|
|
| Line 16: |
Line 16: |
|
| |
|
| =Text= | | =Text= |
| {{#Wiki_filter:ACCELERATED ITRIBUTION DEMONS'ATION SYSTEMREGULATORY INFORMATION DISTRIBUTION SYSTEM(RIDS)ACCESSION NBR:9405190369 DOC.DATE: | | {{#Wiki_filter:ACCELERATED ITRIBUTION DEMONS'ATION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)ACCESSION NBR:9405190369 DOC.DATE: 94/05/10 NOTARIZED: |
| 94/05/10NOTARIZED: | | NO DOCKET g FACIL:50-315 Donald C.Cook Nuclear Power Plant, Unit 1, Indiana M 05000315 AUTH.NAME AUTHOR AFFILIATION WEBER,G.A. |
| NODOCKETgFACIL:50-315 DonaldC.CookNuclearPowerPlant,Unit1,IndianaM05000315AUTH.NAMEAUTHORAFFILIATION WEBER,G.A.
| | Indiana Michigan Power Co.(formerly Indiana&Michigan Ele BLIND,A.A. |
| IndianaMichiganPowerCo.(formerly Indiana&MichiganEleBLIND,A.A.
| | Indiana Michigan Power Co.(formerly Indiana&Michigan Ele RECIP.NAME RECIPIENT AFFILIATION |
| IndianaMichiganPowerCo.(formerly Indiana&MichiganEleRECIP.NAME RECIPIENT AFFILIATION
| |
|
| |
|
| ==SUBJECT:== | | ==SUBJECT:== |
| LER94-004-00:on 940406,three pressurizer safetyvalves,senttotooffsitetestlabfortestingfailedtomeetTSacceptance criteria.
| | LER 94-004-00:on 940406,three pressurizer safety valves, sent to to off site test lab for testing failed to meet TS acceptance criteria.Cause not determined. |
| Causenotdetermined.
| | Nozzle&disc seating surfaces lapped&polished.W/940510 ltr.DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR (ENCL 0 SIZE: TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.NOTES: RECIPIENT ID CODE/NAME PD3-1 PD INTERNAL: AEOD/DOA AEOD/ROAB/DS P NRR/DE/EMEB NRR/DRCH/HHFB NRR/DRCH/HOLB NRR/DRSS/PRPB NRR/DSSA/SRXB RES/DSIR/EIB EXTERNAL: EG&G BRYCE,J.H NRC PDR NSIC POORE,W.COPIES LTTR ENCL 1 1 1 1 2 2 1 1 1 1 1 1 2 2 1 1 1 1 2 2'1 1 1 1 RECIPIENT ID CODE/NAME HICKMAN,J AEOD/DSP/TPAB NRR/DE/EELB NRR/DORS/OEAB NRR/DRCH/H I CB NRR/DRI L/RPEB NRR DSSA/SPLB GREG 02 FILE 01 L ST LOBBY WARD NSIC MURPHY,G.A NUDOCS FULL TXT COPIES LTTR ENCL 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 NOTE TO ALL"RIDS" RECIPIENTS: |
| Nozzle&discseatingsurfaceslapped&polished.W/940510 ltr.DISTRIBUTION CODE:IE22TCOPIESRECEIVED:LTR (ENCL0SIZE:TITLE:50.73/50.9 LicenseeEventReport(LER),IncidentRpt,etc.NOTES:RECIPIENT IDCODE/NAME PD3-1PDINTERNAL: | | PLEASE HELP US TO REDUCE iVASTE!CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT.20079)TO ELIMINATE YOUR NAME FROiVi DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED'TTR 27 ENCL 27 Indiana Michig Power Company Cook Nuclear Plant One Cook Place Bridgrnan, Ml 49106 616 465 5901 INDlANA NICHlGi4N POWE'R May 10, 1994 United States Nuclear Regulatory Commission Document Control Desk Rockville, Maryland 20852 Operating Licenses DPR-58 Docket No.50-315 Document Control Manager: In accordance with the criteria established by 10 CFR 50.73 entitled Licensee Event Re ort S stem the following report is being submitted: |
| AEOD/DOAAEOD/ROAB/DS PNRR/DE/EMEB NRR/DRCH/HHFB NRR/DRCH/HOLB NRR/DRSS/PRPB NRR/DSSA/SRXB RES/DSIR/EIB EXTERNAL: | | 94-004-00 A.A.Blind Plant Manager/sb Attachment c'.B.Martin, Region III E.E.Fitzpatrick P.A.Barrett R.F.Kroeger M.A.Bailey-Ft.Wayne NRC Resident Inspector J.B.Hickman-NRC J.R.Padgett G.Charnoff, Esq.D.Hahn INPO S.J.Brewer q n.~, i~U l,s<I'-v'7405190369 940510 PDR ADOCK 05000315 S PDR NRC FORM 366 (5 92)U.S.NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO.3150.0104 EXP IR ES 5/31/95 LICENSEE EVENT REPORT (LER)(See reverse for required number of digits/characters for each block)ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMADON COLLECTION REQUEST: 50.0 HRS.FORWARD COMMENTS REGARDUIG BURDEN ESDMATE TO THE INFQRMATICN AND RECORDS MANAGEMENT BRANCH (MNBB Tr ta), U.S.NUCLEAR REGULATORY COMMISSION> |
| EG&GBRYCE,J.H NRCPDRNSICPOORE,W.COPIESLTTRENCL11112211111122111122'1111RECIPIENT IDCODE/NAME HICKMAN,J AEOD/DSP/TPAB NRR/DE/EELB NRR/DORS/OEAB NRR/DRCH/H ICBNRR/DRIL/RPEBNRRDSSA/SPLB GREG02FILE01LSTLOBBYWARDNSICMURPHY,G.A NUDOCSFULLTXTCOPIESLTTRENCL111111111111111111111111NOTETOALL"RIDS"RECIPIENTS: | | WASHINGTON, DC 20555 0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150010a), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503, FACILITY NAME (I)D.C.COOK NUCLEAR PLANT-UNIT 1 DOCKET NUMBER (2)05000 315 PAGE P)10F 3 FAILURE OF THREE PRESSURIZER SAFETY VALVES TO'MEET TECHNICAL SPECIFICATION RVEILLANCE TEST CRITERIA EVENT DATE 5 LER NUMBER 6 REPORT NUMBER 7 OTHER FACILITIES INVOLVED 6 MONTH 04 DAY 06 YEAR 94 YEAR 94 SEQUENTIAL NUMBER 004 REVISION NUMBER 0 MONTH DAY 05 10 94 FACIUTY NAME FACILITY NAME DOCKET NUMBER 05000 DOCKET NUMBcR 05000 OPERATING MODE (9)20A02(b)20.405(c)50.73(a)(2)(iv)73.71(b)NT TO THE REQUIREMENTS OF 10 CFR E: Check one or m THIS REPORT IS SUBMITTED PURSUA ore 11 POWER LEVEL (10)0 20.405(a)(1)(i)20.405(a)(1)(ii)20.405(a)(1)(iii)20.405(a)(1)(iv)2o.405(a)(1)(v) 50.36(c)(1)50.36(c)(2)X 50.73(a)(2)(i) 50.73(a)(2)(ii)50.73(a)(2)(iii)LICENSEE CONTACT FOR THIS LER 12 M73(a)(2)(v) 50.73(a)(2)(vii)50.73(a)(2)(viii)(A)50.73(a)(2)(viii)(B) 50.73(a)(2)(x) 73.71(c)OTHER (Speoty in Atrctract tretow and in Text, NRC Form 366A)NAME G.A.WEBER-PLANT ENGINEERING SUPERINTENDENT TELEPHONE NUMBER Fnctvde Area code)(616)465-5902 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 13 CAUSE SYSTEM COMPONENT RV MANUFACTURER C710 REPORTABLE TO NPRDS , D.;j.x',~CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NPRDS SUPPLEMENTAL REPORT EXPECTED 14 YES 0(yee, cornptete ExPEGTED sUBMIssIQN DATE)NO ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single.spaced typewritten lines)(16)EXPFCTFD MONTH DAY YEAR SUBMISSION DATE (15)On Apri.l 6, 1994 with Uni.t 1 in Mode 5 (Cold Shutdown)it was determined that all three of the pressurizer safety valves, which were sent to an off site test laboratory for testing, were found with lift settings outside of the Technical Specification acceptance criteria.Acceptable settings are between, 2461 psig and 2509 psig.Valve 1-SV-45A was found to have a lift setpoint of 2536 psig, valve 1-SV-45B had a lift setpoint of 2535 psig and 1-SV-45C had a lift setpoint of 2538 psig.There was no safety-signi.ficance since the worst case (1-SV-45C-lift setpoint of 2538 psig)would result in a maximum transient pressure of 2615 psig (2538 psig plus 3 percent accumulation to attain its full rated lift).This is below the Technical Specification safety limit of 2735 psig.All three valves were partially disassembled (retaining spring compression) and inspected. |
| PLEASEHELPUSTOREDUCEiVASTE!CONTACTTHEDOCUMENTCONTROLDESK,ROOMPl-37(EXT.20079)TOELIMINATE YOURNAMEFROiViDISTRIBUTION LISTSFORDOCUMENTS YOUDON'TNEED!FULLTEXTCONVERSION REQUIREDTOTALNUMBEROFCOPIESREQUIRED'TTR 27ENCL27 IndianaMichigPowerCompanyCookNuclearPlantOneCookPlaceBridgrnan, Ml491066164655901INDlANANICHlGi4N POWE'RMay10,1994UnitedStatesNuclearRegulatory Commission DocumentControlDeskRockville, Maryland20852Operating LicensesDPR-58DocketNo.50-315DocumentControlManager:Inaccordance withthecriteriaestablished by10CFR50.73entitledLicenseeEventReortSstemthefollowing reportisbeingsubmitted:
| | No problems were noted.The nozzle and disc seating surfaces were lapped and polished.The valves were reassembled and tested satisfactori.ly. |
| 94-004-00 A.A.BlindPlantManager/sbAttachment c'.B.Martin,RegionIIIE.E.Fitzpatrick P.A.BarrettR.F.KroegerM.A.Bailey-Ft.WayneNRCResidentInspector J.B.Hickman-NRCJ.R.PadgettG.Charnoff, Esq.D.HahnINPOS.J.Brewerqn.~,i~Ul,s<I'-v'7405190369 940510PDRADOCK05000315SPDR NRCFORM366(592)U.S.NUCLEARREGULATORY COMMISSION APPROVEDBYOMBNO.3150.0104 EXPIRES5/31/95LICENSEEEVENTREPORT(LER)(Seereverseforrequirednumberofdigits/characters foreachblock)ESTIMATED BURDENPERRESPONSETOCOMPLYWITHTHISINFORMADON COLLECTION REQUEST:50.0HRS.FORWARDCOMMENTSREGARDUIG BURDENESDMATETOTHEINFQRMATICN ANDRECORDSMANAGEMENT BRANCH(MNBBTrta),U.S.NUCLEARREGULATORY COMMISSION> | | The safety valve conditions experienced at the D.C.Cook Plant are similar to current i.ndustry trends/concerns. |
| WASHINGTON, DC205550001,ANDTOTHEPAPERWORK REDUCTION PROJECT(3150010a), | | -Since a specific Root Cause could not be determined, no preventive action i.s planned at this time.However, we will be evaluating the test methods and industry activities pertaining to the pressurizer safety valves.NRC FORM 366 (5-92) |
| OFFICEOFMANAGEMENT ANDBUDGET,WASHINGTON, DC20503,FACILITYNAME(I)D.C.COOKNUCLEARPLANT-UNIT1DOCKETNUMBER(2)05000315PAGEP)10F3FAILUREOFTHREEPRESSURIZER SAFETYVALVESTO'MEETTECHNICAL SPECIFICATION RVEILLANCE TESTCRITERIAEVENTDATE5LERNUMBER6REPORTNUMBER7OTHERFACILITIES INVOLVED6MONTH04DAY06YEAR94YEAR94SEQUENTIAL NUMBER004REVISIONNUMBER0MONTHDAY051094FACIUTYNAMEFACILITYNAMEDOCKETNUMBER05000DOCKETNUMBcR05000OPERATING MODE(9)20A02(b)20.405(c) 50.73(a)(2)(iv)73.71(b)NTTOTHEREQUIREMENTS OF10CFRE:CheckoneormTHISREPORTISSUBMITTED PURSUAore11POWERLEVEL(10)020.405(a)
| | REQUIRED NUMBER OF DIGITS/CHARACTERS FOR EACH BLOCK BLOCK NUMBER NUMBER OF DIGITS/CHARACTERS UP TO 46 8 TOTAL 3 IN ADDITION TO 05000 VARIES TITLE FACILITY NAME DOCKET NUMBER PAGE NUMBER 10 12 13 14 15 UP TO 76 6 TOTAL 2 PER BLOCK 7 TOTAL 2 FOR YEAR 3 FOR SEQUENTIAL NUMBER 2 FOR REVISION NUMBER 6 TOTAL 2 PER BLOCK UP TO 18*.FACILITY NAME 8 TOTAL-DOCKET NUMBER 3 IN ADDITION TO 05000 1 CHECK BOX THAT APPLIES UP TO 50 FOR NAME 14 FOR TELEPHONE CAUSE VARIES 2 FOR SYSTEM 4 FOR COMPONENT 4 FOR MANUFACTURER NPRDS VARIES 1 CHECK BOX THAT APPLIES 6 TOTAL 2 PER BLOCK TITLE EVENT DATE LER NUMBER REPORT DATE OTHER FACILITIES INVOLVED OPERATING MODE POWER LEVEL REQUIREMENTS OF 10 CFR LICENSEE CONTACT EACH COMPONENT FAILURE SUPPLEMENTAL REPORT EXPECTED EXPECTED SUBMISSION DATE NRC FORM 366A r (669)U.S.NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION APPROVED OMB NO.31500104 EXPIRES: 4/30I92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUESTI 50.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (F630), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, OC 20555, AND TO 1HE PAPERWORK REDUCTION PRO)ECT (31500104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.FACILITY NAME (1)DOCKE'T NUMBER (2)LER NUMBER (6)PAGE (3)YEAR Hr5 SEOUENTIAL NUMBER REVISION NUMSFR D.C.COOK NUCLEAR PLANT-UNIT 1 TEXT (if more 4Pece Je eetrr9ed, Iree eddttr'one) |
| (1)(i)20.405(a)(1) | | NRC Forrrr 3IJSA'et (12)o 5 o o o 315 94 0 0 4-0 0 0 2 OF 0 3 Conditions Prior to Occurrence: |
| (ii)20.405(a) | | Unit One-Mode 5 (Cold Shutdown-following refueling). |
| (1)(iii)20.405(a) | | Descri tion of Event: On April 6, 1994, it was determined that all three safety pressurizer safety valves, Crosby Valve Model HB-86-PB, (EZIS/AB-RV) had lift settings outside Technical Specification |
| (1)(iv)2o.405(a)(1)(v) 50.36(c)(1)50.36(c)(2)X50.73(a)(2)(i) 50.73(a)(2)(ii)50.73(a)(2) | | |
| (iii)LICENSEECONTACTFORTHISLER12M73(a)(2)(v) 50.73(a)(2)(vii)50.73(a)(2) | | ====3.4.3 acceptance==== |
| (viii)(A) 50.73(a)(2)(viii)(B) 50.73(a)(2)(x) 73.71(c)OTHER(SpeotyinAtrctract tretowandinText,NRCForm366A)NAMEG.A.WEBER-PLANTENGINEERING SUPERINTENDENT TELEPHONE NUMBERFnctvdeAreacode)(616)465-5902COMPLETEONELINEFOREACHCOMPONENT FAILUREDESCRIBED INTHISREPORT13CAUSESYSTEMCOMPONENT RVMANUFACTURER C710REPORTABLE TONPRDS,D.;j.x',~CAUSESYSTEMCOMPONENT MANUFACTURER REPORTABLE TONPRDSSUPPLEMENTAL REPORTEXPECTED14YES0(yee,cornptete ExPEGTEDsUBMIssIQN DATE)NOABSTRACT(Limitto1400spaces,i.e.,approximately 15single.spacedtypewritten lines)(16)EXPFCTFDMONTHDAYYEARSUBMISSION DATE(15)OnApri.l6,1994withUni.t1inMode5(ColdShutdown) itwasdetermined thatallthreeofthepressurizer safetyvalves,whichweresenttoanoffsitetestlaboratory fortesting,werefoundwithliftsettingsoutsideoftheTechnical Specification acceptance criteria. | | criteria.The safety valves are tested at a test laboratory using steam at nominal temperature and pressure, as required by Technical Specification. |
| Acceptable settingsarebetween,2461psigand2509psig.Valve1-SV-45Awasfoundtohavealiftsetpointof2536psig,valve1-SV-45Bhadaliftsetpointof2535psigand1-SV-45Chadaliftsetpointof2538psig.Therewasnosafety-signi.ficance sincetheworstcase(1-SV-45C-liftsetpointof2538psig)wouldresultinamaximumtransient pressureof2615psig(2538psigplus3percentaccumulation toattainitsfullratedlift).ThisisbelowtheTechnical Specification safetylimitof2735psig.Allthreevalveswerepartially disassembled (retaining springcompression) andinspected. | | The valves are required to lift at 2485 psig plus or minus 1 percent, (i.e.between 2461 and 2509 psig).Valve 1-SV-45A was found to have a lift setpoint of 2536 psig, valve 1-SV-45B had a lift setpoint of 2535 psig and 1-SV-45C had a lift setpoint of 2538 psig.Technical Specification 4.4:3 requires that each Pressurizer Code Safety Valve be demonstrated operable per Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition.Cause of Event: The safety valve conditions experienced at the D.C.Cook Plant are similar to current industry trends/concerns. |
| Noproblemswerenoted.Thenozzleanddiscseatingsurfaceswerelappedandpolished.
| | The phenomenon of safety valve setpoint drift outside of design tolerances is common in the Nuclear industry.However, as yet, no cause for the drift has been determined. |
| Thevalveswerereassembled andtestedsatisfactori.ly.
| | 1-SV-45A, 1-SV-45B and 1-SV-45C were partially disassembled (retaining spring compression) and inspected. |
| Thesafetyvalveconditions experienced attheD.C.CookPlantaresimilartocurrenti.ndustry trends/concerns.
| | No problems were noted.The nozzle and disc seating surfaces were lapped and polished.The valves were reassembled and tested satisfactorily. |
| -SinceaspecificRootCausecouldnotbedetermined, nopreventive actioni.splannedatthistime.However,wewillbeevaluating thetestmethodsandindustryactivities pertaining tothepressurizer safetyvalves.NRCFORM366(5-92) | | The test program (test facility and procedure) for the Pressurizer Safety Valves has not changed from previous testing.Anal sis of Event: This event has been determined to be reportable under the provisions of 10CFR5073(a)(2)(i)(B) as an operation prohibited by Plant Technical Specification 3.4.3, which requires all of the pressurizer safety valves to be operable with a lift setting of 2485 psig+/-1 percent.The as-found lift setpoints of safety valves 1-SV-45A, 1-SV-45B and 1-SV-45C did not have any actual impact on the Reactor Coolant System (RCS)since the safety valves were not challenged during the last fuel cycle.There was no potential impact since the RCS would not have exceeded the maximum transient limit of 2735 psig, which is 110 percent of design pressure (2485 psig).There was no impact on the health or safety of the public.Safety Valve 1-SV-45C (worst case)had a lift setpoint of 2538 psig.The RCS pressure would have to reach a pressure of 2615 psig (2538 psig plus 3 percent accumulation) for this valve to attain its full rated lift.Valve 1-SV-45A would have attained its rated lift at 2612 psig (2536 psig plus 3 percent)and 1-SV-45B would have attained its rated lift at 2612 psig (2535 plus 3 percent).NRC Form 366A (669) |
| REQUIREDNUMBEROFDIGITS/CHARACTERS FOREACHBLOCKBLOCKNUMBERNUMBEROFDIGITS/CHARACTERS UPTO468TOTAL3INADDITIONTO05000VARIESTITLEFACILITYNAMEDOCKETNUMBERPAGENUMBER1012131415UPTO766TOTAL2PERBLOCK7TOTAL2FORYEAR3FORSEQUENTIAL NUMBER2FORREVISIONNUMBER6TOTAL2PERBLOCKUPTO18*.FACILITYNAME8TOTAL-DOCKETNUMBER3INADDITIONTO050001CHECKBOXTHATAPPLIESUPTO50FORNAME14FORTELEPHONE CAUSEVARIES2FORSYSTEM4FORCOMPONENT 4FORMANUFACTURER NPRDSVARIES1CHECKBOXTHATAPPLIES6TOTAL2PERBLOCKTITLEEVENTDATELERNUMBERREPORTDATEOTHERFACILITIES INVOLVEDOPERATING MODEPOWERLEVELREQUIREMENTS OF10CFRLICENSEECONTACTEACHCOMPONENT FAILURESUPPLEMENTAL REPORTEXPECTEDEXPECTEDSUBMISSION DATE NRCFORM366Ar(669)U.S.NUCLEARREGULATORY COMMISSION LICENSEEEVENTREPORT(LER)TEXTCONTINUATION APPROVEDOMBNO.31500104EXPIRES:4/30I92ESTIMATED BURDENPERRESPONSETOCOMPLYWTHTHISINFORMATION COLLECTION REOUESTI50.0HRS.FORWARDCOMMENTSREGARDING BURDENESTIMATETOTHERECORDSANDREPORTSMANAGEMENT BRANCH(F630),U.S.NUCLEARREGULATORY COMMISSION, WASHINGTON, OC20555,ANDTO1HEPAPERWORK REDUCTION PRO)ECT(31500104),
| | NRC FORM 366A (64)9)U.S.NUCLEAR REGULATORY COMMISSIO LICENSEE EVENT REPORT (LER)TEXT CONTINUATION APPROVED OMB NO.31504))04 EXPIRES)4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REGUESTI 500 HRS.FORWARD COMMENTS REGARDING BUADEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH IP430), U.S.NUCLEAR AEGULATOAY COMMISSION, WASHINGTON, OC 20555, AND TO THE PAPERWORK REDUCTION PAOJECT (31500104), OFFICE OF MANAGF MENT AND BUDGET, WASHINGTON, OC 20503.FACILITY NAME (1)DOCKET NUMBER (21 LFR NUMBER (6)PAGE (3)YEAR)'.yi(SSOVSNTIAL NVMSSR REVISION NVM 54 D.C.COOK NUCLEAR PLANT-UNIT 1 TEXT///mttt tptct B nqvnd.IItt tddldoIM//Y/IC Fdmt 3/)BA'4/(I 2)0 5 0 0 0 3 I 5 9 4 0 0 4 00 03 oF 0 Anal sis of Event continued! |
| OFFICEOFMANAGEMENT ANDBUDGET,WASHINGTON, DC20503.FACILITYNAME(1)DOCKE'TNUMBER(2)LERNUMBER(6)PAGE(3)YEARHr5SEOUENTIAL NUMBERREVISIONNUMSFRD.C.COOKNUCLEARPLANT-UNIT1TEXT(ifmore4PeceJeeetrr9ed, Ireeeddttr'one)
| | The reactor vessel and pressurizer were designed to ASME BGPV Section III which permits a maximum transient. |
| NRCForrrr3IJSA'et(12)o5ooo31594004-0002OF03Conditions PriortoOccurrence:
| | pressure of 2735 psig, 110 percent of design pressure (2485 psig).The RCS piping, valves and fittings are designed to ANSI B31.1, 1967 Edition, which permits a maximum transient pressure of 2985 psig, 120 percent of design pressure (2485 psig).In addition, the entire RCS was hydro tested to 3107 psig, 125 percent of design (2485 psig), to demonstrate system integrity prior to initial operation. |
| UnitOne-Mode5(ColdShutdown-following refueling).
| | In conclusion, this event did not have any safety significance and did not represent a hazard to the public health and safety.The safety limit of 2735 psig would not have been exceeded since the maximum RCS pressure would not have exceed 2615 psig (1-SV-45C setpoint of 2538 psig plus 3 percent).In addition to this Safety Analysis, an additional Safety Evaluation is being performed to review the combined effect of recent Main Steam Safety Valve lift setpoints (reported in LER 50-315/94-001 and 94-003)in conjunction with the as-found lift setpoints of the Pressurizer Safety Valves.This evaluation is scheduled to be completed by June 30, 1994.We do not anticipate any adverse conditions to be identified during this evaluation, however, an updated LER will be submitted if deemed necessary. |
| DescritionofEvent:OnApril6,1994,itwasdetermined thatallthreesafetypressurizer safetyvalves,CrosbyValveModelHB-86-PB, (EZIS/AB-RV) hadliftsettingsoutsideTechnical Specification 3.4.3acceptance criteria.
| | Corrective Action: The nozzle and disc seating surfaces were lapped and polished for all three safety valves.Retests were satisfactorily completed for steam set pressure and seat leakage.The spring pressure on 1-SV-45A was adjusted by 1/3 flat to bring the set pressure back into tolerance. |
| Thesafetyvalvesaretestedatatestlaboratory usingsteamatnominaltemperature andpressure, asrequiredbyTechnical Specification.
| | Since a specific Root Cause could not be determined, no preventive action is planned at this time.However, we will continue to follow industry activities pertaining to safety valve setpoint drift.Failed Com onent Identification: |
| Thevalvesarerequiredtoliftat2485psigplusorminus1percent,(i.e.between2461and2509psig).Valve1-SV-45Awasfoundtohavealiftsetpointof2536psig,valve1-SV-45Bhadaliftsetpointof2535psigand1-SV-45Chadaliftsetpointof2538psig.Technical Specification 4.4:3requiresthateachPressurizer CodeSafetyValvebedemonstrated operableperSectionXIoftheASMEBoilerandPressureVesselCode,1974Edition.CauseofEvent:Thesafetyvalveconditions experienced attheD.C.CookPlantaresimilartocurrentindustrytrends/concerns.
| | Pressurizer Safety Valve Plant Designation: |
| Thephenomenon ofsafetyvalvesetpointdriftoutsideofdesigntolerances iscommonintheNuclearindustry.
| | 1-SV-45A, 1-SV-45B and 1-SV-45C Manufacturer: |
| However,asyet,nocauseforthedrifthasbeendetermined. | | Crosby Valve Company Model: HB-86-BP EIIS Code: AB-RV Previous Similar Events: LERS 50-315/90-16, 92-09 LER: 50-316/89-04, 92-06 NRC Form 366A (669)}} |
| 1-SV-45A, 1-SV-45Band1-SV-45Cwerepartially disassembled (retaining springcompression) andinspected. | |
| Noproblemswerenoted.Thenozzleanddiscseatingsurfaceswerelappedandpolished.
| |
| Thevalveswerereassembled andtestedsatisfactorily.
| |
| Thetestprogram(testfacilityandprocedure) forthePressurizer SafetyValveshasnotchangedfromprevioustesting.AnalsisofEvent:Thiseventhasbeendetermined tobereportable undertheprovisions of10CFR5073(a)(2)(i)(B) asanoperation prohibited byPlantTechnical Specification 3.4.3,whichrequiresallofthepressurizer safetyvalvestobeoperablewithaliftsettingof2485psig+/-1percent.Theas-foundliftsetpoints ofsafetyvalves1-SV-45A, 1-SV-45Band1-SV-45CdidnothaveanyactualimpactontheReactorCoolantSystem(RCS)sincethesafetyvalveswerenotchallenged duringthelastfuelcycle.Therewasnopotential impactsincetheRCSwouldnothaveexceededthemaximumtransient limitof2735psig,whichis110percentofdesignpressure(2485psig).Therewasnoimpactonthehealthorsafetyofthepublic.SafetyValve1-SV-45C(worstcase)hadaliftsetpointof2538psig.TheRCSpressurewouldhavetoreachapressureof2615psig(2538psigplus3percentaccumulation) forthisvalvetoattainitsfullratedlift.Valve1-SV-45Awouldhaveattaineditsratedliftat2612psig(2536psigplus3percent)and1-SV-45Bwouldhaveattaineditsratedliftat2612psig(2535plus3percent).
| |
| NRCForm366A(669)
| |
| NRCFORM366A(64)9)U.S.NUCLEARREGULATORY COMMISSIO LICENSEEEVENTREPORT(LER)TEXTCONTINUATION APPROVEDOMBNO.31504))04 EXPIRES)4/30/92ESTIMATED BURDENPERRESPONSETOCOMPLYWTHTHISINFORMATION COLLECTION REGUESTI500HRS.FORWARDCOMMENTSREGARDING BUADENESTIMATETOTHERECORDSANDREPORTSMANAGEMENT BRANCHIP430),U.S.NUCLEARAEGULATOAY COMMISSION, WASHINGTON, OC20555,ANDTOTHEPAPERWORK REDUCTION PAOJECT(31500104),
| |
| OFFICEOFMANAGFMENTANDBUDGET,WASHINGTON, OC20503.FACILITYNAME(1)DOCKETNUMBER(21LFRNUMBER(6)PAGE(3)YEAR)'.yi(SSOVSNTIAL NVMSSRREVISIONNVM54D.C.COOKNUCLEARPLANT-UNIT1TEXT///mttttptctBnqvnd.IItttddldoIM//Y/IC Fdmt3/)BA'4/(I2)050003I5940040003oF0AnalsisofEventcontinued!
| |
| Thereactorvesselandpressurizer weredesignedtoASMEBGPVSectionIIIwhichpermitsamaximumtransient.
| |
| pressureof2735psig,110percentofdesignpressure(2485psig).TheRCSpiping,valvesandfittingsaredesignedtoANSIB31.1,1967Edition,whichpermitsamaximumtransient pressureof2985psig,120percentofdesignpressure(2485psig).Inaddition, theentireRCSwashydrotestedto3107psig,125percentofdesign(2485psig),todemonstrate systemintegrity priortoinitialoperation.
| |
| Inconclusion, thiseventdidnothaveanysafetysignificance anddidnotrepresent ahazardtothepublichealthandsafety.Thesafetylimitof2735psigwouldnothavebeenexceededsincethemaximumRCSpressurewouldnothaveexceed2615psig(1-SV-45C setpointof2538psigplus3percent).
| |
| InadditiontothisSafetyAnalysis, anadditional SafetyEvaluation isbeingperformed toreviewthecombinedeffectofrecentMainSteamSafetyValveliftsetpoints (reported inLER50-315/94-001 and94-003)inconjunction withtheas-foundliftsetpoints ofthePressurizer SafetyValves.Thisevaluation isscheduled tobecompleted byJune30,1994.Wedonotanticipate anyadverseconditions tobeidentified duringthisevaluation, however,anupdatedLERwillbesubmitted ifdeemednecessary.
| |
| Corrective Action:Thenozzleanddiscseatingsurfaceswerelappedandpolishedforallthreesafetyvalves.Retestsweresatisfactorily completed forsteamsetpressureandseatleakage.Thespringpressureon1-SV-45Awasadjustedby1/3flattobringthesetpressurebackintotolerance. | |
| SinceaspecificRootCausecouldnotbedetermined, nopreventive actionisplannedatthistime.However,wewillcontinuetofollowindustryactivities pertaining tosafetyvalvesetpointdrift.FailedComonentIdentification:
| |
| Pressurizer SafetyValvePlantDesignation: | |
| 1-SV-45A, 1-SV-45Band1-SV-45CManufacturer: | |
| CrosbyValveCompanyModel:HB-86-BPEIISCode:AB-RVPreviousSimilarEvents:LERS50-315/90-16, 92-09LER:50-316/89-04, 92-06NRCForm366A(669)}}
| |
Similar Documents at Cook |
|---|
Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML17335A5641999-10-18018 October 1999
LER 99-024-00:on 990708,literal TS Requirements Were Not Met by Accumlator Valve Surveillance.Caused by Misjudgement Made in Conversion from Initial DC Cook TS to W Std Ts.Submitted License Amend Request.With 991018 Ltr
ML17335A5531999-10-0707 October 1999
LER 99-023-00:on 990907,inadequate TS Surveillance Testing of ESW Pump ESF Response Time Noted.Caused by Inadequate Understanding of Plant Design Basis.Surveillance Tests Will Be Revised & Implemented
ML17326A1291999-09-17017 September 1999
LER 99-022-00:on 990609,electrical Bus Degraded Voltage Setpoints Too Low for Safety Related Loads,Was Discovered. Caused by Lack of Understanding of Design of Plant.No Immediate Corrective Actions Necessary
ML17326A1121999-08-27027 August 1999
LER 99-021-00:on 990728,determined That GL 96-01 Test Requirements Were Not Met in Surveillance Tests.Caused by Failure to Understand Full Extent of GL Requirements. Surveillance Procedures Will Be Revised or Developed
ML17326A1011999-08-26026 August 1999
LER 99-020-00:on 990727,EDGs Were Declared Inoperable.Caused by Inadequate Protection of Air Intake,Exhaust & Room Ventilation Structures from Tornado Missile Hazards. Implemented Compensatory Measures in Form of ACs
ML17326A0911999-08-16016 August 1999
LER 99-019-00:on 990716,noted Victoreen Containment Hrrms Not Environmentally Qualified to Withstand post-LOCA Conditions.Caused by Inadequate Design Control.Reviewing Options to Support Hrrms Operability in Modes 1-4
ML17326A0771999-08-0404 August 1999
LER 98-029-01:on 980422,noted That Fuel Handling Area Ventilation Sys Was Inoperable.Caused by Original Design Deficiency.Radiological Analysis for Spent Fuel Handling Accidents in Auxiliary Bldg Will Be Redone by 990830
ML17326A0741999-07-29029 July 1999
LER 99-018-00:on 990629,determined That Valve Yokes May Yield Under Combined Stress of Seismic Event & Static,Valve Closed,Stem Thrust.Caused by Inadequate Design of Associated Movs.Operability Determinations Were Performed for Valves
ML17326A0661999-07-26026 July 1999
LER 99-017-00:on 990625,noted That Improperly Installed Fuel Oil Return Relief Valve Rendered EDG Inoperable.Caused by Personnel Error.Fuel Oil Return Valve Was Replaced with Valve in Correct Orientation.With 990722 Ltr
ML17326A0651999-07-22022 July 1999
LER 98-014-03:on 980310,noted That Response to high-high Containment Pressure Procedure Was Not Consistent with Analysis of Record.Caused by Inadequate Interface with W. FRZ-1 Will Be Revised to Be Consistent with New Analysis
ML17326A0491999-07-13013 July 1999
LER 99-016-00:on 990615,TS Requirements for Source Range Neutron Flux Monitors Not Met.Caused by Failure to Understand Design Basis of Plant.Procedures Revised.With 990713 Ltr
ML17326A0331999-07-0101 July 1999
LER 99-004-01:on 971030,failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed Was Noted.Caused by Ineffective Mgt of Tss.Chemistry Personnel Have Been Instructed on Requirement to Follow TS as Written
ML17326A0151999-06-18018 June 1999
LER 99-014-00:on 990521,determined That Boron Injection Tank Manway Bolts Were Not Included in ISI Program,Creating Missed Exam for Previous ISI Interval.Caused by Programmatic Weakness.Isi Program & Associated ISI Database Modified
ML17325B6311999-06-0101 June 1999
LER 99-S03-00:on 990430,vital Area Barrier Degradation Was Noted.Caused by Inadequate Insp & Maint of Vital Area Barrier.Repairs & Mods Were Made to Barriers to Eliminate Degraded & Nonconforming Conditions
ML17325B6421999-06-0101 June 1999
LER 99-013-00:on 990327,safety Injection & Centrifugal Charging Throttle Valve Cavitation During LOCA Could Have Led to ECCS Pump Failure.Caused by Inadequate Original Design Application of Si.Throttle Valves Will Be Developed
ML17325B6351999-05-28028 May 1999
LER 99-S02-00:on 990428,vulnerability in Safeguard Sys That Could Allow Unauthorized Access to Protected Area Was Noted. Caused by Inadequate Original Plant Design.Mods Were Made to Wall Opening to Eliminate Nonconforming Conditions
ML17265A8231999-05-24024 May 1999
LER 98-037-01:on 990422,determined That Ice Condenser Bypass Leakage Exceeds Design Basis Limit.Caused by Pressure Seal Required by Revised W Design Not Incorporated Into Aep Design.Numerous Matl Condition Walkdowns & Assessments Made
ML17325B6001999-05-20020 May 1999
LER 99-012-00:on 990420,concluded That Auxiliary Bldg ESF Ventilation Sys Not Capable of Maintaining ESF Room Temps post-accident.Caused by Inadequate Control of Sys Design Inputs.Comprehensive Action Plan Being Developed
ML17325B5861999-05-10010 May 1999
LER 99-002-00:on 990415,discovered That TS 4.0.5 Requirements Were Not Met Due to Improperly Performed Test. Caused by Incorrect Interpretation of ASME Code.App J Testing Will Be Completed & Procedures Will Be Revised
ML17325B5811999-05-0404 May 1999
LER 99-011-00:on 990407,air Sys for EDG Will Not Support Long Operability.Caused by Original Design Error.Temporary Mod to Supply Makeup Air Capability in Modes 5 & 6 Was Prepared
ML17325B5771999-05-0303 May 1999
LER 99-010-00:on 990401,RCS Leak Detection Sys Sensitivity Not in Accordance with Design Requirements Occurred.Caused by Inadequate Original Design of Containment Sump Level. Evaluation Will Be Performed to Clearly Define Design
ML17325B5581999-04-16016 April 1999
LER 99-006-00:on 990115,personnel Identified Discrepancy Between TS 3.9.7 Impact Energy Limit & Procedure 12 Ohp 4030.STP.046.Caused by Lack of Design Basis Control.Placed Procedure 12 Ohp 4030.STP.046 on Administrative Hold
ML17325B5471999-04-12012 April 1999
LER 99-009-00:on 990304,as-found RHR Safety Relief Valve Lift Setpoint Greater than TS Limit Occurred.Cause Investigation for Condition Has Not Been Completed.Update to LER Will Be Submitted,Upon Completion of Investigation
ML17325B5321999-04-0707 April 1999
LER 99-S01-00:on 990308,discovered That Lock for Vital Gate Leading to Plant 4KV Switchgear Area Was Nonconforming & Vulnerable to Unauthorized Access.Caused by Inadequate Gate Design & Inadequate Procedures.Mods Are Being Made to Gate
ML17325B5161999-04-0101 April 1999
LER 99-007-00:on 981020,calculations Showed That Divider Barrier Between Upper & Lower Containment Vols Were Overstressed.Engineers Are Currently Working on Analyses of Loads & Stress on Enclosures
ML17325B5221999-03-29029 March 1999
LER 99-001-00:on 960610,degraded Component Cooling Water Flow to Containment Main Steam Line Penetrations,Identified on 990226.Caused by Inadequate Understanding of Design Basis.Additional Investigations Ongoing
ML17325B4801999-03-18018 March 1999
LER 99-004-00:on 971030,failure to Perform TS Surveillance Analyses of Rc Chemistry with Fuel Removed Was Noted.Cause of Event Is Under Investigation.Corrected Written Job Order Activities Used to Control SD Chemistry Sampling
ML17325B4741999-03-18018 March 1999
LER 99-005-00:on 940512,determined That Rt Breaker Manual Actuations During Rod Drop Testing Were Not Previously Reported.Caused by Lack of Training.Addl Corrective Actions,Including Preventative Actions May Be Developed
ML17325B4571999-02-24024 February 1999
LER 99-003-00:on 990107,CR Pressurization Sys Surveillance Test Did Not Test Sys in Normal Operating Condition.Caused by Failure to Recognize Door 12DR-AUX415 as Part of CR Pressure Boundary.Performed Walkdown of Other Doors
ML17335A5171999-02-11011 February 1999
LER 99-002-00:on 990112,determined That RCS Pressurizer PORVs Had Not Been Tested,Per Ts.Caused by Inadequate Scheduling Controls Allowing Personnel Error.Surveillance Procedure Was Completed & Updated LER Will Be Submitted
ML17335A5141999-02-10010 February 1999
LER 99-001-00:on 990106,noted That GE Hfa Relays Installed in EDGs May Not Meet Seismic Qualification.Caused by Operating Experience Info Incorrectly Dispositioned in 1985. Updated LER Will Be Submitted by 990405
ML17335A5011999-02-0101 February 1999
LER 98-060-00:on 981231,identified That Rt Sys Response Time Testing Did Not Comply with TS Definition.Caused by Inadequate Procedures.Corrective Actions Will Be Developed & Update to LER Will Be Submitted by 990415.With 990201 Ltr
ML17335A4951999-01-29029 January 1999
LER 98-059-00:on 981230,interim LER -single Failure in Containment Spray Sys Could Result in Containment Spray Ph Outside Design Occurred.Investigation Into Condition Continuing.Update Will Be Submitted by 990514 Ltr
ML17335A4961999-01-27027 January 1999
LER 98-057-00:on 981228,discovered That AFW Valves Were Not Tested IAW Inservice Testing Program.Caused by Failure to Recognize Design Bases Features Re Afws by Personnel. Updated LER Will Be Submittted by 990415.With 990127 Ltr
ML17335A4921999-01-19019 January 1999
LER 98-052-01:on 981128,no Analysis for NSR Sc Manual Loader for Tdafwp Could Be Found in Original Design.Cause Due to All Failure Modes Not Considered When Compressed Air Sys Originally Designed.Performed Review of Components
ML17335A4721999-01-0606 January 1999
LER 98-055-00:on 981207,potential for Condition Outside of Design Bases for Rod Control Sys Was Noted.Caused by Calibration Error Coupled with Single Rod Failure.Condition Rept Investigation Is Ongoing
ML17335A4691999-01-0606 January 1999
LER 98-056-00:on 981211,hot Leg Nozzle Gaps Resulted in Plant Being in Unanalyzed Condition.Analyses Are Being Performed by W to Resolve Problem.Updated LER Will Be Submitted by 990211.With 990106 Ltr
ML17335A4661999-01-0505 January 1999
LER 98-049-00:on 981020,emergency Boron Injection Flow Path Was Inoperable.Caused by Original Design Deficiency. Engineering Evaluation of Event Is Continuing
ML17335A4631999-01-0404 January 1999
LER 98-054-00:on 981202,discovered That at Least One MSSV Had Not Been Reset as Required by Ts.Engineering Is Continuing Review of Extent of Condition for Event.Updated LER Will Be Submitted by 990129.With 990104 Ltr
ML17335A4481998-12-30030 December 1998
LER 98-053-00:on 981130,discovered Use of Inoperable Substitute Subcooling Margin Monitor.Caused by Condition Existing Since Installation of Plant Process Computer in 1992.Updated LER Will Be Submitted.With 981230 Ltr
ML17335A4581998-12-28028 December 1998
LER 98-052-00:on 981128,turbine Driven AFW Pump Speed Controller Failure Mode Occurred.Caused Because Not All Failure Modes Were Considered When Compressed Air Sys Was Originally Designed.Verified Current Design Change Process
ML17335A4281998-12-22022 December 1998
LER 98-051-00:on 981122,reactor Trip Signal from Manual Safety Injection Not Verified as Required by TS Surveillance,Was Discovered.Maintenance Currently Evaluating Significance & Cause of Event
ML17335A4111998-12-17017 December 1998
LER 98-047-00:on 981117,potential for Increase Leakage from Reactor Coolant Pump Seals Was Identified.Util Is Working with W to Resolve Issue.Current Expectations Are to Submit Update to LER by 990215.With 981217 Ltr
ML17335A4141998-12-16016 December 1998
LER 98-058-00:on 981216,postulated High Line Break Could Result in Condition Outside Design Bases for AF Occurred. Caused by Deficiencies Associated with Administration of HELB Program.Analysis of AF Will Be Completed by 990122
ML17335A4181998-12-16016 December 1998
LER 98-050-00:on 980814,ancillary Equipment Installed in Ice Condenser Was Not Designed to Withstand Design Basis Accident/Earthquake Loads.Caused by Lack of Established Design Criteria.Developed Design Criteria
ML17335A3871998-12-11011 December 1998
LER 98-031-01:on 980610,potential Common Mode Failure of RHR Pumps Were Noted.Caused by Inaccurate Values.Accurate Miniflow Numbers Have Been Determined by Flow Testing
ML17335A3821998-12-0808 December 1998
LER 98-039-01 Re EOP Step Conflicts with Small Break LOCA Analysis.Ler 98-039-00 Has Been Canceled.With 981208 Ltr
ML17335A3781998-12-0707 December 1998
LER 98-007-00:on 981106,high Energy Line Break Effects in Auxiliary FW Sys Was Noted.Cause of Event Is Under Investigation & Will Be Completed by 990220.Updated LER Will Be Submitted by 990310.With 981207 Ltr
ML17335A3771998-12-0303 December 1998
LER 98-046-00:on 981103,determined That Afs Was Unable to Meet Design Flow Requirements During Special Test.Caused by Failure to Consider All Aspects of Sys Operation in Design of Suction Basket Strainers.Sys Will Be Redesigned
ML17335A3741998-12-0202 December 1998
LER 97-011-02:on 970822,operation Was Noted Outside Design Bases for ECCS & CSP for Switchover to Recirculation Sump Suction.Caused by Ineffective Change Mgt.Revised Procedure for Switchover 01(02) Ohp 4023.ES-1.3
1999-09-17
[Table view]
Category:RO)
MONTHYEARML17335A5641999-10-18018 October 1999
LER 99-024-00:on 990708,literal TS Requirements Were Not Met by Accumlator Valve Surveillance.Caused by Misjudgement Made in Conversion from Initial DC Cook TS to W Std Ts.Submitted License Amend Request.With 991018 Ltr
ML17335A5531999-10-0707 October 1999
LER 99-023-00:on 990907,inadequate TS Surveillance Testing of ESW Pump ESF Response Time Noted.Caused by Inadequate Understanding of Plant Design Basis.Surveillance Tests Will Be Revised & Implemented
ML17326A1291999-09-17017 September 1999
LER 99-022-00:on 990609,electrical Bus Degraded Voltage Setpoints Too Low for Safety Related Loads,Was Discovered. Caused by Lack of Understanding of Design of Plant.No Immediate Corrective Actions Necessary
ML17326A1121999-08-27027 August 1999
LER 99-021-00:on 990728,determined That GL 96-01 Test Requirements Were Not Met in Surveillance Tests.Caused by Failure to Understand Full Extent of GL Requirements. Surveillance Procedures Will Be Revised or Developed
ML17326A1011999-08-26026 August 1999
LER 99-020-00:on 990727,EDGs Were Declared Inoperable.Caused by Inadequate Protection of Air Intake,Exhaust & Room Ventilation Structures from Tornado Missile Hazards. Implemented Compensatory Measures in Form of ACs
ML17326A0911999-08-16016 August 1999
LER 99-019-00:on 990716,noted Victoreen Containment Hrrms Not Environmentally Qualified to Withstand post-LOCA Conditions.Caused by Inadequate Design Control.Reviewing Options to Support Hrrms Operability in Modes 1-4
ML17326A0771999-08-0404 August 1999
LER 98-029-01:on 980422,noted That Fuel Handling Area Ventilation Sys Was Inoperable.Caused by Original Design Deficiency.Radiological Analysis for Spent Fuel Handling Accidents in Auxiliary Bldg Will Be Redone by 990830
ML17326A0741999-07-29029 July 1999
LER 99-018-00:on 990629,determined That Valve Yokes May Yield Under Combined Stress of Seismic Event & Static,Valve Closed,Stem Thrust.Caused by Inadequate Design of Associated Movs.Operability Determinations Were Performed for Valves
ML17326A0661999-07-26026 July 1999
LER 99-017-00:on 990625,noted That Improperly Installed Fuel Oil Return Relief Valve Rendered EDG Inoperable.Caused by Personnel Error.Fuel Oil Return Valve Was Replaced with Valve in Correct Orientation.With 990722 Ltr
ML17326A0651999-07-22022 July 1999
LER 98-014-03:on 980310,noted That Response to high-high Containment Pressure Procedure Was Not Consistent with Analysis of Record.Caused by Inadequate Interface with W. FRZ-1 Will Be Revised to Be Consistent with New Analysis
ML17326A0491999-07-13013 July 1999
LER 99-016-00:on 990615,TS Requirements for Source Range Neutron Flux Monitors Not Met.Caused by Failure to Understand Design Basis of Plant.Procedures Revised.With 990713 Ltr
ML17326A0331999-07-0101 July 1999
LER 99-004-01:on 971030,failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed Was Noted.Caused by Ineffective Mgt of Tss.Chemistry Personnel Have Been Instructed on Requirement to Follow TS as Written
ML17326A0151999-06-18018 June 1999
LER 99-014-00:on 990521,determined That Boron Injection Tank Manway Bolts Were Not Included in ISI Program,Creating Missed Exam for Previous ISI Interval.Caused by Programmatic Weakness.Isi Program & Associated ISI Database Modified
ML17325B6311999-06-0101 June 1999
LER 99-S03-00:on 990430,vital Area Barrier Degradation Was Noted.Caused by Inadequate Insp & Maint of Vital Area Barrier.Repairs & Mods Were Made to Barriers to Eliminate Degraded & Nonconforming Conditions
ML17325B6421999-06-0101 June 1999
LER 99-013-00:on 990327,safety Injection & Centrifugal Charging Throttle Valve Cavitation During LOCA Could Have Led to ECCS Pump Failure.Caused by Inadequate Original Design Application of Si.Throttle Valves Will Be Developed
ML17325B6351999-05-28028 May 1999
LER 99-S02-00:on 990428,vulnerability in Safeguard Sys That Could Allow Unauthorized Access to Protected Area Was Noted. Caused by Inadequate Original Plant Design.Mods Were Made to Wall Opening to Eliminate Nonconforming Conditions
ML17265A8231999-05-24024 May 1999
LER 98-037-01:on 990422,determined That Ice Condenser Bypass Leakage Exceeds Design Basis Limit.Caused by Pressure Seal Required by Revised W Design Not Incorporated Into Aep Design.Numerous Matl Condition Walkdowns & Assessments Made
ML17325B6001999-05-20020 May 1999
LER 99-012-00:on 990420,concluded That Auxiliary Bldg ESF Ventilation Sys Not Capable of Maintaining ESF Room Temps post-accident.Caused by Inadequate Control of Sys Design Inputs.Comprehensive Action Plan Being Developed
ML17325B5861999-05-10010 May 1999
LER 99-002-00:on 990415,discovered That TS 4.0.5 Requirements Were Not Met Due to Improperly Performed Test. Caused by Incorrect Interpretation of ASME Code.App J Testing Will Be Completed & Procedures Will Be Revised
ML17325B5811999-05-0404 May 1999
LER 99-011-00:on 990407,air Sys for EDG Will Not Support Long Operability.Caused by Original Design Error.Temporary Mod to Supply Makeup Air Capability in Modes 5 & 6 Was Prepared
ML17325B5771999-05-0303 May 1999
LER 99-010-00:on 990401,RCS Leak Detection Sys Sensitivity Not in Accordance with Design Requirements Occurred.Caused by Inadequate Original Design of Containment Sump Level. Evaluation Will Be Performed to Clearly Define Design
ML17325B5581999-04-16016 April 1999
LER 99-006-00:on 990115,personnel Identified Discrepancy Between TS 3.9.7 Impact Energy Limit & Procedure 12 Ohp 4030.STP.046.Caused by Lack of Design Basis Control.Placed Procedure 12 Ohp 4030.STP.046 on Administrative Hold
ML17325B5471999-04-12012 April 1999
LER 99-009-00:on 990304,as-found RHR Safety Relief Valve Lift Setpoint Greater than TS Limit Occurred.Cause Investigation for Condition Has Not Been Completed.Update to LER Will Be Submitted,Upon Completion of Investigation
ML17325B5321999-04-0707 April 1999
LER 99-S01-00:on 990308,discovered That Lock for Vital Gate Leading to Plant 4KV Switchgear Area Was Nonconforming & Vulnerable to Unauthorized Access.Caused by Inadequate Gate Design & Inadequate Procedures.Mods Are Being Made to Gate
ML17325B5161999-04-0101 April 1999
LER 99-007-00:on 981020,calculations Showed That Divider Barrier Between Upper & Lower Containment Vols Were Overstressed.Engineers Are Currently Working on Analyses of Loads & Stress on Enclosures
ML17325B5221999-03-29029 March 1999
LER 99-001-00:on 960610,degraded Component Cooling Water Flow to Containment Main Steam Line Penetrations,Identified on 990226.Caused by Inadequate Understanding of Design Basis.Additional Investigations Ongoing
ML17325B4801999-03-18018 March 1999
LER 99-004-00:on 971030,failure to Perform TS Surveillance Analyses of Rc Chemistry with Fuel Removed Was Noted.Cause of Event Is Under Investigation.Corrected Written Job Order Activities Used to Control SD Chemistry Sampling
ML17325B4741999-03-18018 March 1999
LER 99-005-00:on 940512,determined That Rt Breaker Manual Actuations During Rod Drop Testing Were Not Previously Reported.Caused by Lack of Training.Addl Corrective Actions,Including Preventative Actions May Be Developed
ML17325B4571999-02-24024 February 1999
LER 99-003-00:on 990107,CR Pressurization Sys Surveillance Test Did Not Test Sys in Normal Operating Condition.Caused by Failure to Recognize Door 12DR-AUX415 as Part of CR Pressure Boundary.Performed Walkdown of Other Doors
ML17335A5171999-02-11011 February 1999
LER 99-002-00:on 990112,determined That RCS Pressurizer PORVs Had Not Been Tested,Per Ts.Caused by Inadequate Scheduling Controls Allowing Personnel Error.Surveillance Procedure Was Completed & Updated LER Will Be Submitted
ML17335A5141999-02-10010 February 1999
LER 99-001-00:on 990106,noted That GE Hfa Relays Installed in EDGs May Not Meet Seismic Qualification.Caused by Operating Experience Info Incorrectly Dispositioned in 1985. Updated LER Will Be Submitted by 990405
ML17335A5011999-02-0101 February 1999
LER 98-060-00:on 981231,identified That Rt Sys Response Time Testing Did Not Comply with TS Definition.Caused by Inadequate Procedures.Corrective Actions Will Be Developed & Update to LER Will Be Submitted by 990415.With 990201 Ltr
ML17335A4951999-01-29029 January 1999
LER 98-059-00:on 981230,interim LER -single Failure in Containment Spray Sys Could Result in Containment Spray Ph Outside Design Occurred.Investigation Into Condition Continuing.Update Will Be Submitted by 990514 Ltr
ML17335A4961999-01-27027 January 1999
LER 98-057-00:on 981228,discovered That AFW Valves Were Not Tested IAW Inservice Testing Program.Caused by Failure to Recognize Design Bases Features Re Afws by Personnel. Updated LER Will Be Submittted by 990415.With 990127 Ltr
ML17335A4921999-01-19019 January 1999
LER 98-052-01:on 981128,no Analysis for NSR Sc Manual Loader for Tdafwp Could Be Found in Original Design.Cause Due to All Failure Modes Not Considered When Compressed Air Sys Originally Designed.Performed Review of Components
ML17335A4721999-01-0606 January 1999
LER 98-055-00:on 981207,potential for Condition Outside of Design Bases for Rod Control Sys Was Noted.Caused by Calibration Error Coupled with Single Rod Failure.Condition Rept Investigation Is Ongoing
ML17335A4691999-01-0606 January 1999
LER 98-056-00:on 981211,hot Leg Nozzle Gaps Resulted in Plant Being in Unanalyzed Condition.Analyses Are Being Performed by W to Resolve Problem.Updated LER Will Be Submitted by 990211.With 990106 Ltr
ML17335A4661999-01-0505 January 1999
LER 98-049-00:on 981020,emergency Boron Injection Flow Path Was Inoperable.Caused by Original Design Deficiency. Engineering Evaluation of Event Is Continuing
ML17335A4631999-01-0404 January 1999
LER 98-054-00:on 981202,discovered That at Least One MSSV Had Not Been Reset as Required by Ts.Engineering Is Continuing Review of Extent of Condition for Event.Updated LER Will Be Submitted by 990129.With 990104 Ltr
ML17335A4481998-12-30030 December 1998
LER 98-053-00:on 981130,discovered Use of Inoperable Substitute Subcooling Margin Monitor.Caused by Condition Existing Since Installation of Plant Process Computer in 1992.Updated LER Will Be Submitted.With 981230 Ltr
ML17335A4581998-12-28028 December 1998
LER 98-052-00:on 981128,turbine Driven AFW Pump Speed Controller Failure Mode Occurred.Caused Because Not All Failure Modes Were Considered When Compressed Air Sys Was Originally Designed.Verified Current Design Change Process
ML17335A4281998-12-22022 December 1998
LER 98-051-00:on 981122,reactor Trip Signal from Manual Safety Injection Not Verified as Required by TS Surveillance,Was Discovered.Maintenance Currently Evaluating Significance & Cause of Event
ML17335A4111998-12-17017 December 1998
LER 98-047-00:on 981117,potential for Increase Leakage from Reactor Coolant Pump Seals Was Identified.Util Is Working with W to Resolve Issue.Current Expectations Are to Submit Update to LER by 990215.With 981217 Ltr
ML17335A4141998-12-16016 December 1998
LER 98-058-00:on 981216,postulated High Line Break Could Result in Condition Outside Design Bases for AF Occurred. Caused by Deficiencies Associated with Administration of HELB Program.Analysis of AF Will Be Completed by 990122
ML17335A4181998-12-16016 December 1998
LER 98-050-00:on 980814,ancillary Equipment Installed in Ice Condenser Was Not Designed to Withstand Design Basis Accident/Earthquake Loads.Caused by Lack of Established Design Criteria.Developed Design Criteria
ML17335A3871998-12-11011 December 1998
LER 98-031-01:on 980610,potential Common Mode Failure of RHR Pumps Were Noted.Caused by Inaccurate Values.Accurate Miniflow Numbers Have Been Determined by Flow Testing
ML17335A3821998-12-0808 December 1998
LER 98-039-01 Re EOP Step Conflicts with Small Break LOCA Analysis.Ler 98-039-00 Has Been Canceled.With 981208 Ltr
ML17335A3781998-12-0707 December 1998
LER 98-007-00:on 981106,high Energy Line Break Effects in Auxiliary FW Sys Was Noted.Cause of Event Is Under Investigation & Will Be Completed by 990220.Updated LER Will Be Submitted by 990310.With 981207 Ltr
ML17335A3771998-12-0303 December 1998
LER 98-046-00:on 981103,determined That Afs Was Unable to Meet Design Flow Requirements During Special Test.Caused by Failure to Consider All Aspects of Sys Operation in Design of Suction Basket Strainers.Sys Will Be Redesigned
ML17335A3741998-12-0202 December 1998
LER 97-011-02:on 970822,operation Was Noted Outside Design Bases for ECCS & CSP for Switchover to Recirculation Sump Suction.Caused by Ineffective Change Mgt.Revised Procedure for Switchover 01(02) Ohp 4023.ES-1.3
1999-09-17
[Table view]
Category:TEXT-SAFETY REPORT
MONTHYEARML17335A5641999-10-18018 October 1999
LER 99-024-00:on 990708,literal TS Requirements Were Not Met by Accumlator Valve Surveillance.Caused by Misjudgement Made in Conversion from Initial DC Cook TS to W Std Ts.Submitted License Amend Request.With 991018 Ltr
ML17335A5531999-10-0707 October 1999
LER 99-023-00:on 990907,inadequate TS Surveillance Testing of ESW Pump ESF Response Time Noted.Caused by Inadequate Understanding of Plant Design Basis.Surveillance Tests Will Be Revised & Implemented
ML17335A5631999-09-30030 September 1999
Monthly Operating Rept for Sept 1999 for DC Cook Nuclear Plant,Unit 1.With 991012 Ltr
ML17335A5621999-09-30030 September 1999
Monthly Operating Rept for Sept 1999 for DC Cook Nuclear Plant,Unit 2.With 991012 Ltr
ML17335A5481999-09-30030 September 1999
Non-proprietary DC Cook Nuclear Plant Units 1 & 2 Mods to Containment Sys W SE (Secl 99-076,Rev 3).
ML17335A5451999-09-28028 September 1999
Rev 1 to Containment Sump Level Design Condition & Failure Effects Analysis for Potential Draindown Scenarios.
ML17326A1291999-09-17017 September 1999
LER 99-022-00:on 990609,electrical Bus Degraded Voltage Setpoints Too Low for Safety Related Loads,Was Discovered. Caused by Lack of Understanding of Design of Plant.No Immediate Corrective Actions Necessary
ML17326A1481999-09-17017 September 1999
Independent Review of Control Rod Insertion Following Cold Leg Lbloca,Dc Cook,Units 1 & 2.
ML17326A1211999-08-31031 August 1999
Monthly Operating Rept for Aug 1999 for Cook Nuclear Plant, Unit 2.With 990915 Ltr
ML17326A1201999-08-31031 August 1999
Monthly Operating Rept for Aug 1999 for Cook Nuclear Plant, Unit 1.With 990915 Ltr
ML17326A1121999-08-27027 August 1999
LER 99-021-00:on 990728,determined That GL 96-01 Test Requirements Were Not Met in Surveillance Tests.Caused by Failure to Understand Full Extent of GL Requirements. Surveillance Procedures Will Be Revised or Developed
ML17326A1011999-08-26026 August 1999
LER 99-020-00:on 990727,EDGs Were Declared Inoperable.Caused by Inadequate Protection of Air Intake,Exhaust & Room Ventilation Structures from Tornado Missile Hazards. Implemented Compensatory Measures in Form of ACs
ML17326A0911999-08-16016 August 1999
LER 99-019-00:on 990716,noted Victoreen Containment Hrrms Not Environmentally Qualified to Withstand post-LOCA Conditions.Caused by Inadequate Design Control.Reviewing Options to Support Hrrms Operability in Modes 1-4
ML17326A0771999-08-0404 August 1999
LER 98-029-01:on 980422,noted That Fuel Handling Area Ventilation Sys Was Inoperable.Caused by Original Design Deficiency.Radiological Analysis for Spent Fuel Handling Accidents in Auxiliary Bldg Will Be Redone by 990830
ML17335A5461999-08-0202 August 1999
Rev 0 to Evaluation of Cook Recirculation Sump Level for Reduced Pump Flow Rates.
ML17326A0871999-07-31031 July 1999
Monthly Operating Rept for July 1999 for DC Cook Nuclear Plant,Unit 1.With 990812 Ltr
ML17326A0861999-07-31031 July 1999
Monthly Operating Rept for July 1999 for DC Cook Nuclear Plant,Units 2.With 990812 Ltr
ML17326A0741999-07-29029 July 1999
LER 99-018-00:on 990629,determined That Valve Yokes May Yield Under Combined Stress of Seismic Event & Static,Valve Closed,Stem Thrust.Caused by Inadequate Design of Associated Movs.Operability Determinations Were Performed for Valves
ML17326A0661999-07-26026 July 1999
LER 99-017-00:on 990625,noted That Improperly Installed Fuel Oil Return Relief Valve Rendered EDG Inoperable.Caused by Personnel Error.Fuel Oil Return Valve Was Replaced with Valve in Correct Orientation.With 990722 Ltr
ML17326A0651999-07-22022 July 1999
LER 98-014-03:on 980310,noted That Response to high-high Containment Pressure Procedure Was Not Consistent with Analysis of Record.Caused by Inadequate Interface with W. FRZ-1 Will Be Revised to Be Consistent with New Analysis
ML17326A0491999-07-13013 July 1999
LER 99-016-00:on 990615,TS Requirements for Source Range Neutron Flux Monitors Not Met.Caused by Failure to Understand Design Basis of Plant.Procedures Revised.With 990713 Ltr
ML17326A0331999-07-0101 July 1999
LER 99-004-01:on 971030,failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed Was Noted.Caused by Ineffective Mgt of Tss.Chemistry Personnel Have Been Instructed on Requirement to Follow TS as Written
ML17326A0511999-06-30030 June 1999
Monthly Operating Rept for June 1999 for DC Cook Nuclear Plant,Unit 2.With 990709 Ltr
ML17326A0501999-06-30030 June 1999
Monthly Operating Rept for June 1999 for DC Cook Nuclear Plant,Unit 1.With 990709 Ltr
ML17326A0151999-06-18018 June 1999
LER 99-014-00:on 990521,determined That Boron Injection Tank Manway Bolts Were Not Included in ISI Program,Creating Missed Exam for Previous ISI Interval.Caused by Programmatic Weakness.Isi Program & Associated ISI Database Modified
ML17325B6421999-06-0101 June 1999
LER 99-013-00:on 990327,safety Injection & Centrifugal Charging Throttle Valve Cavitation During LOCA Could Have Led to ECCS Pump Failure.Caused by Inadequate Original Design Application of Si.Throttle Valves Will Be Developed
ML17325B6311999-06-0101 June 1999
LER 99-S03-00:on 990430,vital Area Barrier Degradation Was Noted.Caused by Inadequate Insp & Maint of Vital Area Barrier.Repairs & Mods Were Made to Barriers to Eliminate Degraded & Nonconforming Conditions
ML17326A0061999-05-31031 May 1999
Monthly Operating Rept for May 1999 for Dcp.With 990609 Ltr
ML17326A0071999-05-31031 May 1999
Monthly Operating Rept for May 1999 for DC Cook Nuclear Plant,Unit 2.With 990609 Ltr
ML17325B6351999-05-28028 May 1999
LER 99-S02-00:on 990428,vulnerability in Safeguard Sys That Could Allow Unauthorized Access to Protected Area Was Noted. Caused by Inadequate Original Plant Design.Mods Were Made to Wall Opening to Eliminate Nonconforming Conditions
ML17265A8231999-05-24024 May 1999
LER 98-037-01:on 990422,determined That Ice Condenser Bypass Leakage Exceeds Design Basis Limit.Caused by Pressure Seal Required by Revised W Design Not Incorporated Into Aep Design.Numerous Matl Condition Walkdowns & Assessments Made
ML17325B6001999-05-20020 May 1999
LER 99-012-00:on 990420,concluded That Auxiliary Bldg ESF Ventilation Sys Not Capable of Maintaining ESF Room Temps post-accident.Caused by Inadequate Control of Sys Design Inputs.Comprehensive Action Plan Being Developed
ML17325B5861999-05-10010 May 1999
LER 99-002-00:on 990415,discovered That TS 4.0.5 Requirements Were Not Met Due to Improperly Performed Test. Caused by Incorrect Interpretation of ASME Code.App J Testing Will Be Completed & Procedures Will Be Revised
ML17325B5811999-05-0404 May 1999
LER 99-011-00:on 990407,air Sys for EDG Will Not Support Long Operability.Caused by Original Design Error.Temporary Mod to Supply Makeup Air Capability in Modes 5 & 6 Was Prepared
ML17325B5771999-05-0303 May 1999
LER 99-010-00:on 990401,RCS Leak Detection Sys Sensitivity Not in Accordance with Design Requirements Occurred.Caused by Inadequate Original Design of Containment Sump Level. Evaluation Will Be Performed to Clearly Define Design
ML17335A5301999-04-30030 April 1999
Monthly Operating Rept for Apr 1999 for DC Cook Nuclear Plant,Unit 1.With 990508 Ltr
ML17335A5291999-04-30030 April 1999
Monthly Operating Rept for Apr 1999 for DC Cook Nuclear Plant,Unit 2.With 990508 Ltr
ML17325B5581999-04-16016 April 1999
LER 99-006-00:on 990115,personnel Identified Discrepancy Between TS 3.9.7 Impact Energy Limit & Procedure 12 Ohp 4030.STP.046.Caused by Lack of Design Basis Control.Placed Procedure 12 Ohp 4030.STP.046 on Administrative Hold
ML17325B5471999-04-12012 April 1999
LER 99-009-00:on 990304,as-found RHR Safety Relief Valve Lift Setpoint Greater than TS Limit Occurred.Cause Investigation for Condition Has Not Been Completed.Update to LER Will Be Submitted,Upon Completion of Investigation
ML17325B5321999-04-0707 April 1999
LER 99-S01-00:on 990308,discovered That Lock for Vital Gate Leading to Plant 4KV Switchgear Area Was Nonconforming & Vulnerable to Unauthorized Access.Caused by Inadequate Gate Design & Inadequate Procedures.Mods Are Being Made to Gate
ML17325B5161999-04-0101 April 1999
LER 99-007-00:on 981020,calculations Showed That Divider Barrier Between Upper & Lower Containment Vols Were Overstressed.Engineers Are Currently Working on Analyses of Loads & Stress on Enclosures
ML17325B5491999-03-31031 March 1999
Monthly Operating Rept for Mar 1999 for DC Cook Nuclear Plant Unit 2.With 990408 Ltr
ML17325B5441999-03-31031 March 1999
Monthly Operating Rept for Mar 1999 for DC Cook Nuclear Plant,Unit 1.With 990408 Ltr
ML17325B5221999-03-29029 March 1999
LER 99-001-00:on 960610,degraded Component Cooling Water Flow to Containment Main Steam Line Penetrations,Identified on 990226.Caused by Inadequate Understanding of Design Basis.Additional Investigations Ongoing
ML17325B4801999-03-18018 March 1999
LER 99-004-00:on 971030,failure to Perform TS Surveillance Analyses of Rc Chemistry with Fuel Removed Was Noted.Cause of Event Is Under Investigation.Corrected Written Job Order Activities Used to Control SD Chemistry Sampling
ML17325B4741999-03-18018 March 1999
LER 99-005-00:on 940512,determined That Rt Breaker Manual Actuations During Rod Drop Testing Were Not Previously Reported.Caused by Lack of Training.Addl Corrective Actions,Including Preventative Actions May Be Developed
ML17325B5671999-03-0202 March 1999
Summary of Unit 1 Steam Generator Layup Chemistry from 980101 to 990218.
ML17325B4631999-02-28028 February 1999
Monthly Operating Rept for Feb 1999 for DC Cook Nuclear Power Station,Unit 2.With 990308 Ltr
ML17325B4621999-02-28028 February 1999
Monthly Operating Rept for Feb 1999 for DC Cook Nuclear Plant,Unit 1.With 990308 Ltr
ML17325B4571999-02-24024 February 1999
LER 99-003-00:on 990107,CR Pressurization Sys Surveillance Test Did Not Test Sys in Normal Operating Condition.Caused by Failure to Recognize Door 12DR-AUX415 as Part of CR Pressure Boundary.Performed Walkdown of Other Doors
1999-09-30
[Table view] |
Text
ACCELERATED ITRIBUTION DEMONS'ATION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)ACCESSION NBR:9405190369 DOC.DATE: 94/05/10 NOTARIZED:
NO DOCKET g FACIL:50-315 Donald C.Cook Nuclear Power Plant, Unit 1, Indiana M 05000315 AUTH.NAME AUTHOR AFFILIATION WEBER,G.A.
Indiana Michigan Power Co.(formerly Indiana&Michigan Ele BLIND,A.A.
Indiana Michigan Power Co.(formerly Indiana&Michigan Ele RECIP.NAME RECIPIENT AFFILIATION
SUBJECT:
LER 94-004-00:on 940406,three pressurizer safety valves, sent to to off site test lab for testing failed to meet TS acceptance criteria.Cause not determined.
Nozzle&disc seating surfaces lapped&polished.W/940510 ltr.DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR (ENCL 0 SIZE: TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.NOTES: RECIPIENT ID CODE/NAME PD3-1 PD INTERNAL: AEOD/DOA AEOD/ROAB/DS P NRR/DE/EMEB NRR/DRCH/HHFB NRR/DRCH/HOLB NRR/DRSS/PRPB NRR/DSSA/SRXB RES/DSIR/EIB EXTERNAL: EG&G BRYCE,J.H NRC PDR NSIC POORE,W.COPIES LTTR ENCL 1 1 1 1 2 2 1 1 1 1 1 1 2 2 1 1 1 1 2 2'1 1 1 1 RECIPIENT ID CODE/NAME HICKMAN,J AEOD/DSP/TPAB NRR/DE/EELB NRR/DORS/OEAB NRR/DRCH/H I CB NRR/DRI L/RPEB NRR DSSA/SPLB GREG 02 FILE 01 L ST LOBBY WARD NSIC MURPHY,G.A NUDOCS FULL TXT COPIES LTTR ENCL 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 NOTE TO ALL"RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE iVASTE!CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT.20079)TO ELIMINATE YOUR NAME FROiVi DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED'TTR 27 ENCL 27 Indiana Michig Power Company Cook Nuclear Plant One Cook Place Bridgrnan, Ml 49106 616 465 5901 INDlANA NICHlGi4N POWE'R May 10, 1994 United States Nuclear Regulatory Commission Document Control Desk Rockville, Maryland 20852 Operating Licenses DPR-58 Docket No.50-315 Document Control Manager: In accordance with the criteria established by 10 CFR 50.73 entitled Licensee Event Re ort S stem the following report is being submitted:
94-004-00 A.A.Blind Plant Manager/sb Attachment c'.B.Martin, Region III E.E.Fitzpatrick P.A.Barrett R.F.Kroeger M.A.Bailey-Ft.Wayne NRC Resident Inspector J.B.Hickman-NRC J.R.Padgett G.Charnoff, Esq.D.Hahn INPO S.J.Brewer q n.~, i~U l,s<I'-v'7405190369 940510 PDR ADOCK 05000315 S PDR NRC FORM 366 (5 92)U.S.NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO.3150.0104 EXP IR ES 5/31/95 LICENSEE EVENT REPORT (LER)(See reverse for required number of digits/characters for each block)ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMADON COLLECTION REQUEST: 50.0 HRS.FORWARD COMMENTS REGARDUIG BURDEN ESDMATE TO THE INFQRMATICN AND RECORDS MANAGEMENT BRANCH (MNBB Tr ta), U.S.NUCLEAR REGULATORY COMMISSION>
WASHINGTON, DC 20555 0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150010a), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503, FACILITY NAME (I)D.C.COOK NUCLEAR PLANT-UNIT 1 DOCKET NUMBER (2)05000 315 PAGE P)10F 3 FAILURE OF THREE PRESSURIZER SAFETY VALVES TO'MEET TECHNICAL SPECIFICATION RVEILLANCE TEST CRITERIA EVENT DATE 5 LER NUMBER 6 REPORT NUMBER 7 OTHER FACILITIES INVOLVED 6 MONTH 04 DAY 06 YEAR 94 YEAR 94 SEQUENTIAL NUMBER 004 REVISION NUMBER 0 MONTH DAY 05 10 94 FACIUTY NAME FACILITY NAME DOCKET NUMBER 05000 DOCKET NUMBcR 05000 OPERATING MODE (9)20A02(b)20.405(c)50.73(a)(2)(iv)73.71(b)NT TO THE REQUIREMENTS OF 10 CFR E: Check one or m THIS REPORT IS SUBMITTED PURSUA ore 11 POWER LEVEL (10)0 20.405(a)(1)(i)20.405(a)(1)(ii)20.405(a)(1)(iii)20.405(a)(1)(iv)2o.405(a)(1)(v) 50.36(c)(1)50.36(c)(2)X 50.73(a)(2)(i) 50.73(a)(2)(ii)50.73(a)(2)(iii)LICENSEE CONTACT FOR THIS LER 12 M73(a)(2)(v) 50.73(a)(2)(vii)50.73(a)(2)(viii)(A)50.73(a)(2)(viii)(B) 50.73(a)(2)(x) 73.71(c)OTHER (Speoty in Atrctract tretow and in Text, NRC Form 366A)NAME G.A.WEBER-PLANT ENGINEERING SUPERINTENDENT TELEPHONE NUMBER Fnctvde Area code)(616)465-5902 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 13 CAUSE SYSTEM COMPONENT RV MANUFACTURER C710 REPORTABLE TO NPRDS , D.;j.x',~CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NPRDS SUPPLEMENTAL REPORT EXPECTED 14 YES 0(yee, cornptete ExPEGTED sUBMIssIQN DATE)NO ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single.spaced typewritten lines)(16)EXPFCTFD MONTH DAY YEAR SUBMISSION DATE (15)On Apri.l 6, 1994 with Uni.t 1 in Mode 5 (Cold Shutdown)it was determined that all three of the pressurizer safety valves, which were sent to an off site test laboratory for testing, were found with lift settings outside of the Technical Specification acceptance criteria.Acceptable settings are between, 2461 psig and 2509 psig.Valve 1-SV-45A was found to have a lift setpoint of 2536 psig, valve 1-SV-45B had a lift setpoint of 2535 psig and 1-SV-45C had a lift setpoint of 2538 psig.There was no safety-signi.ficance since the worst case (1-SV-45C-lift setpoint of 2538 psig)would result in a maximum transient pressure of 2615 psig (2538 psig plus 3 percent accumulation to attain its full rated lift).This is below the Technical Specification safety limit of 2735 psig.All three valves were partially disassembled (retaining spring compression) and inspected.
No problems were noted.The nozzle and disc seating surfaces were lapped and polished.The valves were reassembled and tested satisfactori.ly.
The safety valve conditions experienced at the D.C.Cook Plant are similar to current i.ndustry trends/concerns.
-Since a specific Root Cause could not be determined, no preventive action i.s planned at this time.However, we will be evaluating the test methods and industry activities pertaining to the pressurizer safety valves.NRC FORM 366 (5-92)
REQUIRED NUMBER OF DIGITS/CHARACTERS FOR EACH BLOCK BLOCK NUMBER NUMBER OF DIGITS/CHARACTERS UP TO 46 8 TOTAL 3 IN ADDITION TO 05000 VARIES TITLE FACILITY NAME DOCKET NUMBER PAGE NUMBER 10 12 13 14 15 UP TO 76 6 TOTAL 2 PER BLOCK 7 TOTAL 2 FOR YEAR 3 FOR SEQUENTIAL NUMBER 2 FOR REVISION NUMBER 6 TOTAL 2 PER BLOCK UP TO 18*.FACILITY NAME 8 TOTAL-DOCKET NUMBER 3 IN ADDITION TO 05000 1 CHECK BOX THAT APPLIES UP TO 50 FOR NAME 14 FOR TELEPHONE CAUSE VARIES 2 FOR SYSTEM 4 FOR COMPONENT 4 FOR MANUFACTURER NPRDS VARIES 1 CHECK BOX THAT APPLIES 6 TOTAL 2 PER BLOCK TITLE EVENT DATE LER NUMBER REPORT DATE OTHER FACILITIES INVOLVED OPERATING MODE POWER LEVEL REQUIREMENTS OF 10 CFR LICENSEE CONTACT EACH COMPONENT FAILURE SUPPLEMENTAL REPORT EXPECTED EXPECTED SUBMISSION DATE NRC FORM 366A r (669)U.S.NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION APPROVED OMB NO.31500104 EXPIRES: 4/30I92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUESTI 50.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (F630), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, OC 20555, AND TO 1HE PAPERWORK REDUCTION PRO)ECT (31500104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.FACILITY NAME (1)DOCKE'T NUMBER (2)LER NUMBER (6)PAGE (3)YEAR Hr5 SEOUENTIAL NUMBER REVISION NUMSFR D.C.COOK NUCLEAR PLANT-UNIT 1 TEXT (if more 4Pece Je eetrr9ed, Iree eddttr'one)
NRC Forrrr 3IJSA'et (12)o 5 o o o 315 94 0 0 4-0 0 0 2 OF 0 3 Conditions Prior to Occurrence:
Unit One-Mode 5 (Cold Shutdown-following refueling).
Descri tion of Event: On April 6, 1994, it was determined that all three safety pressurizer safety valves, Crosby Valve Model HB-86-PB, (EZIS/AB-RV) had lift settings outside Technical Specification
3.4.3 acceptance
criteria.The safety valves are tested at a test laboratory using steam at nominal temperature and pressure, as required by Technical Specification.
The valves are required to lift at 2485 psig plus or minus 1 percent, (i.e.between 2461 and 2509 psig).Valve 1-SV-45A was found to have a lift setpoint of 2536 psig, valve 1-SV-45B had a lift setpoint of 2535 psig and 1-SV-45C had a lift setpoint of 2538 psig.Technical Specification 4.4:3 requires that each Pressurizer Code Safety Valve be demonstrated operable per Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition.Cause of Event: The safety valve conditions experienced at the D.C.Cook Plant are similar to current industry trends/concerns.
The phenomenon of safety valve setpoint drift outside of design tolerances is common in the Nuclear industry.However, as yet, no cause for the drift has been determined.
1-SV-45A, 1-SV-45B and 1-SV-45C were partially disassembled (retaining spring compression) and inspected.
No problems were noted.The nozzle and disc seating surfaces were lapped and polished.The valves were reassembled and tested satisfactorily.
The test program (test facility and procedure) for the Pressurizer Safety Valves has not changed from previous testing.Anal sis of Event: This event has been determined to be reportable under the provisions of 10CFR5073(a)(2)(i)(B) as an operation prohibited by Plant Technical Specification 3.4.3, which requires all of the pressurizer safety valves to be operable with a lift setting of 2485 psig+/-1 percent.The as-found lift setpoints of safety valves 1-SV-45A, 1-SV-45B and 1-SV-45C did not have any actual impact on the Reactor Coolant System (RCS)since the safety valves were not challenged during the last fuel cycle.There was no potential impact since the RCS would not have exceeded the maximum transient limit of 2735 psig, which is 110 percent of design pressure (2485 psig).There was no impact on the health or safety of the public.Safety Valve 1-SV-45C (worst case)had a lift setpoint of 2538 psig.The RCS pressure would have to reach a pressure of 2615 psig (2538 psig plus 3 percent accumulation) for this valve to attain its full rated lift.Valve 1-SV-45A would have attained its rated lift at 2612 psig (2536 psig plus 3 percent)and 1-SV-45B would have attained its rated lift at 2612 psig (2535 plus 3 percent).NRC Form 366A (669)
NRC FORM 366A (64)9)U.S.NUCLEAR REGULATORY COMMISSIO LICENSEE EVENT REPORT (LER)TEXT CONTINUATION APPROVED OMB NO.31504))04 EXPIRES)4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REGUESTI 500 HRS.FORWARD COMMENTS REGARDING BUADEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH IP430), U.S.NUCLEAR AEGULATOAY COMMISSION, WASHINGTON, OC 20555, AND TO THE PAPERWORK REDUCTION PAOJECT (31500104), OFFICE OF MANAGF MENT AND BUDGET, WASHINGTON, OC 20503.FACILITY NAME (1)DOCKET NUMBER (21 LFR NUMBER (6)PAGE (3)YEAR)'.yi(SSOVSNTIAL NVMSSR REVISION NVM 54 D.C.COOK NUCLEAR PLANT-UNIT 1 TEXT///mttt tptct B nqvnd.IItt tddldoIM//Y/IC Fdmt 3/)BA'4/(I 2)0 5 0 0 0 3 I 5 9 4 0 0 4 00 03 oF 0 Anal sis of Event continued!
The reactor vessel and pressurizer were designed to ASME BGPV Section III which permits a maximum transient.
pressure of 2735 psig, 110 percent of design pressure (2485 psig).The RCS piping, valves and fittings are designed to ANSI B31.1, 1967 Edition, which permits a maximum transient pressure of 2985 psig, 120 percent of design pressure (2485 psig).In addition, the entire RCS was hydro tested to 3107 psig, 125 percent of design (2485 psig), to demonstrate system integrity prior to initial operation.
In conclusion, this event did not have any safety significance and did not represent a hazard to the public health and safety.The safety limit of 2735 psig would not have been exceeded since the maximum RCS pressure would not have exceed 2615 psig (1-SV-45C setpoint of 2538 psig plus 3 percent).In addition to this Safety Analysis, an additional Safety Evaluation is being performed to review the combined effect of recent Main Steam Safety Valve lift setpoints (reported in LER 50-315/94-001 and 94-003)in conjunction with the as-found lift setpoints of the Pressurizer Safety Valves.This evaluation is scheduled to be completed by June 30, 1994.We do not anticipate any adverse conditions to be identified during this evaluation, however, an updated LER will be submitted if deemed necessary.
Corrective Action: The nozzle and disc seating surfaces were lapped and polished for all three safety valves.Retests were satisfactorily completed for steam set pressure and seat leakage.The spring pressure on 1-SV-45A was adjusted by 1/3 flat to bring the set pressure back into tolerance.
Since a specific Root Cause could not be determined, no preventive action is planned at this time.However, we will continue to follow industry activities pertaining to safety valve setpoint drift.Failed Com onent Identification:
Pressurizer Safety Valve Plant Designation:
1-SV-45A, 1-SV-45B and 1-SV-45C Manufacturer:
Crosby Valve Company Model: HB-86-BP EIIS Code: AB-RV Previous Similar Events: LERS 50-315/90-16, 92-09 LER: 50-316/89-04, 92-06 NRC Form 366A (669)
