input in the cooldown analysis. |
| AnotherquestiondealtwiththeCCWcoolingflowtotheresidualheatremoval(RHR)heatexchangers.
| | ATTACHMENT 2 TO AEP:NRC:1280G8 FLOW CHART DEPICTING INFORMATION PROVIDED IN ATTACHMENT 1 |
| Thisvaluewas5,000gpminthecooldownanalysis.
| | hccacbsienc 2 so AEP/NRCI126008 |
| TheNRCinquiredwhetherinstrument uncertainty hadbeenappliedtotheflowvalue.TheNRCcommented thatatotherplantsitwasnotuncommontoseeaflowuncertainty ontheorderof1,000gpm.Toaddresstheconcern,weprovidedtheNRCwithinformation showingthattheuncertainty intheflowindication loopwasonlyontheorderofp200gpm.Additionally, theCCWflowtotheRHRheatexchanger isonourcriticalparameters list.Perapreviouscommitment madetotheNRCinourletterAEP:NRC:1260G3, datedDecember2,1997,instrument uncertainty willbeincorporated intotheoperating procedures foreachvalueonthecriticalparameters list.Thiscommitment hasaduedateofDecember31,1998.Thus,theissueaboutuncertainty intheCCWcoolingflowratetotheRHRheatexchangers isalreadyscheduled tobeaddressed underapxeviouscommitment.
| | .QUESTION: The cakuhtion assumes 5000 g pm of CCW coohng flow to the RHR heat exchanger. |
| totheNRC.Inadditiontoreviewing theactualcooldowncalculation, theNRCalsoreviewedthesupporting documentation.
| | Was hstrument uncertainty appfied?ANSWER: The CCW ihw to the RHR heat exchangers is on the critical parameters gsL All procedures which reference the 5000 gpm used in the analysts wfll be revised to account for instrument uncertainty. |
| Includedinthatdocumentation wasSECL97-189completed byWestinghouse tosupportoperation oftheCCWsystemat.theelevatedsupplytemperature of1204F.IntheirSECL,Westinghouse statedthecontrolvalve.regulating theCCWflowtotheletdownheatexchanger willgofullopenduringa36hourcooldown.
| | This ls a gbbal comnitment to the NRC which is due by the end of 1998.r NRC/AE,INSPECTION August-September 1997 CAL ITEM¹3 36 Hour Cooldown I I I NRC Inspector Reviewed Westinghouse Plant Cooldown Analysis NRC Inspector Reviewed Westinghouse SECL and Our Safety Evaluation QUESllON;The cooldown analysis assumes a total COW lhwof 8000 gpm.The FSAR aflaws a total CCW flaw of 809 gpm through the COW heat exchanger. |
| TheNRCnotedthatWestinghouse assumed,withthevalvefullopen,atotalCCWflowof1,000gpmthroughtheletdownheatexchanger.
| | Wiry is there no margin between the osculation hpul and the phot design basis?QUESTION: Westinghouse stated that the COW control valve for the letdown heat exchanger wgl go fufl open.Using a CCW flaw rate of 1000 gpm to the letdown heat exchanger, they cahuhted the magnum letdown temperature. |
| Theinspector pointedoutthattable9.5-2oftheUFSARliststhedesignCCWflowtotheletdownheatexchanger as984gpmandquestioned whytheapparentdiscrepancy wasnotcalledoutinthesafetyreview.Thefollowing explanation wasprovided.
| | The FSAR states that the CCW Ihw to the letdown heat exchanger Is 984 gpss Why wasnl this addressed in lhe safety evahation? |
| Theactualdesigncoolingflowratefortheletdownheatexchanger is492,000lb./hr.-amassflowrate,notavolumetric flowrate.Thisisshownintable9.2-3oftheUFSAR.Theflowrateof984gpm,asshownintable9.5-2oftheUFSAR,isthevolumetric flowratethatcorresponds to492,000lb./hr.atstandardconditions.
| | I I I I I I I I I I I I I I I RESOLUTION: |
| Theflowrateof1,000gpm,asshownintheSECL,isthevolumetric flowratethatcorresponds to492,000lb./hr.attheCCWreturntemperature expectedduringa36hourcooldown.
| | The 50.59 review to aflaw 9000 g pm through the CCW heat exchanger was expedited.(Comphted 1/23/98)RESOLUTION: |
| Thus,thevolumetric flowratesof984gpmand1,000gpmarebothcorrectastheyarebothderivedfromthedesignmassflowratespecified fortheletdownheatexchanger. | | As shawn h the FSAR, the design flaw for the letdown heat exchanger is 492,000 ib/hr.This is equfiratent to 984 gprn at 76'F and 1000 gpm at 120'F.Westinghouse correcfly used 492,000 Ib/hr to cahutate Ihe temperature. |
| Attachment 1toAEP:NRC:1260G8 Page3Asafollowupquestiontotheoneabove,theNRCaskedhowmuchCCWflowisexpectedtopassthroughtheletdownheatexchanger withthecontrolvalvefullopen.Apreliminary calculation determined withthecontrolvalvefullopen,approximately 1,400gpmwouldpassthroughtheletdownheatexchanger. | | '0 Subsequent to the hspection, a technkst review was completed to determine a maximum aflowabh caofing ihw rate for each component served by CCW.(CR 97-2378)The maxhxrm aflawabfe ihw through the CCW heat exchanger is 9000 gpm.Completed 12/2N97, The maximum aflawable llaw through cooflng fhw through the letdown heat exchanger Is twice design (l.e.984,000 lb/br).ted 12/20/97.A part F convnitment was made In corxfitian report 97-2378 to hcorporate the maximum flaw fimits hto the FSAIL Commitment made: 11/23I97 Commitment due: Apnl 1999 QUESTION: If the control valve does go fufl open, what wifl be the CCW fhw rate to the letdown heat exchanger? |
| TheNRCfurtherquestioned whytheflowrate,whichisgreaterthanthevaluelistedintheUFSAR,wasnotevaluated inthe36hourcooldownsafetyevaluation.
| | ANSWER With the control valve full open, the CCW flow wg be approxhnatety 1400 gpm..This Is greater than the design fhw rate shown h the FSAIL RESOLUTION: |
| Toaddresstheimmediate concern,a10CFR50.59evaluation wasexpedited toevaluatetwicethedesigncoolingflowratethroughtheletdownheatexchanger (approximately 2,000gpm).Asdiscussed above,theneedforthis10CFR50.59reviewhadalreadybeenrecognized andcommitted toaspartoftheinvestigation ofCR97-2378.Althoughtheletdownheatexchanger coolingflowrateof1,400gpmwasnotspecifically addressed inthe36hourcooldownsafetyreview,aboundinganalysishadbeencompleted andacommitment torevisetheUFSARalreadyexisted.Itisimportant torecognize that,althoughtheincreased CCWflowratetotheletdownheatexchanger isaconsequence ofthehigherCCWsupplytemperature, itisnotaninputinthecooldownanalysis.
| | The 50.59 review to aihw twice design flaw rate through the letdown heat exchanger was expefited.(Completed 1/23lg}} |
| ATTACHMENT 2TOAEP:NRC:1280G8 FLOWCHARTDEPICTING INFORMATION PROVIDEDINATTACHMENT 1 | |
| hccacbsienc 2soAEP/NRCI126008 | |
| .QUESTION: | |
| Thecakuhtion assumes5000gpmofCCWcoohngflowtotheRHRheatexchanger.
| |
| Washstrument uncertainty appfied?ANSWER:TheCCWihwtotheRHRheatexchangers isonthecriticalparameters gsLAllprocedures whichreference the5000gpmusedintheanalystswfllberevisedtoaccountforinstrument uncertainty.
| |
| Thislsagbbalcomnitment totheNRCwhichisduebytheendof1998.rNRC/AE,INSPECTION August-September 1997CALITEM¹336HourCooldownIIINRCInspector ReviewedWestinghouse PlantCooldownAnalysisNRCInspector ReviewedWestinghouse SECLandOurSafetyEvaluation QUESllON; ThecooldownanalysisassumesatotalCOWlhwof8000gpm.TheFSARaflawsatotalCCWflawof809gpmthroughtheCOWheatexchanger.
| |
| Wiryistherenomarginbetweentheosculation hpulandthephotdesignbasis?QUESTION:
| |
| Westinghouse statedthattheCOWcontrolvalvefortheletdownheatexchanger wglgofuflopen.UsingaCCWflawrateof1000gpmtotheletdownheatexchanger, theycahuhtedthemagnumletdowntemperature. | |
| TheFSARstatesthattheCCWIhwtotheletdownheatexchanger Is984gpssWhywasnlthisaddressed inlhesafetyevahation?
| |
| IIIIIIIIIIIIIIIRESOLUTION:
| |
| The50.59reviewtoaflaw9000gpmthroughtheCCWheatexchanger wasexpedited.
| |
| (Comphted 1/23/98)RESOLUTION: | |
| AsshawnhtheFSAR,thedesignflawfortheletdownheatexchanger is492,000ib/hr.Thisisequfiratent to984gprnat76'Fand1000gpmat120'F.Westinghouse correcfly used492,000Ib/hrtocahutateIhetemperature.
| |
| '0Subsequent tothehspection, atechnkstreviewwascompleted todetermine amaximumaflowabhcaofingihwrateforeachcomponent servedbyCCW.(CR97-2378)Themaxhxrmaflawabfe ihwthroughtheCCWheatexchanger is9000gpm.Completed 12/2N97,Themaximumaflawable llawthroughcooflngfhwthroughtheletdownheatexchanger Istwicedesign(l.e.984,000lb/br).ted12/20/97. | |
| ApartFconvnitment wasmadeIncorxfitian report97-2378tohcorporate themaximumflawfimitshtotheFSAILCommitment made:11/23I97Commitment due:Apnl1999QUESTION:
| |
| Ifthecontrolvalvedoesgofuflopen,whatwiflbetheCCWfhwratetotheletdownheatexchanger?
| |
| ANSWERWiththecontrolvalvefullopen,theCCWflowwgbeapproxhnatety 1400gpm..This IsgreaterthanthedesignfhwrateshownhtheFSAILRESOLUTION:
| |
| The50.59reviewtoaihwtwicedesignflawratethroughtheletdownheatexchanger wasexpefited.
| |
| (Completed 1/23lg}} | |
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ML17326A0981999-08-23023 August 1999
Forwards fitness-for-duty Program Performance Data for Period of 990101-0630 for DC Cook Nuclear Plants,Units 1 & 2,per 10CFR26.71(d)
ML17326A0891999-08-16016 August 1999
Forwards LER 99-019-00,re Victoreen Containment High Range Monitors Not Beign Environmentally Qualified to Withstand post-LOCA Conditions.Commitments Made by Util Are Listed
ML17326A0811999-08-10010 August 1999
Notifies NRC of Changes in Commitments Made in Response to GL 98-01,supplement 1, Yr 2000 Readiness of Computer Sys Ar Npps, Dtd 990623
ML17326A0821999-08-0606 August 1999
Informs That Util Is Submitting Encl Scope & Objectives for 991026 DC Cook Nuclear Plant Emergency Plan Exercise to G Shear of NRC Plant Support Branch.Exercise Will Include Full State & County Participation
ML17326A1451999-08-0404 August 1999
Requests Withholding of WCAP-15246, Control Rod Insertion Following Cold Leg Lbloca.
ML17326A0751999-08-0404 August 1999
Forwards LER 98-029-01, Fuel Handling Area Ventilation Sys Inoperable Due to Original Design Deficiency. Supplemental Rept Represents Extensive Rev to Original LER & Replaces Rept in Entirely.Commitment Listed
ML17326A0721999-07-29029 July 1999
Forwards LER 99-018-00 Re Refueling Water Storage Tank Suction Motor Operated Valves Inoperable,Due to Inadequate Design.Listed Commitments Were Identified in LER
ML17326A0711999-07-27027 July 1999
Responds to 980123 RAI Re NRC GL 87-02, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors,Unresolved Safety Issue (USI) A-46.
ML17326A0601999-07-22022 July 1999
Forwards UFSAR, IAW 10CFR50.71(e) & Rept of Changes,Tests & Experiments as Required by 10CFR50.59(b)(2) for DC Cook Nuclear Plant,Units 1 & 2.Without UFSAR
ML17326A0631999-07-22022 July 1999
Forwards LER 98-014-03, Response to High-High Containment Pressure Procedure Not Consistent with Analysis of Record. Revised Info Marked by Sidebars in Right Hand Margin. Commitments Made by Util,Listed
ML17326A0311999-07-0101 July 1999
Forwards LER 99-004-01 Re Failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed. Commitments Made by Util Are Listed
ML20196K5961999-06-30030 June 1999
Ltr Contract:Task Order 40, DC Cook Extended Sys Regulatory Review Oversight Insp, Under Contract NRC-03-98-021
ML17326A0281999-06-28028 June 1999
Provides Response to 981116 & 960228 RAIs Re GL 92-01. Revised Pressurized Thermal Shock Evaluation Based on New Weld Chemistry Info & Copy of W Rept WCAP-15074, Evaluation of 1P3571 Weld Metal from Surveillance Programs... Encl
ML17326A0241999-06-23023 June 1999
Responds to GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Nuclear Power Plants. Y2K Readiness Disclosure for Plant & List of Commitments Encl
ML17326A0121999-06-18018 June 1999
Forwards LER 99-014-00 Re Requirement of TS 4.0.5 Not Met for Boron Injection Tank Bolting.Commitments Identified in Submittal Listed
ML17326A0111999-06-11011 June 1999
Provides Response to NRC RAI Re GL 97-01, Degradation of Crdm/Cedm Nozzle & Other Vessel Closure Head Penetrations.
ML17325B6281999-06-0101 June 1999
Forwards LER 99-S03-00,re Nonconforming Vital Area Barriers.Commitments Made by Util Are Listed
ML17325B6401999-06-0101 June 1999
Forwards LER 99-013-00 Re Safety Injection & Centrifugal Charging Throttle Valve Cavitation During LOCA Could Lead to ECCS Pump Failure.Listed Commitments Identified in Submittal
ML17325B6331999-05-28028 May 1999
Forwards LER 99-S02-00,re Vulnerability in Safeguard Sys That Could Allow Unauthorized or Undetected Access to Protected Area.Commitments Made by Util Are Listed
ML17265A8201999-05-24024 May 1999
Forwards LER 98-037-01,representing Extensive Rev to Original LER & Replacing Rept in Entirety.Listed Commitments Identified in Submittal
ML20207A9201999-05-21021 May 1999
Ack Receipt of 990319 Response to Notice of Violation & Proposed Imposition of Civil Penalty .On 981124, Licensee Remitted Check for Payment of Civil Penalties. Licensee Requests for Extension for Response,Granted
ML17325B6111999-05-21021 May 1999
Forwards Annual Radioactive Effluent Release Rept for 980101-1231 for DC Cook Nuclear Plant,Units 1 & 2. Transmittal of Submittal Was Delayed Due to Administrative Error in Regulatory Affairs Dept
ML17325B6031999-05-21021 May 1999
Provides Response to NRC GL 98-04, Potential for Degradation of ECCS & Containment Spray Sys After LOCA Because of Const & Protective Coating Deficiencies & Foreign Matl in Containment.
ML17325B5971999-05-20020 May 1999
Forwards LER 99-012-00,re Auxiliary Building ESF Ventilation Sys Not Being Capable of Maintaining ESF Room Temps post-accident.Commitment,listed
ML17335A5281999-05-12012 May 1999
Forwards DC Cook Nuclear Plant Fitness for Duty Program Performance Dtd for six-month Period of 980701-1231,IAW 10CFR26.71(d).Info Was Delayed Due to Administrative Error in Regulatory Affairs Dept
ML17335A5271999-05-11011 May 1999
Forwards Details Re Sources & Levels of Insurance Maintained for DC Cook,Units 1 & 2,as of 990401,per 10CFR50.54(w)(3). Info Was Delayed Beyond Required Date Due to Internal Oversight
ML17325B5841999-05-10010 May 1999
Forwards LER 99-002-00 Re TS 4.0.5 Requirements Not Being Met Due to Improperly Performed Test.Commitments Identified in Ler,Listed
ML17325B5871999-05-0707 May 1999
Forwards Current Revs of Expanded Sys Readiness Review (Essr) Implementing Procedures,For Info Purposes to Support Current NRC Insps.Current Esrr Schedule Provided for Info Purposes,Reflecting Revised Target Dates
ML17325B5791999-05-0404 May 1999
Forwards LER 99-011-00,concerning Air Sys for EDG Not Supporting Long Term Operability.Commitments Made by Util Listed
ML17325B5821999-05-0404 May 1999
Provides Addl Background,Description & Clarification of Previous & Revised Commitments Re UFSAR Revalidation Effort. Commitment Change Involved Alignment of UFSAR Revalidation Program Methodology to Strategy Contained in Current Plan
ML17325B5741999-05-0303 May 1999
Forwards LER 99-010-00 Re RCS Leak Detection Sys Sensitivity Not in Accoradnce with Design Requirements.Listed Commitments Identified in Submittal
ML17325B5631999-04-22022 April 1999
Forwards Results of Independent Chemical Evaluations Performed from Sept 1997 Through Feb 1999,re Resolution of Issues Related to License Amend 227
ML17325B5561999-04-16016 April 1999
Forwards LER 99-006-00, Fuel Crane Loads Lifted Over SFP Could Impact Energies Greater than TS Limits, IAW 10CFR50.73.Submittal Was Delayed to Allow for Resolution of Questions.Commitment Made by Licensee,Listed
ML20205P0591999-04-14014 April 1999
Ninth Partial Response to FOIA Request for Documents.App Records Already Available in Pdr.Records in App T Encl & Being Made Available in Pdr.App U Records Being Released in Part (Ref FOIA Exemption 7).App V Records Withheld Entirely
ML17325B5451999-04-12012 April 1999
Forwards LER 99-009-00 Re as-found RHR Safety Relief Valve Lift Setpoint Greater than TS Limit.Commitments Identified in Submittal Listed
ML17325B5301999-04-0707 April 1999
Forwards LER 99-S01-01, Vulnerability in Locking Mechanism of Four Vital Area Gates, Per 10CFR50.73.Commitments Made by Util,Listed
ML17325B5241999-04-0505 April 1999
Forwards Revs 0 & 1 to Cook Nuclear Plant Restart Plan, Dtd 980307 & 0407.Rev 5 Is Current Cook Nuclear Plant Restart & Supercedes Previous Revs in All Respects
ML17325B5121999-04-0101 April 1999
Forwards LER 99-007-00, Calculations Show That Divider Barrier Between Upper & Lower Containment Vols May Be Overstressed. Commitments Made by Util Are Listed
1999-09-30
[Table view]
Category:INCOMING CORRESPONDENCE
MONTHYEARML17335A5511999-10-0707 October 1999
Forwards LER 99-023-00, Inadequate TS Surveillance Testing of ESW Pump ESF Response Time. Commitments Identified in LER Listed
ML20217D9361999-09-30030 September 1999
FOIA Request for Document Re Section 9.7 of SE by Directorate of Licensing,Us Ae Commission in Matter of Indiana & Michigan Electric Co & Indiana & Michigan Power Co,Dc Cook Nuclear Plan,Units 1 & 2
ML17326A1541999-09-20020 September 1999
Provides Notification of Change in Senior Licensed Operator Status.Operating Licenses for CR Smith,License SOP-30159-4 & Tw Welch,License SOP-30654-2 Are No Longer Required & Should Be Withdrawn
ML17326A1261999-09-17017 September 1999
Forwards LER 99-022-00 Re Electrical Bus Degraded Voltage Setpoints Too Low for Safety Related Loads.Listed Commitment Identified in Submittal
ML17326A1441999-09-17017 September 1999
Submits Trace on Second Shipment of Two Plant,Unit 2 Steam Generators.Info Re Shipment Submitted
ML17326A1531999-09-16016 September 1999
Submits Info Pertaining to Plant Proposed Operator Licensing Exam Requirements Through Yr 2003.NRC Form 536, Operator Licensing Exam Data, Which Provides Required Info Encl
ML17326A1101999-08-27027 August 1999
Forwards LER 99-021-00, GL 96-01 Test Requirements Not Met in Surveillance Tests. List of Commitments Identified in LER Provided
ML17326A0991999-08-26026 August 1999
Forwards LER 99-020-00,re EDGs Being Declared Inoperable. Commitments Made by Util Are Listed
ML17326A1221999-08-23023 August 1999
Forwards Revised Page 2 to 1998 Annual Environ Operating Rept, for DC Cook Nuclear Plant,Correcting Omission to App I
ML17326A0981999-08-23023 August 1999
Forwards fitness-for-duty Program Performance Data for Period of 990101-0630 for DC Cook Nuclear Plants,Units 1 & 2,per 10CFR26.71(d)
ML17326A0891999-08-16016 August 1999
Forwards LER 99-019-00,re Victoreen Containment High Range Monitors Not Beign Environmentally Qualified to Withstand post-LOCA Conditions.Commitments Made by Util Are Listed
ML17326A0811999-08-10010 August 1999
Notifies NRC of Changes in Commitments Made in Response to GL 98-01,supplement 1, Yr 2000 Readiness of Computer Sys Ar Npps, Dtd 990623
ML17326A0821999-08-0606 August 1999
Informs That Util Is Submitting Encl Scope & Objectives for 991026 DC Cook Nuclear Plant Emergency Plan Exercise to G Shear of NRC Plant Support Branch.Exercise Will Include Full State & County Participation
ML17326A1451999-08-0404 August 1999
Requests Withholding of WCAP-15246, Control Rod Insertion Following Cold Leg Lbloca.
ML17326A0751999-08-0404 August 1999
Forwards LER 98-029-01, Fuel Handling Area Ventilation Sys Inoperable Due to Original Design Deficiency. Supplemental Rept Represents Extensive Rev to Original LER & Replaces Rept in Entirely.Commitment Listed
ML17326A0721999-07-29029 July 1999
Forwards LER 99-018-00 Re Refueling Water Storage Tank Suction Motor Operated Valves Inoperable,Due to Inadequate Design.Listed Commitments Were Identified in LER
ML17326A0711999-07-27027 July 1999
Responds to 980123 RAI Re NRC GL 87-02, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors,Unresolved Safety Issue (USI) A-46.
ML17326A0601999-07-22022 July 1999
Forwards UFSAR, IAW 10CFR50.71(e) & Rept of Changes,Tests & Experiments as Required by 10CFR50.59(b)(2) for DC Cook Nuclear Plant,Units 1 & 2.Without UFSAR
ML17326A0631999-07-22022 July 1999
Forwards LER 98-014-03, Response to High-High Containment Pressure Procedure Not Consistent with Analysis of Record. Revised Info Marked by Sidebars in Right Hand Margin. Commitments Made by Util,Listed
ML17326A0311999-07-0101 July 1999
Forwards LER 99-004-01 Re Failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed. Commitments Made by Util Are Listed
ML17326A0281999-06-28028 June 1999
Provides Response to 981116 & 960228 RAIs Re GL 92-01. Revised Pressurized Thermal Shock Evaluation Based on New Weld Chemistry Info & Copy of W Rept WCAP-15074, Evaluation of 1P3571 Weld Metal from Surveillance Programs... Encl
ML17326A0241999-06-23023 June 1999
Responds to GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Nuclear Power Plants. Y2K Readiness Disclosure for Plant & List of Commitments Encl
ML17326A0121999-06-18018 June 1999
Forwards LER 99-014-00 Re Requirement of TS 4.0.5 Not Met for Boron Injection Tank Bolting.Commitments Identified in Submittal Listed
ML17326A0111999-06-11011 June 1999
Provides Response to NRC RAI Re GL 97-01, Degradation of Crdm/Cedm Nozzle & Other Vessel Closure Head Penetrations.
ML17325B6401999-06-0101 June 1999
Forwards LER 99-013-00 Re Safety Injection & Centrifugal Charging Throttle Valve Cavitation During LOCA Could Lead to ECCS Pump Failure.Listed Commitments Identified in Submittal
ML17325B6281999-06-0101 June 1999
Forwards LER 99-S03-00,re Nonconforming Vital Area Barriers.Commitments Made by Util Are Listed
ML17325B6331999-05-28028 May 1999
Forwards LER 99-S02-00,re Vulnerability in Safeguard Sys That Could Allow Unauthorized or Undetected Access to Protected Area.Commitments Made by Util Are Listed
ML17265A8201999-05-24024 May 1999
Forwards LER 98-037-01,representing Extensive Rev to Original LER & Replacing Rept in Entirety.Listed Commitments Identified in Submittal
ML17325B6111999-05-21021 May 1999
Forwards Annual Radioactive Effluent Release Rept for 980101-1231 for DC Cook Nuclear Plant,Units 1 & 2. Transmittal of Submittal Was Delayed Due to Administrative Error in Regulatory Affairs Dept
ML17325B6031999-05-21021 May 1999
Provides Response to NRC GL 98-04, Potential for Degradation of ECCS & Containment Spray Sys After LOCA Because of Const & Protective Coating Deficiencies & Foreign Matl in Containment.
ML17325B5971999-05-20020 May 1999
Forwards LER 99-012-00,re Auxiliary Building ESF Ventilation Sys Not Being Capable of Maintaining ESF Room Temps post-accident.Commitment,listed
ML17335A5281999-05-12012 May 1999
Forwards DC Cook Nuclear Plant Fitness for Duty Program Performance Dtd for six-month Period of 980701-1231,IAW 10CFR26.71(d).Info Was Delayed Due to Administrative Error in Regulatory Affairs Dept
ML17335A5271999-05-11011 May 1999
Forwards Details Re Sources & Levels of Insurance Maintained for DC Cook,Units 1 & 2,as of 990401,per 10CFR50.54(w)(3). Info Was Delayed Beyond Required Date Due to Internal Oversight
ML17325B5841999-05-10010 May 1999
Forwards LER 99-002-00 Re TS 4.0.5 Requirements Not Being Met Due to Improperly Performed Test.Commitments Identified in Ler,Listed
ML17325B5871999-05-0707 May 1999
Forwards Current Revs of Expanded Sys Readiness Review (Essr) Implementing Procedures,For Info Purposes to Support Current NRC Insps.Current Esrr Schedule Provided for Info Purposes,Reflecting Revised Target Dates
ML17325B5821999-05-0404 May 1999
Provides Addl Background,Description & Clarification of Previous & Revised Commitments Re UFSAR Revalidation Effort. Commitment Change Involved Alignment of UFSAR Revalidation Program Methodology to Strategy Contained in Current Plan
ML17325B5791999-05-0404 May 1999
Forwards LER 99-011-00,concerning Air Sys for EDG Not Supporting Long Term Operability.Commitments Made by Util Listed
ML17325B5741999-05-0303 May 1999
Forwards LER 99-010-00 Re RCS Leak Detection Sys Sensitivity Not in Accoradnce with Design Requirements.Listed Commitments Identified in Submittal
ML17325B5631999-04-22022 April 1999
Forwards Results of Independent Chemical Evaluations Performed from Sept 1997 Through Feb 1999,re Resolution of Issues Related to License Amend 227
ML17325B5561999-04-16016 April 1999
Forwards LER 99-006-00, Fuel Crane Loads Lifted Over SFP Could Impact Energies Greater than TS Limits, IAW 10CFR50.73.Submittal Was Delayed to Allow for Resolution of Questions.Commitment Made by Licensee,Listed
ML17325B5451999-04-12012 April 1999
Forwards LER 99-009-00 Re as-found RHR Safety Relief Valve Lift Setpoint Greater than TS Limit.Commitments Identified in Submittal Listed
ML17325B5301999-04-0707 April 1999
Forwards LER 99-S01-01, Vulnerability in Locking Mechanism of Four Vital Area Gates, Per 10CFR50.73.Commitments Made by Util,Listed
ML17325B5241999-04-0505 April 1999
Forwards Revs 0 & 1 to Cook Nuclear Plant Restart Plan, Dtd 980307 & 0407.Rev 5 Is Current Cook Nuclear Plant Restart & Supercedes Previous Revs in All Respects
ML17325B5121999-04-0101 April 1999
Forwards LER 99-007-00, Calculations Show That Divider Barrier Between Upper & Lower Containment Vols May Be Overstressed. Commitments Made by Util Are Listed
ML17325B5141999-03-30030 March 1999
Forwards Rept on Status of Decommissioning Funding.Attached Rept Includes Amount of Decommissioning Funds Estimated to Be Required Pursuant to 10CFR50.75(b) & (C)
ML17325B5191999-03-29029 March 1999
Forwards LER 99-001-00,re Degraded Component Cw Flow to Containment Main Steam Line Penetrations.Commitment, Listed
ML20204F6401999-03-19019 March 1999
Responds to NRC 981013 NOV & Proposed Imposition of Civil Penalty.Violations Cited in Subject NOV Were Initially Identified in Referenced Five Insp Repts.Corrective Actions: Ice Condensers Have Been Completely Thawed of Any Blockage
ML17325B4751999-03-18018 March 1999
Forwards LER 99-004-00,re Failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed. Commitment Made by Util,Listed
ML17325B4721999-03-18018 March 1999
Forwards LER 99-005-00,re Reactor Trip Breaker Manual Actuations During Rod Drop Testing Not Previously Reported. Listed Commitments Identified in Submittal
ML17325B4641999-03-17017 March 1999
Withdraws Response to Issue 1 of NRC Cal,Dtd 970919. Comprehensive Design Review Effort in Progress to Validate Resolution of Issue for Future Operation
1999-09-30
[Table view] |
Text
CATEGORY 1~REGULATOR INFORMATION DISTRIBUTIO?PKYSTEM (RIDS)ACCESSION NBR:9802120308 DOC.DATE: 98/01/29 NOTARIZED:
NO FACIL'50-315 Donald C.Cook Nuclear Power Plant, Unit 1, Indiana M 650-316'onald C.Cook Nuclear Power Plant, Unit 2, Indiana,M AUTH.NAME AUTHOR AFFILIATION FITZPATRICK,E.E Indiana Michigan Power Co.RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Forwards confirmatory action resporise validation insp for RAI re item 3,"36 Hour Cooldown.DOCKET 05000315 05000316 DISTRIBUTION CODE: IE36D COPIES RECEIVED:LTR i ENCL J STEE-.TITLE: Immediate/Confirmatory Action Ltr (50 Dkt-Other Than Emergency Prepar NOTES: RECIPIENT ID CODE/NAME PD3-3 PD COPIES RECIPIENT LTTR ENCL ID CODE/NAME' 1 T HICKMAN<J COPIES LTTR ENCL 1 1 E G 0 INTERNA: FILE CE T EIB EXTERNAL: NOAC 01 1 1 1 1 1 1 NRR/DRPM/PECB NRC PDR 1 1 R D E N NOTE TO ALL"RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE.TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD)ON EXTENSION 415-2083 TOTAL NUMBER OF COPIES REQUIRED: LTTR 7 ENCL 7 t*>L$t~
Indiana Michigan Power Company 500 Circle Drive Buchanan, Ml 49107 1395 INSlANA N1CNESSl N PQWM January 29, 1998 AEP:NRC:1260G8 Docket Nos.: 50-315 50-316 U.S.Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop 0-Pl-17 Washington, D.C.20555-0001 Gentlemen:
Donald C.Cook Nuclear Plant Units 1 and 2 CONFIRMATORY ACTION RESPONSE VALIDATION INSPECTION RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING ITEM 3,"36 HOUR COOLDOWN" During the time period January 9, 1998, through January 23, 1998, the NRC conducted a confirmatory action letter (CAL)followup inspection.
During the exit meeting for the inspection, on January 26, 1998, we were requested to docket additional information regarding CAL item 3, involving the ability of the plant to be cooled down in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.Attachment 1 to this letter contains the information requested.
Attachment 2 contains a flow chart depicting the information in attachment 1.Sincerely, EF~pm'.E.Fitzpatrick Vice President/vlb Attachments J.A.Abramson A.B.Beach MDEQ-DW fc RPD NRC Resident Inspector J.R.Sampson 9802i20308 980i29 PDR ADQCK 050003i5)PDR J j.M~~W+IIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIII
'
ATTACHMENT 1 TO AEP:NRC:1260G8 ADDITIONAL INFORMATION RELATED TO 36 HOUR COOLDOHN CONFIRMATORY ACTION LETTER ITEM 3 Attachment 1 to AEP:NRC:1260G8 Page 1 Introduction Item 3 of the confirmatory action letter (CAL)dealt specifically with the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown.During the CAL validation inspection, several concerns were raised by the NRC relating to this issue.This letter is provided to assist with the characterization of'hose concerns.From August 4, 1997, through September 12, 1997, the NRC conducted an architect engineering (AE)inspection at Cook Nuclear Plant.During that inspection, several deficiencies were identified related to the component cooling water (CCW)system.Two such deficiencies were: the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown analysis completed by Westinghouse for Cook Nuclear Plant contained errors, and no maximum cooling flow limits existed to protect the components cooled by CCW from high, potentially damaging flow rates.Subsequently, concurrent efforts were undertaken to correct the cooldown analysis and establish maximum flow limits.On August 29, 1997, condition report (CR)97-2378 was written to capture the AE inspector's concern that maximum flow limits for the CCW system did not exist.The condition report investigation was completed on December 20, 1997.As part of that investigation, technical reviews were completed that established an upper flow limit for each of the components cooled by the CCW system.A commitment was made in that CR response to incorporate the upper flow limits into the updated final safety analysis report (UFSAR).The due date for that commitment is April 1999.The April 1999 date corresponds to the next planned revision of the UFSAR.As part of the process of updating the UFSAR, a 10 CFR 50.59 evaluation is performed.
On September 19, 1997, the NRC issued a CAL to Cook Nuclear Plant.The CAL detailed eight specific items that require resolution prior to restart, one of which is the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown issue.In our letter AEP:NRC:1260G3, dated December 2, 1997, we provided our CAL response to the NRC.From January 9, 1998, to January 23, 1998, the NRC conducted a CAL validation inspection to review and verify the CAL response.Discussion During the validation inspection, the NRC reviewed CAL Item 3, 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown.The review was two-fold.First, the NRC reviewed the actual plant cooldown calculation completed by Westinghouse, as well as the supporting documentation that had been compiled by both us and Westinghouse.
That documentation included Westinghouse's safety evaluation check list (SECL)97-189, and our 10 CFR 50.59 evaluation completed for the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown.It was determined that, during a 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown, the CCW supply temperature may exceed the previously analyzed 95~F and reach a maximum of 120~F.As part of design change 12-DCP-855, the necessary technical reviews and plant modifications were completed to allow the CCW Attachment 1 to AEP:NRC:1260GB Page 2 supply temperature to reach 1204 F.In this letter, the 10 CFR 50.59 evaluation that was completed for 12-DCP-855 is referred to as the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown safety evaluation.
Upon reviewing the Westinghouse cooldown calculation, the NRC questioned two of the inputs used in the analysis.One question dealt with the total CCW flow.In the analysis, the total CCW flow rate was 8,000 gpm (i.e., 4.0E6 lb./hr.).This value was taken directly from table 9.5-3 of the UFSAR.The NRC expressed concern that there was no margin between the analysis input value and the UFSAR value.To address the concern, a 10 CFR 50.59 evaluation was expedited to revise the UFSAR to allow 9,000 gpm of total CCW flow through a single CCW heat exchanger.
As discussed above, the need for this 10 CFR 50.59 review had already been recognized and committed to as part of the investigation of CR 97-2378.It is important to recognize that the analysis input of 8,000 gpm was taken directly from the UFSAR and that the 9,000 gpm value is to be added to the UFSAR to show that margin exists in the cooldown analysis.Another question dealt with the CCW cooling flow to the residual heat removal (RHR)heat exchangers.
This value was 5,000 gpm in the cooldown analysis.The NRC inquired whether instrument uncertainty had been applied to the flow value.The NRC commented that at other plants it was not uncommon to see a flow uncertainty on the order of 1,000 gpm.To address the concern, we provided the NRC with information showing that the uncertainty in the flow indication loop was only on the order of p200 gpm.Additionally, the CCW flow to the RHR heat exchanger is on our critical parameters list.Per a previous commitment made to the NRC in our letter AEP:NRC:1260G3, dated December 2, 1997, instrument uncertainty will be incorporated into the operating procedures for each value on the critical parameters list.This commitment has a due date of December 31, 1998.Thus, the issue about uncertainty in the CCW cooling flow rate to the RHR heat exchangers is already scheduled to be addressed under a pxevious commitment.
to the NRC.In addition to reviewing the actual cooldown calculation, the NRC also reviewed the supporting documentation.
Included in that documentation was SECL 97-189 completed by Westinghouse to support operation of the CCW system at.the elevated supply temperature of 1204 F.In their SECL, Westinghouse stated the control valve.regulating the CCW flow to the letdown heat exchanger will go full open during a 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown.The NRC noted that Westinghouse assumed, with the valve full open, a total CCW flow of 1,000 gpm through the letdown heat exchanger.
The inspector pointed out that table 9.5-2 of the UFSAR lists the design CCW flow to the letdown heat exchanger as 984 gpm and questioned why the apparent discrepancy was not called out in the safety review.The following explanation was provided.The actual design cooling flow rate for the letdown heat exchanger is 492,000 lb./hr.-a mass flow rate, not a volumetric flow rate.This is shown in table 9.2-3 of the UFSAR.The flow rate of 984 gpm, as shown in table 9.5-2 of the UFSAR, is the volumetric flow rate that corresponds to 492,000 lb./hr.at standard conditions.
The flow rate of 1,000 gpm, as shown in the SECL, is the volumetric flow rate that corresponds to 492,000 lb./hr.at the CCW return temperature expected during a 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown.Thus, the volumetric flow rates of 984 gpm and 1,000 gpm are both correct as they are both derived from the design mass flow rate specified for the letdown heat exchanger.
Attachment 1 to AEP:NRC:1260G8 Page 3 As a follow up question to the one above, the NRC asked how much CCW flow is expected to pass through the letdown heat exchanger with the control valve full open.A preliminary calculation determined with the control valve full open, approximately 1,400 gpm would pass through the letdown heat exchanger.
The NRC further questioned why the flow rate, which is greater than the value listed in the UFSAR, was not evaluated in the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown safety evaluation.
To address the immediate concern, a 10 CFR 50.59 evaluation was expedited to evaluate twice the design cooling flow rate through the letdown heat exchanger (approximately 2,000 gpm).As discussed above, the need for this 10 CFR 50.59 review had already been recognized and committed to as part of the investigation of CR 97-2378.Although the letdown heat exchanger cooling flow rate of 1,400 gpm was not specifically addressed in the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown safety review, a bounding analysis had been completed and a commitment to revise the UFSAR already existed.It is important to recognize that, although the increased CCW flow rate to the letdown heat exchanger is a consequence of the higher CCW supply temperature, it is not an input in the cooldown analysis.
ATTACHMENT 2 TO AEP:NRC:1280G8 FLOW CHART DEPICTING INFORMATION PROVIDED IN ATTACHMENT 1
hccacbsienc 2 so AEP/NRCI126008
.QUESTION: The cakuhtion assumes 5000 g pm of CCW coohng flow to the RHR heat exchanger.
Was hstrument uncertainty appfied?ANSWER: The CCW ihw to the RHR heat exchangers is on the critical parameters gsL All procedures which reference the 5000 gpm used in the analysts wfll be revised to account for instrument uncertainty.
This ls a gbbal comnitment to the NRC which is due by the end of 1998.r NRC/AE,INSPECTION August-September 1997 CAL ITEM¹3 36 Hour Cooldown I I I NRC Inspector Reviewed Westinghouse Plant Cooldown Analysis NRC Inspector Reviewed Westinghouse SECL and Our Safety Evaluation QUESllON;The cooldown analysis assumes a total COW lhwof 8000 gpm.The FSAR aflaws a total CCW flaw of 809 gpm through the COW heat exchanger.
Wiry is there no margin between the osculation hpul and the phot design basis?QUESTION: Westinghouse stated that the COW control valve for the letdown heat exchanger wgl go fufl open.Using a CCW flaw rate of 1000 gpm to the letdown heat exchanger, they cahuhted the magnum letdown temperature.
The FSAR states that the CCW Ihw to the letdown heat exchanger Is 984 gpss Why wasnl this addressed in lhe safety evahation?
I I I I I I I I I I I I I I I RESOLUTION:
The 50.59 review to aflaw 9000 g pm through the CCW heat exchanger was expedited.(Comphted 1/23/98)RESOLUTION:
As shawn h the FSAR, the design flaw for the letdown heat exchanger is 492,000 ib/hr.This is equfiratent to 984 gprn at 76'F and 1000 gpm at 120'F.Westinghouse correcfly used 492,000 Ib/hr to cahutate Ihe temperature.
'0 Subsequent to the hspection, a technkst review was completed to determine a maximum aflowabh caofing ihw rate for each component served by CCW.(CR 97-2378)The maxhxrm aflawabfe ihw through the CCW heat exchanger is 9000 gpm.Completed 12/2N97, The maximum aflawable llaw through cooflng fhw through the letdown heat exchanger Is twice design (l.e.984,000 lb/br).ted 12/20/97.A part F convnitment was made In corxfitian report 97-2378 to hcorporate the maximum flaw fimits hto the FSAIL Commitment made: 11/23I97 Commitment due: Apnl 1999 QUESTION: If the control valve does go fufl open, what wifl be the CCW fhw rate to the letdown heat exchanger?
ANSWER With the control valve full open, the CCW flow wg be approxhnatety 1400 gpm..This Is greater than the design fhw rate shown h the FSAIL RESOLUTION:
The 50.59 review to aihw twice design flaw rate through the letdown heat exchanger was expefited.(Completed 1/23lg
