|
|
| Line 16: |
Line 16: |
|
| |
|
| =Text= | | =Text= |
| {{#Wiki_filter:ACCELERATED DISJBUTION DEMONS~ION SYSTEMREGULATORY INFORMATION DISTRIBUTION SYSTEM(RIDS)ACCESSION NBR:9005160332 DOC.DATE: | | {{#Wiki_filter:ACCELERATED DISJBUTION DEMONS~ION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)ACCESSION NBR:9005160332 DOC.DATE: 90/05/10 NOTARIZED: |
| 90/05/10NOTARIZED: | | NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH.NAME AUTHOR AFFILIATION GORSKI,P.Rochester Gas&Electric Corp.MECREDY,R.C. |
| NODOCKETFACIL:50-244 RobertEmmetGinnaNuclearPlant,Unit1,Rochester G05000244AUTH.NAMEAUTHORAFFILIATION GORSKI,P.
| | Rochester Gas&Electric Corp.RECIP.NAME RECIPIENT AFFILIATION |
| Rochester Gas&ElectricCorp.MECREDY,R.C. | |
| Rochester Gas&ElectricCorp.RECIP.NAME RECIPIENT AFFILIATION | |
|
| |
|
| ==SUBJECT:== | | ==SUBJECT:== |
| LER90-004-00:on 900416,steam generator tubedegradation duetoIGA/SCCcausesQAmanualreportable limitstobereached.W/9ltr.DISTRIBUTION CODE:IE22TCOPIESRECEIVED:LTR 2ENCLLSIZE:TITLE:50.73/50.9 LicenseeEventReport(LER),IncidentRpt,etc.NOTES:License Expdateinaccordance with10CFR2,2.109(9/19/72).
| | LER 90-004-00:on 900416,steam generator tube degradation due to IGA/SCC causes QA manual reportable limits to be reached.W/9 ltr.DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR 2 ENCL L SIZE: TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). |
| 05000244RECIPIENT IDCODE/NAME PDl-3LAJOHNSON,A INTERNAL:
| | 05000244 RECIPIENT ID CODE/NAME PDl-3 LA JOHNSON,A INTERNAL: ACNW AEOD/DSP/TPAB DEDRO NRR/DET/EMEB9H3 NRR/DLPQ/LPEB10 NRR/DREP/PRPB11 NRR/DST/SICB 7E NRR/DST/SRXB SE RES/DSIR/EIB EXTERNAL: EG&G STUART,V.A LPDR NSIC MAYS,G NUDOCS FULL TXT COPIES LTTR ENCL 1" 1 1 1 2 2 1 1 1 1 1 1 1 1 2 2 1 1 1.1 1 1 4 4 1 1 1 1 1 1 RECIPIENT ID CODE/NAME PD1-3 PD AEOD/DOA AEOD/ROAB/DSP NRR/DET/ECMB 9H NRR/DLPQ/LHFB11 NRR/DOEA/OEAB11 NRR/DST/SELB SD NRRQBSQ'LBSD1 |
| ACNWAEOD/DSP/TPAB DEDRONRR/DET/EMEB9H3 NRR/DLPQ/LPEB10 NRR/DREP/PRPB11 NRR/DST/SICB 7ENRR/DST/SRXB SERES/DSIR/EIB EXTERNAL:
| | , EG FILE 02 R E 01 L ST LOBBY WARD NRC PDR NSIC MURPHYiG A COPIES LTTR ENCL 1 1 ,1 1'2 2 1 1 1 1 1 1~1 1 1, 1 1 1 1 1 1 1 1 1 1 1 P>govs 7MB 1 NOTE TO ALL"RIDS" RECIPIENTS: |
| EG&GSTUART,V.A LPDRNSICMAYS,GNUDOCSFULLTXTCOPIESLTTRENCL1"111221111111122111.11144111111RECIPIENT IDCODE/NAME PD1-3PDAEOD/DOAAEOD/ROAB/DSP NRR/DET/ECMB 9HNRR/DLPQ/LHFB11 NRR/DOEA/OEAB11 NRR/DST/SELB SDNRRQBSQ'LBSD1 | | PLEASE HELP US TO REDUCE iVASTE!CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT.20079)TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 34 ENCL 34 |
| ,EGFILE02RE01LSTLOBBYWARDNRCPDRNSICMURPHYiGACOPIESLTTRENCL11,11'22111111~111,11111111111P>govs7MB1NOTETOALL"RIDS"RECIPIENTS: | |
| PLEASEHELPUSTOREDUCEiVASTE!CONTACTTHEDOCUMENTCONTROLDESK,ROOMPl-37(EXT.20079)TOELIMINATE YOURNAMEFROMDISTRIBUTION LISTSFORDOCUMENTS YOUDON'TNEED!FULLTEXTCONVERSION REQUIREDTOTALNUMBEROFCOPIESREQUIRED:
| |
| LTTR34ENCL34
| |
|
| |
|
| QsCea~5.~r~15I+A55ACCROCHESTER GASANDELECTRICCORPORATION
| | Q sCe a~5.~r~15I+A 55ACC ROCHESTER GAS AND ELECTRIC CORPORATION |
| ~89EASTAVENUE,ROCHESTER, N.Y.14649-0001 TCLCa~OhC, | | ~89 EAST AVENUE, ROCHESTER, N.Y.14649-0001 TCLCa~OhC,*RCA COOt.1III 54G.2700 May 10, 1990 U.S.Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 |
| *RCACOOt.1III54G.2700May10,1990U.S.NuclearRegulatory Commission DocumentControlDeskWashington, DC20555 | |
|
| |
|
| ==Subject:== | | ==Subject:== |
| LER90-004,SteamGenerator TubeDegradation DueToIGA/SCCCausesQ.A.ManualReportable LimitsToBeReachedR.E.GinnaNuclearPowerPlantDocketNo.50-244Inaccordance with10CFR50.73,LicenseeEventReportSystem,item(Other),andtheGinnaStationQuality,.
| | LER 90-004, Steam Generator Tube Degradation Due To IGA/SCC Causes Q.A.Manual Reportable Limits To Be Reached R.E.Ginna Nuclear Power Plant Docket No.50-244 In accordance with 10 CFR 50.73, Licensee Event Report System, item (Other), and the Ginna Station Quality,.Assurance'anual Appendix B;which requires that,"If the number of tubes in a generator falling into categories (a)or (b)exceeds the criteria, then results of the inspection shall be considered a Reportable Event pursuant to 10 CFR 50.73", the attached Licensee Event Report LER 90-004 is hereby submitted. |
| Assurance'anual AppendixB;whichrequiresthat,"Ifthenumberoftubesinagenerator fallingintocategories (a)or(b)exceedsthecriteria, thenresultsoftheinspection shallbeconsidered aReportable Eventpursuantto10CFR50.73",theattachedLicenseeEventReportLER90-004isherebysubmitted. | | This event has in no way affected the public's health and safety.Ve truly yours, o ert C.M redy Division Manager Nuclear Production xco U.S.Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna USNRC Senior Resident Inspector yA555 pgiP |
| Thiseventhasinnowayaffectedthepublic'shealthandsafety.Vetrulyyours,oertC.MredyDivisionManagerNuclearProduction xcoU.S.NuclearRegulatory Commission RegionI475Allendale RoadKingofPrussia,PA19406GinnaUSNRCSeniorResidentInspector yA555pgiP
| |
|
| |
|
| NACtea)00flOIILICENSEEEVENTREPORT{LER)VSNUCLCAAAIOVLATOARCOLNIIHION APPAOVIOOLIINO.IIIOOIOHcxf)At~~IIIIIIPACILITYNAMElllR.E.GinnaNuclearPowerPlantOOCXI'1NUMIIANlo6oo024co(07"'"'"SteamGenerator TubeDegradation DueToIGA/SCCCausesQ.A.ManualReportable LimitsToBeReachedEVINTOATIIIILtANVMICA(4AttOATOATtltlOTHERIACILITIII INVOLVEOIIIUOHTNOAYYCAAYEAHIEOUINZIAL "AIvrjONNUUIIAsfHUU~~AMONTHOAYYIAAPACILITYH*ULIOOCXtTNVUIIAI40500004909000400050900s000OtlAATIHOMOOCIli~OwtxLlvlLQ0QIIIIII.LOIILI II.LOI41(I I(0IOAOI4)lll(4 IIAOIIL)llllw)IOAOI4)(1)
| | NAC tea)00 flOII LICENSEE EVENT REPORT{LER)VS NUCLCAA AIOVLATOARCOLNIIHION APPAOVIO OLII NO.IIIO OIOH cxf)At~~IIIIII PACILITY NAME lll R.E.Ginna Nuclear Power Plant OOCXI'1 NUMIIA Nl o 6 o o 024 co(07"'"'" Steam Generator Tube Degradation Due To IGA/SCC Causes Q.A.Manual Reportable Limits To Be Reached EVINT OATI III LtA NVMICA (4 AttOAT OATt ltl OTHER IACILITIII INVOLV EO III UOHTN OAY YCAA YEAH IEOUINZIAL |
| IHIIO.H)04)(1)l
| | " AIvrjON NUUI I A sf HUU~~A MONTH OAY YIAA PACILITY H*ULI OOCXtT NVUIIAI4 0 5 0 0 0 0 4 9 0 9 0 004 00 05 090 0 s 0 0 0 Otl A AT I HO MOO C Ili~Owtx LlvlL Q 0 Q II II II.LOIILI II.LOI41(I I(0 IOAOI4)lll(4 IIAOIIL)llllw)IOAOI4)(1) |
| ~)10.MN(hIOJI(a)IIIIOMlelNI~0.114IN)II 00.114)N)lll IO.114IN)IwlIO.)14IN)(HI~0.114)NI(hIO.114IN)IMII0.114IN)ItW)IA)~O.f14)(11(HI)Ill IO.TI(IIN)la) 0THCAIOVIAIMENTI Ot10CPA$:IOUctArtWrettIAPIUNrUPI111THIIACPOATIIIUILIITTCO tUAIVANT1OT)IEAIIALtltt4AUYrtt~rprtp0MTMLHACPprrJIIA)HALICPaulGorskiMechanical Maintenance ManagerLICCNIIICONTACTPOATH)ILIAl11)*AIACOOtTELEPHONE HUMICA1552-4446COMtLCTIONtLINtfOAEACHCOLIPOHtNT IAILUACOIICAIItO INTHIIAIPOAT(11(CAUltIYCTIMCOMPOHIHTMAHUfACTVAEACOMPOHCHT MAHVfACTVAEAEPOATAILTOHtADIl~4+8ggjr~eq,Sg IVttLCMlHTAL AltOATCxtlCTCOl(4It~I>'+%/K<"'MONTHCAYYCAAtxtCCTtOlUILNCCIOH OATI(I~I'YCIl)ttw.~NCIPICTCOIUIUIIJIOH OATCIXHOLAITAACTI@cutIPIt00NWN,II,tyyntrrleltAlrrLalLPtmCtnttrrINAVM(14Duringthe1990AnnualRefueling andMaintenance Outagesubsequent totheeddycurrentexamination performed onboththe"A"and"B"Westinghouse Series44SteamGenerators (S/G),75tubesinthe"A"S/Gand211tubesinthe"B"S/Grequiredcorrective actionduetotubedegradation. | | IHI IO.H)04)(1)l |
| Thisdefectpopulation includes28tubesinthe"B"S/Gthathadknowndefectspluggedinprioroutages.Thesetubeswereunplugged forfulllengtheddycurrentexamination andwerereturnedtoservicewithasleeverepairinthedegradedregion.Theimmediate causeoftheeventwasthatthe"A"and"B"S/Gtubedegradation wasinexcessoftheGinnaQualityAssurance Manualreportability limits.Theunderlying causeofthetubedegradation isacommonS/Gproblemofapartially rolledtubesheetcrevicewithrecurring Zntergranular Attack/Stress Corrosion Cracking(EGA/SCC) andPrimaryWaterStressCorrosion Cracking(PWSCC)attackonS/Gtubing.Corrective actiontakenwastoeithersleeveorplugtheaffectedtubeswithacceptedindustryrepairmethods.NACtersIIIIIIII
| | ~)10.MN(h IO JI(a)II I IOM lelNI~0.114IN)II 00.114)N)lll IO.11 4 IN)I wl IO.)14 IN)(HI~0.114)N I(h IO.1 14 IN)IMI I0.1 14 IN)I tW)IA)~O.f 14)(11(HI)Ill IO.TI(IIN)la) 0 THC AIOVIAIMENTI Ot 10CPA$: IOUct Art W re tt IAP IUNrUPI 111 THII ACPOAT II IUILIITTCO tUAIVANT 1 OT)IEA IIALtltt 4 AUYrtt~rpr tp0 M TML HAC Pprr JIIA)HALIC Paul Gorski Mechanical Maintenance Manager LICCNIII CONTACT POA TH)I LIA l11)*AIA COOt TELEPHONE HUMICA 1552-4446 COMtLCTI ONt LINt fOA EACH COLIPOHtNT I AILUAC OIICAIItO IN THII AIPOAT (11(CAUlt IYCTIM COMPO H I HT MAHUf AC TVAEA COMPOHCHT MAHVfAC TVAEA EPOATAIL TO HtADI l~4+8ggjr~eq,Sg IVttLCMlHTAL AltOAT CxtlCTCO l(4 It~I>'+%/K<"'MONTH CAY YCAA txtCCTtO lUILNCCIOH OATI (I~I'YCI l)t tw.~N CIPICTCO IUIUIIJIOH OATCI X HO LAITA ACT I@cut IP I t00 NWN, I I, tyyntrr lelt AlrrL alLPtmCt nttrrINA VM (14 During the 1990 Annual Refueling and Maintenance Outage subsequent to the eddy current examination performed on both the"A" and"B" Westinghouse Series 44 Steam Generators (S/G), 75 tubes in the"A" S/G and 211 tubes in the"B" S/G required corrective action due to tube degradation. |
| | This defect population includes 28 tubes in the"B" S/G that had known defects plugged in prior outages.These tubes were unplugged for full length eddy current examination and were returned to service with a sleeve repair in the degraded region.The immediate cause of the event was that the"A" and"B" S/G tube degradation was in excess of the Ginna Quality Assurance Manual reportability limits.The underlying cause of the tube degradation is a common S/G problem of a partially rolled tube sheet crevice with recurring Zntergranular Attack/Stress Corrosion Cracking (EGA/SCC)and Primary Water Stress Corrosion Cracking (PWSCC)attack on S/G tubing.Corrective action taken was to either sleeve or plug the affected tubes with accepted industry repair methods.NAC ters III IIIII |
|
| |
|
| NRCForm944Al943ILICENSEEEVENTREPORT(LER) TEXTCONTINUATION U.S, NUCLEARREOULATORY COMMISSION APPROVEOOMENO3140& l04EXPIRES'ISI/SS FACILITYNAME>> IOOCKETNUMEERQlLERNUMEER( 41wE::,SEOVENTIAL NVMEAAEVOIOHMVMER~ AOEIS)R.E. GinnaNuclearPowerPlantTEXTIIFuuuu<<MPI<<newt'.uup<<PPppA<< IYIICAvwSOSASIIITI0500024490- 004- 0002OF07PRE-EVENT PLANTCONDITIONS Theunitwasincold/refueling shutdownfortheAnnualRefueling Maintenance Outage. AllfuelhadbeenremovedfromtheReactorVessel. SteamGenerator eddycurrentinspection wasinprogress. | | NRC Form 944A l943I LICENSEE EVENT REPORT (LER) TEXT CONTINUATION U.S, NUCLEAR REOULATORY COMMISSION APPROVEO OME NO 3140& l04 EXPIRES'ISI/SS FACILITY NAME>> I OOCKET NUMEER Ql LER NUMEER ( 41 wE::, SEOVENTIAL NVM EA AEVOIOH MVM ER~ AOE IS)R.E. Ginna Nuclear Power Plant TEXT IIF uuuu<<MPI<<newt'.uup<<PPppA<< IYIIC Avw SOSASI I I TI 0 5 0 0 0 2 4 4 90- 0 04- 000 2 OF 0 7 PRE-EVENT PLANT CONDITIONS The unit was in cold/refueling shutdown for the Annual Refueling Maintenance Outage. All fuel had been removed from the Reactor Vessel. Steam Generator eddy current inspection was in progress.DESCRIPTION OP EVENT A. DATES AND APPROXIMATE TIMES FOR MAJOR OCCURRENCES: |
| DESCRIPTION OPEVENTA.DATESANDAPPROXIMATE TIMESFORMAJOROCCURRENCES: | | o April 16, 1990, 1711 EDST: Event date and time.~t>o.':pril 16, 1990, 1711 EDST: Discovery date and".;.me.o:.ril 18, 1990, 1000 EDST: Oral notification made to the NRC Office of Nuclear Reactor."Regulation (NRR).o A--.ril 21, 1990, 1800 EDST: Steam Generator rap'rs completed. |
| oApril16,1990,1711EDST:Eventdateandtime.~t>o.':pril16,1990,1711EDST:Discovery dateand".;.me.o:.ril18,1990,1000EDST:Oralnotification madetotheNRCOfficeofNuclearReactor."Regulation (NRR).oA--.ril21,1990,1800EDST:SteamGenerator rap'rscompleted.
| | o May 1, 1990: Followup report sent to NRC Office of Nuclear Reactor Regulation (NRR).B.EVENT: During the 1990 Annual Refueling and Maintenance Outage, an eddy current examination was performed in both the>>A>>and>>B>>Westinghouse Series 44 Design recirculating steam generators. |
| oMay1,1990:FollowupreportsenttoNRCOfficeofNuclearReactorRegulation (NRR).B.EVENT:Duringthe1990AnnualRefueling andMaintenance Outage,aneddycurrentexamination wasperformed inboththe>>A>>and>>B>>Westinghouse Series44Designrecirculating steamgenerators.
| | The purpose of the eddy current examination was to assess any corrosion or mechanical damage that may have occurred during the cycle since May 1989.'AC FORM 944A 1943 I NRC Perm SSBA (9451 LICENSEE EVENT REPORT (LERI TEXTCONTINUATION V.S.NUCLEAR REOULATORY COMMISSION APPROVED OMB NO 5150&(95 ExPIRES: BISI/BS PACILITY NAME (11 DOCKET NUMBER (l(LER NUMBER (51 55OVENTIAL NVM REVISION I4vM eo~AOE (SI R.E..Ginna Nuclear Power Plant TEXT lll more 5oece r reeeeerf.Pee~Arr(C forrrr JRLI Sl (Ill o s o o o 2 4 4 90-004 00 03 oF 07 The examination was performed by personnel from Rochester Gas and Electric Corporation (RG&E)f and Allen Nuclear Associates. |
| Thepurposeoftheeddycurrentexamination wastoassessanycorrosion ormechanical damagethatmayhaveoccurredduringthecyclesinceMay1989.'ACFORM944A1943I NRCPermSSBA(9451LICENSEEEVENTREPORT(LERITEXTCONTINUATION V.S.NUCLEARREOULATORY COMMISSION APPROVEDOMBNO5150&(95ExPIRES:BISI/BSPACILITYNAME(11DOCKETNUMBER(l(LERNUMBER(5155OVENTIAL NVMREVISIONI4vMeo~AOE(SIR.E..Ginna NuclearPowerPlantTEXTlllmore5oecerreeeeerf.
| | All personnel had been trained and qualified in the eddy current examination method and had been certified to a minimum of Level I for data acquisition and Level II for data analysis.The eddy current examination of the"A" and"B" steam generators was performed utilizing the Zetec MIZ-18 Digital Data Acquisition System.The frequencies selected were 400, 200, 100, and 25 KHZ.The inlet or hot leg examination program plan was generated to provide the examination of 1004 of each open (not sleeved, or plugged)steam generator tube from the tube end to the first tube support.Zn addition, 204 of these tubes were selected and examined for their full length as recommended in the EPRZ guidelines. |
| Pee~Arr(CforrrrJRLISl(Illosooo24490-0040003oF07Theexamination wasperformed bypersonnel fromRochester GasandElectricCorporation (RG&E)fandAllenNuclearAssociates. | | All tubes with previous indications greater than 20%through wall (TW)depth were examined at a minimum to the location of their degradation. |
| Allpersonnel hadbeentrainedandqualified intheeddycurrentexamination methodandhadbeencertified toaminimumofLevelIfordataacquisition andLevelIIfordataanalysis.
| | Approximately 20%of all open Row 1 U-bend regions were examined with the Motorized Rotating Pancake Coil (MRPC)between the g6 Tube Support Plate Hot (TSPH)and the 56 Tube Support Plant Cold (TSPC)from the cold leg side.Results of the above inspections indicated that 75 tubes in the"A" steam generator and 211 tubes in the"B" steam generator (183 new repairs plus 28 previously plugged tubes)required corrective action.C.INOPERABLE STRUCTURES f COMPONENTS f OR SYSTEMS THAT CONTRIBUTED TO THE EVENT: D.None.OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED: None.E.METHOD OF DISCOVERY: |
| Theeddycurrentexamination ofthe"A"and"B"steamgenerators wasperformed utilizing theZetecMIZ-18DigitalDataAcquisition System.Thefrequencies selectedwere400,200,100,and25KHZ.Theinletorhotlegexamination programplanwasgenerated toprovidetheexamination of1004ofeachopen(notsleeved,orplugged)steamgenerator tubefromthetubeendtothefirsttubesupport.Znaddition, 204ofthesetubeswereselectedandexaminedfortheirfulllengthasrecommended intheEPRZguidelines.
| | The event was apparent after the review of the eddy current examination results.NRC POIIM 555A (9451 NRC Form 844A 19')5 I D LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S.NUCLEAR REOULATORY COMMISSION APPROVED OM8 NO S)50&104 EXPIRES.8181/85 FACILITY NAME 111 R.E.Ginna Nuclear Power Plant TEXT lll mcpp N>>ce N>>EMcpd.p>>cHtcNanV HRC%%dmc JR)A 9)I IT)DOCKET NUMSER Ill o 5 o o o 244 I.ER NUMSER 14)54OVSNTIAL 45.EVIS o NVMSSR-.I" MVM 90-004 00 PACE (8)4 oF 0 7 F.OPERATOR ACTION: None.G.SAFETY SYSTEM RESPONSES: |
| Alltubeswithpreviousindications greaterthan20%throughwall(TW)depthwereexaminedataminimumtothelocationoftheirdegradation.
| | None.III. CAUSE OF EVENT A.IMMEDIATE CAUSE: The Immediate Cause of the event was that the<<A<<and'B<< steam generator tube degradation was in excess of the Ginna Quality Assurance Manual Reportable Limits.B. ROOT CAUSE: The results of the examination indicate that the ZGA and SCC continue to be active within the tube sheet crevice region on the inlet side of each steam generator. |
| Approximately 20%ofallopenRow1U-bendregionswereexaminedwiththeMotorized RotatingPancakeCoil(MRPC)betweentheg6TubeSupportPlateHot(TSPH)andthe56TubeSupportPlantCold(TSPC)fromthecoldlegside.Resultsoftheaboveinspections indicated that75tubesinthe"A"steamgenerator and211tubesinthe"B"steamgenerator (183newrepairsplus28previously pluggedtubes)requiredcorrective action.C.INOPERABLE STRUCTURES fCOMPONENTS fORSYSTEMSTHATCONTRIBUTED TOTHEEVENT:D.None.OTHERSYSTEMSORSECONDARY FUNCTIONS AFFECTED: | | As in the past, the IGA/ SCC is much more prevalent in the<<B<< steam generator with 108 ZGA indications and 49 SCC indications reported. Zn the<<A<< steam generator 16 ZGA indications and 22 SCC indications were reported. The majority of the inlet tube sheet crevice corrosion indications are IGA/ SCC of the Mil-annealed Inconel 600 Tube Material. This form of corrosion is believed to be the result of the tube sheet crevices forming an alkaline environment. |
| None.E.METHODOFDISCOVERY: | | This environment has developed over the years as deposits and active species have reacted with sodium and phosphate, changing a neutral or inhibited crevice into the aggressive environment that presently exists. In addition to the IGA/ SCC in the crevices PWSCC indications continue to be found at the roll transi-tion. This year there were 23 PWSCC indications in the<<B<< steam generator and 37 PWSCC indications in the<<A<<. steam generator. |
| Theeventwasapparentafterthereviewoftheeddycurrentexamination results.NRCPOIIM555A(9451 NRCForm844A19')5IDLICENSEEEVENTREPORT(LER)TEXTCONTINUATION U.S.NUCLEARREOULATORY COMMISSION APPROVEDOM8NOS)50&104EXPIRES.8181/85FACILITYNAME111R.E.GinnaNuclearPowerPlantTEXTlllmcppN>>ceN>>EMcpd.p>>cHtcNanVHRC%%dmcJR)A9)IIT)DOCKETNUMSERIllo5ooo244I.ERNUMSER14)54OVSNTIAL 45.EVISoNVMSSR-.I"MVM90-00400PACE(8)4oF07F.OPERATORACTION:None.G.SAFETYSYSTEMRESPONSES:
| | NAC FOIIM 545A 19851 HRC form 559A (9451 PACILITY NAME 111UCENSEE EVENT REPORT ILERI TEXT CONTINUATION I/.9.NI/CLEAR REGULATORY COMMIEEION APPROVEO OME I/O 515OPOIOA EXP/RES: 9/91/85 LER NUMEER 191 R.E.Ginna Nuclear Power Plant TEXT/ll mIVO Opoco IP Olvrror/, vro~/V/IC fo/III PS@'p/IITI o s o o o 2 4 4 vEAR 9 0 9 9 O I/C/IT I A L II I/II C R 0 0 4 R 9 V If IO Ir Ir I/II 9 A 000 5 OF 07 Ivo ANALYSIS OF EVENT The event is reportable in accordance with 10 CFR 50.73, Licensee Event Report Item (other)and the Ginna Station Quality Assurance Manual Appendix B which requires that,"If the number of tubes in a generator falling into'ategories (a)or (b)exceeds the criteria, then results of the inspection shall be considered a Reportable Event pursuant to 10 CFR 50.73." The tube degradation in the"A" and"B" steam generators exceeded the criterion of (b)which states,"More than 1%of the total tubes inspected are degraded, (imperfections greater than the repair limit)." This repair limit is defined as,"steam generator tubes that have imperfections greater than 40%through wall, as'ndicated by eddy current, shall be repaired by-,@ging or sleeving." An assessment was performed considering.the safety con-sequences and implications of this event with the following results and conclusions: |
| None.III. CAUSEOFEVENTA.IMMEDIATE CAUSE: TheImmediate Causeoftheeventwasthatthe<<A<<and'B<< steamgenerator tubedegradation wasinexcessoftheGinnaQualityAssurance ManualReportable Limits.B. ROOTCAUSE: Theresultsoftheexamination indicatethattheZGAandSCCcontinuetobeactivewithinthetubesheetcreviceregionontheinletsideofeachsteamgenerator. | | There were no operational or safety consequences or safety im",.'ations resulting from the steam generator tube dec;dation in excess of the Q.A.Manual reportable limits bee..use: o The degraded tubes were identified and repaired prior to any significant leakage or S/G tube rupture occurring. |
| Asinthepast,theIGA/SCCismuchmoreprevalent inthe<<B<<steamgenerator with108ZGAindications and49SCCindications reported.
| | o Even assuming a complete severance of a steam generator tube at full power, as stated in the R.E.Ginna Nuclear Power Plant Updated Final Safety Analysis Report (Ginna/UFSAR) |
| Znthe<<A<< steamgenerator 16ZGAindications and22SCCindications werereported. | | Section 15.6.3, (Steam-Generator Tube Rupture)the sequence, of recovery actions ensures early termination of primary to secondary leakage with or without offsite power available thus limiting offsite radiation doses to within the guidelines of 10 CFR 100.Based on the above, it can be concluded that the public's health and safety were assured at all times.NRC PORM POOA 19 EE I NRC Form SSSA 19431 IEACILITY NAME III LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.5.NUCLEAR REGULATORY COMMISSION APPROVEO OMS HO 5150WI05 EXPIRES SISI/95 LER NUMSER IS)R.E.Ginna Nuclear Power Plant TEXT Ill more opec@rp orrrrerF.1>>poooronolHlIC Form PxWSI IITI o s o o o2 44 90 55GUSNTrAL |
| Themajorityoftheinlettubesheetcrevicecorrosion indications areIGA/SCCoftheMil-annealed Inconel600TubeMaterial.
| | ~rUM 51 004 rr 5 Y l5 lO 1~rUrr 51-0 0 06oF 7 CORRECTIVE ACTION A.ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS: o Of the 75 degraded tubes in the"A" Steam Generator, 51 tubes were repaired using a Combustion Engineering welded 27" sleeve in the hot leg and these tubes will remain in service.The remaining 24 tubes were removed from active service by plugging both the hot and cold leg tube ends.o Of the 211 degraded tubes in the"B" steam'enerator, 191 tubes were repaired using a Combustion Engineering welded 27" sleeve in the hot leg and these tubes will remain in service.The remaining 20 tubes were removed from active service by plugging both the hot and cold leg tube ends.B.ACTION TAKEN OR PLANNED TO PREVENT RECUSANCE: |
| Thisformofcorrosion isbelievedtobetheresultofthetubesheetcrevicesforminganalkalineenvironment.
| | The occurrence/presence of IGA, SCC and PWSCC is a common PHR Steam Generator problem.Utile~"ies with susceptible tubing and partially rolled crevices must deal with this recurring attack on steam generator tubing.R.E.Ginna Station will continue careful monitoring of both primary RCS and secondary side water chemistry parameters. |
| Thisenvironment hasdeveloped overtheyearsasdepositsandactivespecieshavereactedwithsodiumandphosphate, changinganeutralorinhibited creviceintotheaggressive environment thatpresently exists. InadditiontotheIGA/ SCCinthecrevicesPWSCCindications continuetobefoundattherolltransi-tion. Thisyeartherewere23PWSCCindications inthe<<B<< steamgenerator and37PWSCCindications inthe<<A<<. steamgenerator. | | These water chemistry parameters will be evaluated against accepted industry guidelines in order to minimize harmful primary and/or secondary side environments. |
| NACFOIIM545A19851 HRCform559A(9451PACILITYNAME111UCENSEEEVENTREPORTILERITEXTCONTINUATION I/.9.NI/CLEARREGULATORY COMMIEEION APPROVEOOMEI/O515OPOIOA EXP/RES:9/91/85LERNUMEER191R.E.GinnaNuclearPowerPlantTEXT/llmIVOOpocoIPOlvrror/,
| | Degraded S/G tubes shall be sleeved or plugged in accordance with the inservice inspection program and accepted industry repair methods.~roc po1M 555k 19M I NRC FoIm 344A (943)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION V.S.NUCLEAR REGULATORY COMMISSION APPROVEO OMB NO, 3)50M)04 EXPIRESI 813)185 FACILITY NAME 111 OOCKET NUMBER Ill YEAR LER NUMBER IS)44OUENTIAL NUM SR REVISION NUM 4 II~AGE IS)R.E.Ginna Nuclear Power Plant TEXT llf mCVP g>>44 N fPqviNL MW~ff)IC Ann 3RLISl)IT)0 5 0 0 0 2 4 4 9 p 004 0 0 7 QF p ADDITIONAL INFORMATION A.FAILED COMPONENTS: |
| vro~/V/ICfo/IIIPS@'p/IITIosooo244vEAR9099OI/C/ITIALIII/IICR004R9VIfIOIrIrI/II9A0005OF07IvoANALYSISOFEVENTTheeventisreportable inaccordance with10CFR50.73,LicenseeEventReportItem(other)andtheGinnaStationQualityAssurance ManualAppendixBwhichrequiresthat,"Ifthenumberoftubesinagenerator fallinginto'ategories (a)or(b)exceedsthecriteria, thenresultsoftheinspection shallbeconsidered aReportable Eventpursuantto10CFR50.73."Thetubedegradation inthe"A"and"B"steamgenerators exceededthecriterion of(b)whichstates,"Morethan1%ofthetotaltubesinspected aredegraded, (imperfections greaterthantherepairlimit)."Thisrepairlimitisdefinedas,"steamgenerator tubesthathaveimperfections greaterthan40%throughwall,as'ndicated byeddycurrent,shallberepairedby-,@gingorsleeving." | | None.B.PREVIOUS LERs ON SIMILAR EVENTS: A similar LER event historical search was conducted with the following results: The crevice indications are similar to those reported in A0-74-02, A0-75-07, R0-75-013, and LER's 76-008, 77-008, 78-003, 79-006, 79-022, 80-003, 81-009, 82-003, 82-022, 83-013, and 89-001.C.SPECIAL COMMENTS: None.NRC FORM 344A 1983) 0 E, J r}} |
| Anassessment wasperformed considering
| |
| .thesafetycon-sequences andimplications ofthiseventwiththefollowing resultsandconclusions: | |
| Therewerenooperational orsafetyconsequences orsafetyim",.'ationsresulting fromthesteamgenerator tubedec;dationinexcessoftheQ.A.Manualreportable limitsbee..use:
| |
| oThedegradedtubeswereidentified andrepairedpriortoanysignificant leakageorS/Gtuberuptureoccurring.
| |
| oEvenassumingacompleteseverance ofasteamgenerator tubeatfullpower,asstatedintheR.E.GinnaNuclearPowerPlantUpdatedFinalSafetyAnalysisReport(Ginna/UFSAR)
| |
| Section15.6.3,(Steam-Generator TubeRupture)thesequence, ofrecoveryactionsensuresearlytermination ofprimarytosecondary leakagewithorwithoutoffsitepoweravailable thuslimitingoffsiteradiation dosestowithintheguidelines of10CFR100.Basedontheabove,itcanbeconcluded thatthepublic'shealthandsafetywereassuredatalltimes.NRCPORMPOOA19EEI NRCFormSSSA19431IEACILITY NAMEIIILICENSEEEVENTREPORT(LER)TEXTCONTINUATION U.5.NUCLEARREGULATORY COMMISSION APPROVEOOMSHO5150WI05EXPIRESSISI/95LERNUMSERIS)R.E.GinnaNuclearPowerPlantTEXTIllmoreopec@rporrrrerF.
| |
| 1>>poooronolHlIC FormPxWSIIITIosooo2449055GUSNTrAL | |
| ~rUM51004rr5Yl5lO1~rUrr51-0006oF7CORRECTIVE ACTIONA.ACTIONTAKENTORETURNAFFECTEDSYSTEMSTOPRE-EVENT NORMALSTATUS:oOfthe75degradedtubesinthe"A"SteamGenerator, 51tubeswererepairedusingaCombustion Engineering welded27"sleeveinthehotlegandthesetubeswillremaininservice.Theremaining 24tubeswereremovedfromactiveservicebypluggingboththehotandcoldlegtubeends.oOfthe211degradedtubesinthe"B"steam'enerator, 191tubeswererepairedusingaCombustion Engineering welded27"sleeveinthehotlegandthesetubeswillremaininservice.Theremaining 20tubeswereremovedfromactiveservicebypluggingboththehotandcoldlegtubeends.B.ACTIONTAKENORPLANNEDTOPREVENTRECUSANCE: | |
| Theoccurrence/presence ofIGA,SCCandPWSCCisacommonPHRSteamGenerator problem.Utile~"ieswithsusceptible tubingandpartially rolledcrevicesmustdealwiththisrecurring attackonsteamgenerator tubing.R.E.GinnaStationwillcontinuecarefulmonitoring ofbothprimaryRCSandsecondary sidewaterchemistry parameters.
| |
| Thesewaterchemistry parameters willbeevaluated againstacceptedindustryguidelines inordertominimizeharmfulprimaryand/orsecondary sideenvironments.
| |
| DegradedS/Gtubesshallbesleevedorpluggedinaccordance withtheinservice inspection programandacceptedindustryrepairmethods.~rocpo1M555k19MI NRCFoIm344A(943)LICENSEEEVENTREPORT(LER)TEXTCONTINUATION V.S.NUCLEARREGULATORY COMMISSION APPROVEOOMBNO,3)50M)04EXPIRESI813)185FACILITYNAME111OOCKETNUMBERIllYEARLERNUMBERIS)44OUENTIAL NUMSRREVISIONNUM4II~AGEIS)R.E.GinnaNuclearPowerPlantTEXTllfmCVPg>>44NfPqviNLMW~ff)ICAnn3RLISl)IT)050002449p004007QFpADDITIONAL INFORMATION A.FAILEDCOMPONENTS:
| |
| None.B.PREVIOUSLERsONSIMILAREVENTS:AsimilarLEReventhistorical searchwasconducted withthefollowing results:Thecreviceindications aresimilartothosereportedinA0-74-02, A0-75-07, R0-75-013, andLER's76-008,77-008,78-003,79-006,79-022,80-003,81-009,82-003,82-022,83-013,and89-001.C.SPECIALCOMMENTS: | |
| None.NRCFORM344A1983) 0E,Jr}} | |
Similar Documents at Ginna |
|---|
Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML17265A7541999-09-22022 September 1999
LER 99-011-00:on 990823,small Tears Were Discovered in Flexible Duct Work Connector at Inlet of CR HVAC Sys Return Air Fan (AKF08).Caused by in-leakage Greater than That Assumed.Implemented Temporary Mod 99-029.With 990922 Ltr
ML17265A7431999-08-24024 August 1999
LER 99-004-01:on 990412,discovered That Containment Recirculation Fan Chevron Separator Vanes Were Installed Backwards.Caused by Improper Assembly by Mfg.Moisture Separator Vanes Were Dismantled & Correctly re-installed
ML17265A7181999-07-23023 July 1999
LER 99-007-01:on 990423,reactor Trip Occurred Due to Instrument & Control Technicians Inadvertently Pulling Fuses from Wrong Nuclear Instrument Channel.Setpoint Adjustments Were Completed by Different Crew of Technicians
ML17265A7081999-07-22022 July 1999
LER 98-003-02:on 980904,actuations of CR Emergency Air Treatment Sys Was Noted Due to Invalid Causes.Caused by Various Degraded Components in CR RM Sys.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored
ML17265A7031999-07-19019 July 1999
LER 99-S01-00:on 990617,determined That Temporary Unescorted Access Had Been Granted to Contractor Employee.Caused by Incomplete Info Re Circumstances of Individual Military Separation.Individual Access Was Revoked.With 990719 Ltr
ML17265A7021999-07-15015 July 1999
LER 99-010-00:on 990615,ventilation Isolation of Auxiliary Bldg Occurred When Auxiliary Bldg Gas Radiation Monitor R-14 Reached High Alarm Setpoint.Cr Operators Rest Auxiliary Bldg Ventilation Isolation Signal.With 990715 Ltr
ML17265A6851999-06-21021 June 1999
LER 99-001-01:on 990222,deficiencies in NSSS Vendor steam- Line Brake Mass & Energy Release Analysis Results in Plant Being Outside Design Bases Occurred.Caused by Deficiencies in W.Temporary Administrative Replaced.With 990621 Ltr
ML17265A6661999-06-0202 June 1999
LER 99-009-00:on 990503,instrumentation Declared Inoperable in Multiple Channels Resulted in Condition Prohibited by Ts. Caused by Unanticipated High Frequency AC Voltage Ripple. Entered TS LCO 3.0.3.With 990602 Ltr
ML17309A6541999-05-27027 May 1999
LER 99-008-00:on 990427,overtemperature Delta T Reactor Trip Occurred Due to Faulted Bistable During Calibr of Redundant Channel.Plant Was Stabilized in Mode 3 & Faulted Bistable Was Subsequently Replaced.With 990527 Ltr
ML17265A6631999-05-24024 May 1999
LER 99-007-00:on 990423,technicians Inadvertently Pulled Fuses from Wrong Nuclear Instrument Cahnnel,Causing Reactor Trip,Due to High Range Flux Trip.Caused by Personnel Error. Labeling Scheme Improved
ML17265A6601999-05-21021 May 1999
LER 99-006-00:on 990421,start of turbine-driven Auxiliary Feedwater Pump Was Noted.Caused by MOV Being Left in Open Position.Closed Manual Isolation Valve to Secure Steam to Pump.With 990521 Ltr
ML17265A6441999-05-13013 May 1999
LER 99-005-00:on 990413,undervoltage Signal of Safeguards Bus During Testing Resulted in Automatic Start of B Edg. Caused by Personnel Error.Blown Fuse Was Replaced & Offsite Power Was Restored to Safeguards Bus 17.With 990513 Ltr
ML17265A6431999-05-12012 May 1999
LER 99-004-00:on 990412,discovered That Containment Recirculation Fan Moisture Separator Vanes Were Incorrectly Installed,Per 10CFR21.Caused by Improper Assembly by Mfg. Subject Vanes Were Dismantled & Correctly re-installed
ML17265A6141999-03-31031 March 1999
LER 99-003-00:on 990301,two Main Steam non-return Check Valves Were Declared Inoperable Due to Exceedance of Acceptance Criteria.Caused by Changes in Methodology & Matls.Packing Gland Torque Will Be Adjusted.With 990331 Ltr
ML17265A6131999-03-29029 March 1999
LER 99-002-00:on 990227,discovered That Surveillance Had Not Been Performed at Frequency,Per Ts.Caused by Personnel Error.Procedure O-6.13 Will Be Evaluated for Enhancement Documentation of Completion of ITS Srs.With 990329 Ltr
ML17265A6061999-03-24024 March 1999
LER 99-001-00:on 990222,plant Was Noted Outside Design Basis.Caused by Deficiencies in NSSS Vendor Slb Mass & Energy Release.Placed Temporary Administrative Restriction 40 Degrees F Max on Screenhouse Bay Temp
ML17265A4951998-12-21021 December 1998
LER 98-005-00:on 981120,loss of 34.5 Kv Offsite Power Circuit 751,resulted in Automatic Start of B Edg.Caused by Faulted Cable Splice.Performed Appropriate Actions of Abnormal Procedure AP-ELEC.1.With 981221 Ltr
ML17265A4931998-12-17017 December 1998
LER 98-004-00:on 971030,determined That Improperly Performed Surveillance Resulted in Condition Prohibited by Ts.Caused by Procedure non-adherence.Appropriate Calibr Procedures Were Properly Performed with 24 H of Condition Discovery
ML17265A4691998-11-25025 November 1998
LER 98-003-01:on 980904,actuations of CR Emergency Air Treatment Systems (Creats) Occurred.Caused by Radon build-up During Temp Inversion.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored to CR
ML17265A4271998-10-0505 October 1998
LER 98-003-00:on 980904,actuations of CR Emergency Air Treatment Sys Occurred.Caused by Radon build-up During Temp Inversion.Air Samples Were Taken & Determined That Source of Radiation Was Naturally Occurring Radon.With 981005 Ltr
ML17265A3671998-07-14014 July 1998
LER 98-002-00:on 971019,CR Emergency Air Treatment Sys Actuating Function Was Not Operable.Caused by Mispositioned Switch.Revised Procedure CPI-MON-R37.W/980714 Ltr
ML17265A1921998-03-11011 March 1998
LER 98-001-00:on 980209,discovered That Boraflex Degradation in SPF Was Greater than Was Assumed.Caused by Dissolution of Boron on Boraflex Matrix,Per 10CFR50.21.Removed Spent Fuel Assemblies from Selected Degraded Storage Rack Cells
ML17265A1641998-02-0606 February 1998
LER 97-007-01:on 971117,reactor Engineer Recognized That Neutron Flux Low Range Trip Circuitry for Channel Was Not in Tripped Condition as Required.Caused by Technical Inadequacies.Channel Defeat Will Be Identified
ML17265A1601998-02-0606 February 1998
LER 97-006-01:on 971103,verification of B Concentration Was Not Performed Due to Misinterpretation of Event Sequence. Audible Count Rate Function Was Restored to Operable Status
ML17264B1441997-12-17017 December 1997
LER 97-007-00:on 971117,NF Low Range Trip Circuitry for Channel N-44 Was Not Placed in Tripped Condition.Caused by Technical Inadequacies in Procedures.Implemented EWR 4862 to Resolve Design deficiency.W/971217 Ltr
ML17264B1291997-12-0303 December 1997
LER 97-006-00:on 971103,NIS Audible Count Rate Function Was Inoperable.Caused by Misinterpretation of Event Sequence Due to Not Verifying Boron Concentration.B Verification Occurred Every 12 H Per ITS LCO Action 3.9.2.C.3.W/971203 Ltr
ML17264B1271997-12-0101 December 1997
LER 97-005-00:on 971031,undetected Unblocking of SI Actuation Signal Occurred at Low Pressure Condition,Due to Faulty Bistable Which Resulted in Inadvertent SI Actuation Signal.Sias,Ci & CVI Signals Were Reset
ML17264B1211997-11-24024 November 1997
LER 97-004-00:on 971024,radiation Monitor Alarm Were Noted Due to Higher than Normal Radioactive Gas Concentration Resulted in Cvi.New R-12 Alarm Setpoint Was Maintained for Duration of Refueling Outage
ML17264B0461997-09-29029 September 1997
LER 97-003-01:on 970730,bistable Instrument Trip Setpoint Could Have Exceeded Allowable Value.Caused by Insufficient Existing Margin Between Trip Setpoint & Allowable Value. Held Switches in Tripped configuration.W/970929 Ltr
ML17264B0111997-08-27027 August 1997
LER 97-003-00:on 970730,high Steam Flow Bistable Instrument Setpoint Plus Instrument Uncertainty Could Exceed Allowable Value in ITS Was Identified.Caused by Entry Into ITS LCO 3.0.3.Switches Placed in Tripped configuration.W/970827 Ltr
ML17264A9941997-08-19019 August 1997
LER 97-002-00:on 970720,34.5 Kv Offsite Power Circuit 751 Was Lost.Caused by Automatic Actuation of B Emergency DG Due to Undervoltage on Safeguards Buses 16 & 17.Offsite Power Restored to Safeguards Buses 16 & 17.W/970819 Ltr
ML17264A9911997-08-11011 August 1997
LER 96-009-02:on 960723,determined That Leak Rate Outside Containment Was Greater than Program Limit.Caused by Weld Defect.Isolated Leak & Cut Out & Replaced Leaking Pipe
ML17264A8271997-03-0303 March 1997
LER 97-001-00:on 970131,discovered Service Water Temp Was Less than Specified Value.Caused by non-representative Method of Monitoring.Increased Water Temp in Screenhouse Bay to Greater than 35 Degrees F.W/970303 Ltr
ML17264A8071997-01-22022 January 1997
LER 96-015-00:on 961223,discovered Thermally Induced Overpressure Transient Could Occur.Caused by Thermal Expansion of Fluid During Design Basis Accident Condition. Installed Relief Valve on Affected line.W/970122 Ltr
ML17264A7471996-11-27027 November 1996
LER 96-013-00:on 961029,circuit Breakers Closed While in Mode 3 & Resulted in Condition Prohibited by TS Due to Personnel Error.Circuit Breakers for MOV-878B & MOV-878D Were re-opened.W/961127 Ltr
ML17264A6051996-09-19019 September 1996
LER 96-012-00:on 960820,feedwater Transient Occurred,Due to Closure of Feedwater Regulating Valve,Causing Lo Lo Steam Generator Level Reactor Trip.Sgs Were Restored & Missing Screw in 1/P-476 Was replaced.W/960919 Ltr
ML17264A6061996-09-19019 September 1996
LER 96-009-01:on 960723,leakage Outside Containment Occurred,Due to Weld Defect,Resulting in Leak Rate Greater than Program Limits.Source of Leakage Isolated from RWST by Freeze Seal,Allowing Exit from ITS LCO 3.0.3.W/960919 Ltr
ML17264A5911996-09-0505 September 1996
LER 96-011-00:on 960807,improper Configuration of Circuit Breaker Occurred,Due to Undetected Internal Interference, Resulting in Automatic Start of Both Auxiliary Feedwater Pumps.Running AFW Pumps Were secured.W/960905 Ltr
ML17264A5921996-09-0505 September 1996
LER 96-010-00:on 960806,latching of Main Turbine While in Mode 4 Occurred,Due to Defective Procedure,Resulting in Automatic Start of Auxiliary Feedwater Pump.Caused by Defective Maint Procedure.Procedure revised.W/960905 Ltr
ML17264A5891996-08-22022 August 1996
LER 96-009-00:on 960723,determined Leak on Piping Sys Outside Containment Greater than Program Limit.Caused by Weld Defect.Pipe & Socket Welds Were Cut Out & Replaced. W/960822 Ltr
ML17264A5781996-08-0606 August 1996
LER 96-008-00:on 960707,main Feedwater Pump Breakers Opened. Caused by Change in Seal Water Differential Pressure Occurred During Sys Realignment.Afw Flow Controlled as Desired to Maintain S/G level.W/960806 Ltr
ML17264A5561996-07-12012 July 1996
LER 96-007-00:on 960612,CR Operators Identified Control Rods Misaligned & Not Moving in Proper Sequence.Caused by Faulty Firing Circuit Card in Rod Control Sys.Faulty Firing Circuit Card in 1BD Power Cabinet replaced.W/960712 Ltr
ML17264A5421996-06-20020 June 1996
LER 96-006-00:on 960521,discovered Containment Penetration Not in Required Status.Caused by Personnel Error.Installed Flange Inside Containment Penetration 2.W/960620 Ltr
ML17264A5411996-06-17017 June 1996
LER 96-005-00:on 960516,PORC Determined Deficient Procedures Do Not Meet SRs for Testing safety-related Logic Circuits. Caused by Inadequancies in Individual Testing Procedures. Procedures Re Improved TSs revised.W/960617 Ltr
ML17264A5051996-05-17017 May 1996
LER 96-003-01:on 960308,identified That Both Pressurizer PORVs Inoperable Concurrently Due to Disconnection of Flex Hose to Both PORV Actuators to Install air-sets for Benchset & Limit Switch Activities.Hpes Completed
ML17264A4481996-04-0808 April 1996
LER 96-003-00:on 960308,both Pressurizer Relief Valves Inoperable.Hpes Evaluation Is Being Conducted to Determined Cause of Event.C/As:Both PORVs restored.W/960408 Ltr
ML17264A4471996-04-0808 April 1996
LER 96-002-00:on 960307,secondary Transient Occurred.Caused by Loss of B Condenser Circulating Water Pump.C/As: Thermography performed.W/960408 Ltr
ML17264A4101996-03-18018 March 1996
LER 96-001-00:on 950504,inservice Test Not Performed During Refueling Outage.Caused by Inadequate Tracking of Surveillance Frequency.Valve Test Performed & Disassembled. W/960318 Ltr
ML17264A2971995-12-14014 December 1995
LER 95-009-00:on 950817,surveillance Was Not Performed Due to Improper Application of TS Requirements Resulting in TS Violation.Testing of MOV-515 Was Performed on 951115.W/ 951214 Ltr
ML17264A1711995-09-25025 September 1995
LER 95-008-00:on 950825,secondary Transient Occurred.Caused by Loss of B Condenser Circulating Water Pump That Resulted in Manual Rt.Returned S/G Levels to Normal Operating levels.W/950925 Ltr
1999-09-22
[Table view]
Category:RO)
MONTHYEARML17265A7541999-09-22022 September 1999
LER 99-011-00:on 990823,small Tears Were Discovered in Flexible Duct Work Connector at Inlet of CR HVAC Sys Return Air Fan (AKF08).Caused by in-leakage Greater than That Assumed.Implemented Temporary Mod 99-029.With 990922 Ltr
ML17265A7431999-08-24024 August 1999
LER 99-004-01:on 990412,discovered That Containment Recirculation Fan Chevron Separator Vanes Were Installed Backwards.Caused by Improper Assembly by Mfg.Moisture Separator Vanes Were Dismantled & Correctly re-installed
ML17265A7181999-07-23023 July 1999
LER 99-007-01:on 990423,reactor Trip Occurred Due to Instrument & Control Technicians Inadvertently Pulling Fuses from Wrong Nuclear Instrument Channel.Setpoint Adjustments Were Completed by Different Crew of Technicians
ML17265A7081999-07-22022 July 1999
LER 98-003-02:on 980904,actuations of CR Emergency Air Treatment Sys Was Noted Due to Invalid Causes.Caused by Various Degraded Components in CR RM Sys.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored
ML17265A7031999-07-19019 July 1999
LER 99-S01-00:on 990617,determined That Temporary Unescorted Access Had Been Granted to Contractor Employee.Caused by Incomplete Info Re Circumstances of Individual Military Separation.Individual Access Was Revoked.With 990719 Ltr
ML17265A7021999-07-15015 July 1999
LER 99-010-00:on 990615,ventilation Isolation of Auxiliary Bldg Occurred When Auxiliary Bldg Gas Radiation Monitor R-14 Reached High Alarm Setpoint.Cr Operators Rest Auxiliary Bldg Ventilation Isolation Signal.With 990715 Ltr
ML17265A6851999-06-21021 June 1999
LER 99-001-01:on 990222,deficiencies in NSSS Vendor steam- Line Brake Mass & Energy Release Analysis Results in Plant Being Outside Design Bases Occurred.Caused by Deficiencies in W.Temporary Administrative Replaced.With 990621 Ltr
ML17265A6661999-06-0202 June 1999
LER 99-009-00:on 990503,instrumentation Declared Inoperable in Multiple Channels Resulted in Condition Prohibited by Ts. Caused by Unanticipated High Frequency AC Voltage Ripple. Entered TS LCO 3.0.3.With 990602 Ltr
ML17309A6541999-05-27027 May 1999
LER 99-008-00:on 990427,overtemperature Delta T Reactor Trip Occurred Due to Faulted Bistable During Calibr of Redundant Channel.Plant Was Stabilized in Mode 3 & Faulted Bistable Was Subsequently Replaced.With 990527 Ltr
ML17265A6631999-05-24024 May 1999
LER 99-007-00:on 990423,technicians Inadvertently Pulled Fuses from Wrong Nuclear Instrument Cahnnel,Causing Reactor Trip,Due to High Range Flux Trip.Caused by Personnel Error. Labeling Scheme Improved
ML17265A6601999-05-21021 May 1999
LER 99-006-00:on 990421,start of turbine-driven Auxiliary Feedwater Pump Was Noted.Caused by MOV Being Left in Open Position.Closed Manual Isolation Valve to Secure Steam to Pump.With 990521 Ltr
ML17265A6441999-05-13013 May 1999
LER 99-005-00:on 990413,undervoltage Signal of Safeguards Bus During Testing Resulted in Automatic Start of B Edg. Caused by Personnel Error.Blown Fuse Was Replaced & Offsite Power Was Restored to Safeguards Bus 17.With 990513 Ltr
ML17265A6431999-05-12012 May 1999
LER 99-004-00:on 990412,discovered That Containment Recirculation Fan Moisture Separator Vanes Were Incorrectly Installed,Per 10CFR21.Caused by Improper Assembly by Mfg. Subject Vanes Were Dismantled & Correctly re-installed
ML17265A6141999-03-31031 March 1999
LER 99-003-00:on 990301,two Main Steam non-return Check Valves Were Declared Inoperable Due to Exceedance of Acceptance Criteria.Caused by Changes in Methodology & Matls.Packing Gland Torque Will Be Adjusted.With 990331 Ltr
ML17265A6131999-03-29029 March 1999
LER 99-002-00:on 990227,discovered That Surveillance Had Not Been Performed at Frequency,Per Ts.Caused by Personnel Error.Procedure O-6.13 Will Be Evaluated for Enhancement Documentation of Completion of ITS Srs.With 990329 Ltr
ML17265A6061999-03-24024 March 1999
LER 99-001-00:on 990222,plant Was Noted Outside Design Basis.Caused by Deficiencies in NSSS Vendor Slb Mass & Energy Release.Placed Temporary Administrative Restriction 40 Degrees F Max on Screenhouse Bay Temp
ML17265A4951998-12-21021 December 1998
LER 98-005-00:on 981120,loss of 34.5 Kv Offsite Power Circuit 751,resulted in Automatic Start of B Edg.Caused by Faulted Cable Splice.Performed Appropriate Actions of Abnormal Procedure AP-ELEC.1.With 981221 Ltr
ML17265A4931998-12-17017 December 1998
LER 98-004-00:on 971030,determined That Improperly Performed Surveillance Resulted in Condition Prohibited by Ts.Caused by Procedure non-adherence.Appropriate Calibr Procedures Were Properly Performed with 24 H of Condition Discovery
ML17265A4691998-11-25025 November 1998
LER 98-003-01:on 980904,actuations of CR Emergency Air Treatment Systems (Creats) Occurred.Caused by Radon build-up During Temp Inversion.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored to CR
ML17265A4271998-10-0505 October 1998
LER 98-003-00:on 980904,actuations of CR Emergency Air Treatment Sys Occurred.Caused by Radon build-up During Temp Inversion.Air Samples Were Taken & Determined That Source of Radiation Was Naturally Occurring Radon.With 981005 Ltr
ML17265A3671998-07-14014 July 1998
LER 98-002-00:on 971019,CR Emergency Air Treatment Sys Actuating Function Was Not Operable.Caused by Mispositioned Switch.Revised Procedure CPI-MON-R37.W/980714 Ltr
ML17265A1921998-03-11011 March 1998
LER 98-001-00:on 980209,discovered That Boraflex Degradation in SPF Was Greater than Was Assumed.Caused by Dissolution of Boron on Boraflex Matrix,Per 10CFR50.21.Removed Spent Fuel Assemblies from Selected Degraded Storage Rack Cells
ML17265A1641998-02-0606 February 1998
LER 97-007-01:on 971117,reactor Engineer Recognized That Neutron Flux Low Range Trip Circuitry for Channel Was Not in Tripped Condition as Required.Caused by Technical Inadequacies.Channel Defeat Will Be Identified
ML17265A1601998-02-0606 February 1998
LER 97-006-01:on 971103,verification of B Concentration Was Not Performed Due to Misinterpretation of Event Sequence. Audible Count Rate Function Was Restored to Operable Status
ML17264B1441997-12-17017 December 1997
LER 97-007-00:on 971117,NF Low Range Trip Circuitry for Channel N-44 Was Not Placed in Tripped Condition.Caused by Technical Inadequacies in Procedures.Implemented EWR 4862 to Resolve Design deficiency.W/971217 Ltr
ML17264B1291997-12-0303 December 1997
LER 97-006-00:on 971103,NIS Audible Count Rate Function Was Inoperable.Caused by Misinterpretation of Event Sequence Due to Not Verifying Boron Concentration.B Verification Occurred Every 12 H Per ITS LCO Action 3.9.2.C.3.W/971203 Ltr
ML17264B1271997-12-0101 December 1997
LER 97-005-00:on 971031,undetected Unblocking of SI Actuation Signal Occurred at Low Pressure Condition,Due to Faulty Bistable Which Resulted in Inadvertent SI Actuation Signal.Sias,Ci & CVI Signals Were Reset
ML17264B1211997-11-24024 November 1997
LER 97-004-00:on 971024,radiation Monitor Alarm Were Noted Due to Higher than Normal Radioactive Gas Concentration Resulted in Cvi.New R-12 Alarm Setpoint Was Maintained for Duration of Refueling Outage
ML17264B0461997-09-29029 September 1997
LER 97-003-01:on 970730,bistable Instrument Trip Setpoint Could Have Exceeded Allowable Value.Caused by Insufficient Existing Margin Between Trip Setpoint & Allowable Value. Held Switches in Tripped configuration.W/970929 Ltr
ML17264B0111997-08-27027 August 1997
LER 97-003-00:on 970730,high Steam Flow Bistable Instrument Setpoint Plus Instrument Uncertainty Could Exceed Allowable Value in ITS Was Identified.Caused by Entry Into ITS LCO 3.0.3.Switches Placed in Tripped configuration.W/970827 Ltr
ML17264A9941997-08-19019 August 1997
LER 97-002-00:on 970720,34.5 Kv Offsite Power Circuit 751 Was Lost.Caused by Automatic Actuation of B Emergency DG Due to Undervoltage on Safeguards Buses 16 & 17.Offsite Power Restored to Safeguards Buses 16 & 17.W/970819 Ltr
ML17264A9911997-08-11011 August 1997
LER 96-009-02:on 960723,determined That Leak Rate Outside Containment Was Greater than Program Limit.Caused by Weld Defect.Isolated Leak & Cut Out & Replaced Leaking Pipe
ML17264A8271997-03-0303 March 1997
LER 97-001-00:on 970131,discovered Service Water Temp Was Less than Specified Value.Caused by non-representative Method of Monitoring.Increased Water Temp in Screenhouse Bay to Greater than 35 Degrees F.W/970303 Ltr
ML17264A8071997-01-22022 January 1997
LER 96-015-00:on 961223,discovered Thermally Induced Overpressure Transient Could Occur.Caused by Thermal Expansion of Fluid During Design Basis Accident Condition. Installed Relief Valve on Affected line.W/970122 Ltr
ML17264A7471996-11-27027 November 1996
LER 96-013-00:on 961029,circuit Breakers Closed While in Mode 3 & Resulted in Condition Prohibited by TS Due to Personnel Error.Circuit Breakers for MOV-878B & MOV-878D Were re-opened.W/961127 Ltr
ML17264A6051996-09-19019 September 1996
LER 96-012-00:on 960820,feedwater Transient Occurred,Due to Closure of Feedwater Regulating Valve,Causing Lo Lo Steam Generator Level Reactor Trip.Sgs Were Restored & Missing Screw in 1/P-476 Was replaced.W/960919 Ltr
ML17264A6061996-09-19019 September 1996
LER 96-009-01:on 960723,leakage Outside Containment Occurred,Due to Weld Defect,Resulting in Leak Rate Greater than Program Limits.Source of Leakage Isolated from RWST by Freeze Seal,Allowing Exit from ITS LCO 3.0.3.W/960919 Ltr
ML17264A5911996-09-0505 September 1996
LER 96-011-00:on 960807,improper Configuration of Circuit Breaker Occurred,Due to Undetected Internal Interference, Resulting in Automatic Start of Both Auxiliary Feedwater Pumps.Running AFW Pumps Were secured.W/960905 Ltr
ML17264A5921996-09-0505 September 1996
LER 96-010-00:on 960806,latching of Main Turbine While in Mode 4 Occurred,Due to Defective Procedure,Resulting in Automatic Start of Auxiliary Feedwater Pump.Caused by Defective Maint Procedure.Procedure revised.W/960905 Ltr
ML17264A5891996-08-22022 August 1996
LER 96-009-00:on 960723,determined Leak on Piping Sys Outside Containment Greater than Program Limit.Caused by Weld Defect.Pipe & Socket Welds Were Cut Out & Replaced. W/960822 Ltr
ML17264A5781996-08-0606 August 1996
LER 96-008-00:on 960707,main Feedwater Pump Breakers Opened. Caused by Change in Seal Water Differential Pressure Occurred During Sys Realignment.Afw Flow Controlled as Desired to Maintain S/G level.W/960806 Ltr
ML17264A5561996-07-12012 July 1996
LER 96-007-00:on 960612,CR Operators Identified Control Rods Misaligned & Not Moving in Proper Sequence.Caused by Faulty Firing Circuit Card in Rod Control Sys.Faulty Firing Circuit Card in 1BD Power Cabinet replaced.W/960712 Ltr
ML17264A5421996-06-20020 June 1996
LER 96-006-00:on 960521,discovered Containment Penetration Not in Required Status.Caused by Personnel Error.Installed Flange Inside Containment Penetration 2.W/960620 Ltr
ML17264A5411996-06-17017 June 1996
LER 96-005-00:on 960516,PORC Determined Deficient Procedures Do Not Meet SRs for Testing safety-related Logic Circuits. Caused by Inadequancies in Individual Testing Procedures. Procedures Re Improved TSs revised.W/960617 Ltr
ML17264A5051996-05-17017 May 1996
LER 96-003-01:on 960308,identified That Both Pressurizer PORVs Inoperable Concurrently Due to Disconnection of Flex Hose to Both PORV Actuators to Install air-sets for Benchset & Limit Switch Activities.Hpes Completed
ML17264A4481996-04-0808 April 1996
LER 96-003-00:on 960308,both Pressurizer Relief Valves Inoperable.Hpes Evaluation Is Being Conducted to Determined Cause of Event.C/As:Both PORVs restored.W/960408 Ltr
ML17264A4471996-04-0808 April 1996
LER 96-002-00:on 960307,secondary Transient Occurred.Caused by Loss of B Condenser Circulating Water Pump.C/As: Thermography performed.W/960408 Ltr
ML17264A4101996-03-18018 March 1996
LER 96-001-00:on 950504,inservice Test Not Performed During Refueling Outage.Caused by Inadequate Tracking of Surveillance Frequency.Valve Test Performed & Disassembled. W/960318 Ltr
ML17264A2971995-12-14014 December 1995
LER 95-009-00:on 950817,surveillance Was Not Performed Due to Improper Application of TS Requirements Resulting in TS Violation.Testing of MOV-515 Was Performed on 951115.W/ 951214 Ltr
ML17264A1711995-09-25025 September 1995
LER 95-008-00:on 950825,secondary Transient Occurred.Caused by Loss of B Condenser Circulating Water Pump That Resulted in Manual Rt.Returned S/G Levels to Normal Operating levels.W/950925 Ltr
1999-09-22
[Table view]
Category:TEXT-SAFETY REPORT
MONTHYEARML17265A7601999-10-0505 October 1999
Part 21 Rept Re W2 Switch Supplied by W Drawn from Stock, Did Not Operate Properly After Being Installed on 990409. Switch Returned to W on 990514 for Evaluation & Root Cause Analysis
ML17265A7621999-09-30030 September 1999
Monthly Operating Rept for Sept 1999 for Re Ginna Npp.With 991008 Ltr
ML17265A7531999-09-23023 September 1999
Part 21 Rept Re Corrective Action & Closeout of 10CFR21 Rept of Noncompliance Re Unacceptable Part for 30-4 Connector. Unacceptable Parts Removed from Stock & Scrapped
ML17265A7541999-09-22022 September 1999
LER 99-011-00:on 990823,small Tears Were Discovered in Flexible Duct Work Connector at Inlet of CR HVAC Sys Return Air Fan (AKF08).Caused by in-leakage Greater than That Assumed.Implemented Temporary Mod 99-029.With 990922 Ltr
ML17265A7471999-08-31031 August 1999
Monthly Operating Rept for Aug 1999 for Re Ginna Npp.With 990909 Ltr
ML17265A7431999-08-24024 August 1999
LER 99-004-01:on 990412,discovered That Containment Recirculation Fan Chevron Separator Vanes Were Installed Backwards.Caused by Improper Assembly by Mfg.Moisture Separator Vanes Were Dismantled & Correctly re-installed
ML17265A7341999-07-31031 July 1999
Monthly Operating Rept for July 1999 for Re Ginna Npp.With 990806 Ltr
ML17265A7291999-07-29029 July 1999
Interim Part 21 Rept Re safety-related DB-25 Breaker Mechanism Procured from W Did Not Pas Degradatin Checks When Drawn from Stock to Be Installed Into BUS15/03A.Holes Did Not line-up & Tripper Pan Bent
ML17265A7181999-07-23023 July 1999
LER 99-007-01:on 990423,reactor Trip Occurred Due to Instrument & Control Technicians Inadvertently Pulling Fuses from Wrong Nuclear Instrument Channel.Setpoint Adjustments Were Completed by Different Crew of Technicians
ML17265A7081999-07-22022 July 1999
LER 98-003-02:on 980904,actuations of CR Emergency Air Treatment Sys Was Noted Due to Invalid Causes.Caused by Various Degraded Components in CR RM Sys.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored
ML17265A7131999-07-22022 July 1999
Special Rept:On 990407,radiation Monitor RM-14A Was Declared Inoperable.Caused by Failed Communication Link from TSC to Plant Process Computer Sys.Communication Link Was re-established & RM-14A Was Declaed Operable on 990521
ML17265A7031999-07-19019 July 1999
LER 99-S01-00:on 990617,determined That Temporary Unescorted Access Had Been Granted to Contractor Employee.Caused by Incomplete Info Re Circumstances of Individual Military Separation.Individual Access Was Revoked.With 990719 Ltr
ML17265A7211999-07-19019 July 1999
ISI Rept for Third Interval (1990-1999) Third Period, Second Outage (1999) at Re Ginna Npp.
ML17265A7021999-07-15015 July 1999
LER 99-010-00:on 990615,ventilation Isolation of Auxiliary Bldg Occurred When Auxiliary Bldg Gas Radiation Monitor R-14 Reached High Alarm Setpoint.Cr Operators Rest Auxiliary Bldg Ventilation Isolation Signal.With 990715 Ltr
ML17265A7661999-06-30030 June 1999
1999 Rept of Facility Changes,Tests & Experiments Conducted Without Prior NRC Approval for Jan 1998 Through June 1999, Per 10CFR50.59.With 991020 Ltr
ML17265A7011999-06-30030 June 1999
Monthly Operating Rept for June 1999 for Re Ginna Npp.With 990712 Ltr
ML17265A6851999-06-21021 June 1999
LER 99-001-01:on 990222,deficiencies in NSSS Vendor steam- Line Brake Mass & Energy Release Analysis Results in Plant Being Outside Design Bases Occurred.Caused by Deficiencies in W.Temporary Administrative Replaced.With 990621 Ltr
ML17265A6761999-06-16016 June 1999
Part 21 Rept Re Defects & noncompliances,10CFR21(d)(3)(ii), Which Requires Written Notification to NRC on Identification of Defect or Failure to Comply. Relays Were Returned to Eaton for Evaluation & Root Cause Analysis
ML17265A6661999-06-0202 June 1999
LER 99-009-00:on 990503,instrumentation Declared Inoperable in Multiple Channels Resulted in Condition Prohibited by Ts. Caused by Unanticipated High Frequency AC Voltage Ripple. Entered TS LCO 3.0.3.With 990602 Ltr
ML17265A6681999-05-31031 May 1999
Monthly Operating Rept for May 1999 for Re Ginna Nuclear Power Plant.With 990608 Ltr
ML17265A6651999-05-27027 May 1999
Interim Rept Re W2 Control Switch,Procured from W,Did Not Operate Satisfactorily When Drawn from Stock to Be Installed in Main Control Board for 1C2 Safety Injection Pump. Estimated That Evaluation Will Be Completed by 991001
ML17309A6541999-05-27027 May 1999
LER 99-008-00:on 990427,overtemperature Delta T Reactor Trip Occurred Due to Faulted Bistable During Calibr of Redundant Channel.Plant Was Stabilized in Mode 3 & Faulted Bistable Was Subsequently Replaced.With 990527 Ltr
ML17265A6631999-05-24024 May 1999
LER 99-007-00:on 990423,technicians Inadvertently Pulled Fuses from Wrong Nuclear Instrument Cahnnel,Causing Reactor Trip,Due to High Range Flux Trip.Caused by Personnel Error. Labeling Scheme Improved
ML17265A6601999-05-21021 May 1999
LER 99-006-00:on 990421,start of turbine-driven Auxiliary Feedwater Pump Was Noted.Caused by MOV Being Left in Open Position.Closed Manual Isolation Valve to Secure Steam to Pump.With 990521 Ltr
ML17265A6591999-05-17017 May 1999
Part 21 Rept Re Relay Deficiency Detected During pre-installation Testing.Caused by Incorrectly Wired Relay Coil.Relays Were Returned to Eaton Corp for Investigation. Relays Were Repaired & Retested
ML17265A6441999-05-13013 May 1999
LER 99-005-00:on 990413,undervoltage Signal of Safeguards Bus During Testing Resulted in Automatic Start of B Edg. Caused by Personnel Error.Blown Fuse Was Replaced & Offsite Power Was Restored to Safeguards Bus 17.With 990513 Ltr
ML17265A6431999-05-12012 May 1999
LER 99-004-00:on 990412,discovered That Containment Recirculation Fan Moisture Separator Vanes Were Incorrectly Installed,Per 10CFR21.Caused by Improper Assembly by Mfg. Subject Vanes Were Dismantled & Correctly re-installed
ML17265A6381999-05-0707 May 1999
Part 21 Rept Re Replacement Turbocharger Exhaust Turbine Side Drain Port Not Functioning as Design Intended.Caused by Manufacturing Deficiency.Turbocharger Was Reaasembled & Reinstalled on B EDG
ML17265A6391999-04-30030 April 1999
Monthly Operating Rept for Apr 1999 for Re Ginna Nuclear Power Plant.With 990510 Ltr
ML17265A6361999-04-23023 April 1999
Part 21 Rept Re Power Supply That Did Not Work Properly When Drawn from Stock & Installed in -25 Vdc Slot.Power Supply Will Be Sent to Vendor to Perform Failure Mode Assessment.Evaluation Will Be Completed by 991001
ML17265A6301999-04-18018 April 1999
Rev 1 to Cycle 28 COLR for Re Ginna Npp.
ML17265A6251999-04-15015 April 1999
Special Rept:On 990309,halon Systems Were Removed from Svc & Fire Door F502 Was Blocked Open.Caused by Mods Being Made to CR Emergency Air Treatment Sys.Continuous Fire Watch Was Established with Backup Fire Suppression Equipment
ML17265A6551999-04-0909 April 1999
Initial Part 21 Rept Re Mfg Deficiency in Replacement Turbocharger for B EDG Supplied by Coltec Industries. Deficiency Consisted of Missing Drain Port in Intermediate Casing.Required Oil Drain Port Machined Open
ML17265A6291999-03-31031 March 1999
Rev 0 to Cycle 28 COLR for Re Ginna Npp.
ML17265A6241999-03-31031 March 1999
Monthly Operating Rept for Mar 1999 for Ginna Station.With 990409 Ltr
ML17265A6141999-03-31031 March 1999
LER 99-003-00:on 990301,two Main Steam non-return Check Valves Were Declared Inoperable Due to Exceedance of Acceptance Criteria.Caused by Changes in Methodology & Matls.Packing Gland Torque Will Be Adjusted.With 990331 Ltr
ML17265A6131999-03-29029 March 1999
LER 99-002-00:on 990227,discovered That Surveillance Had Not Been Performed at Frequency,Per Ts.Caused by Personnel Error.Procedure O-6.13 Will Be Evaluated for Enhancement Documentation of Completion of ITS Srs.With 990329 Ltr
ML17265A6061999-03-24024 March 1999
LER 99-001-00:on 990222,plant Was Noted Outside Design Basis.Caused by Deficiencies in NSSS Vendor Slb Mass & Energy Release.Placed Temporary Administrative Restriction 40 Degrees F Max on Screenhouse Bay Temp
ML17265A5661999-03-0101 March 1999
Rev 26 to QA Program for Station Operation.
ML17265A5961999-02-28028 February 1999
Monthly Operating Rept for Feb 1999 for Ginna Nuclear Power Plant.With 990310 Ltr
ML17265A5371999-01-31031 January 1999
Monthly Operating Rept for Jan 1999 for Re Ginna Nuclear Power Plant.With 990205 Ltr
ML17265A5951998-12-31031 December 1998
Rg&E 1998 Annual Rept.
ML17265A5001998-12-21021 December 1998
Rev 26 to QA Program for Station Operation.
ML17265A4951998-12-21021 December 1998
LER 98-005-00:on 981120,loss of 34.5 Kv Offsite Power Circuit 751,resulted in Automatic Start of B Edg.Caused by Faulted Cable Splice.Performed Appropriate Actions of Abnormal Procedure AP-ELEC.1.With 981221 Ltr
ML17265A4931998-12-17017 December 1998
LER 98-004-00:on 971030,determined That Improperly Performed Surveillance Resulted in Condition Prohibited by Ts.Caused by Procedure non-adherence.Appropriate Calibr Procedures Were Properly Performed with 24 H of Condition Discovery
ML17265A4761998-11-30030 November 1998
Monthly Operating Rept for Nov 1998 for Re Ginna Nuclear Power Plant.With 981210 Ltr
ML17265A4691998-11-25025 November 1998
LER 98-003-01:on 980904,actuations of CR Emergency Air Treatment Systems (Creats) Occurred.Caused by Radon build-up During Temp Inversion.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored to CR
ML17265A4531998-10-31031 October 1998
Monthly Operating Rept for Oct 1998 for Re Ginna Nuclear Power Plant.With 981110 Ltr
ML17265A4271998-10-0505 October 1998
LER 98-003-00:on 980904,actuations of CR Emergency Air Treatment Sys Occurred.Caused by Radon build-up During Temp Inversion.Air Samples Were Taken & Determined That Source of Radiation Was Naturally Occurring Radon.With 981005 Ltr
ML17265A4291998-09-30030 September 1998
Monthly Operating Rept for Sept 1998 for Re Ginna Nuclear Power Plant.With 981009 Ltr
1999-09-30
[Table view] |
Text
ACCELERATED DISJBUTION DEMONS~ION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)ACCESSION NBR:9005160332 DOC.DATE: 90/05/10 NOTARIZED:
NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH.NAME AUTHOR AFFILIATION GORSKI,P.Rochester Gas&Electric Corp.MECREDY,R.C.
Rochester Gas&Electric Corp.RECIP.NAME RECIPIENT AFFILIATION
SUBJECT:
LER 90-004-00:on 900416,steam generator tube degradation due to IGA/SCC causes QA manual reportable limits to be reached.W/9 ltr.DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR 2 ENCL L SIZE: TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).
05000244 RECIPIENT ID CODE/NAME PDl-3 LA JOHNSON,A INTERNAL: ACNW AEOD/DSP/TPAB DEDRO NRR/DET/EMEB9H3 NRR/DLPQ/LPEB10 NRR/DREP/PRPB11 NRR/DST/SICB 7E NRR/DST/SRXB SE RES/DSIR/EIB EXTERNAL: EG&G STUART,V.A LPDR NSIC MAYS,G NUDOCS FULL TXT COPIES LTTR ENCL 1" 1 1 1 2 2 1 1 1 1 1 1 1 1 2 2 1 1 1.1 1 1 4 4 1 1 1 1 1 1 RECIPIENT ID CODE/NAME PD1-3 PD AEOD/DOA AEOD/ROAB/DSP NRR/DET/ECMB 9H NRR/DLPQ/LHFB11 NRR/DOEA/OEAB11 NRR/DST/SELB SD NRRQBSQ'LBSD1
, EG FILE 02 R E 01 L ST LOBBY WARD NRC PDR NSIC MURPHYiG A COPIES LTTR ENCL 1 1 ,1 1'2 2 1 1 1 1 1 1~1 1 1, 1 1 1 1 1 1 1 1 1 1 1 P>govs 7MB 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 FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 34 ENCL 34
Q sCe a~5.~r~15I+A 55ACC ROCHESTER GAS AND ELECTRIC CORPORATION
~89 EAST AVENUE, ROCHESTER, N.Y.14649-0001 TCLCa~OhC,*RCA COOt.1III 54G.2700 May 10, 1990 U.S.Nuclear Regulatory Commission Document Control Desk Washington, DC 20555
Subject:
LER 90-004, Steam Generator Tube Degradation Due To IGA/SCC Causes Q.A.Manual Reportable Limits To Be Reached R.E.Ginna Nuclear Power Plant Docket No.50-244 In accordance with 10 CFR 50.73, Licensee Event Report System, item (Other), and the Ginna Station Quality,.Assurance'anual Appendix B;which requires that,"If the number of tubes in a generator falling into categories (a)or (b)exceeds the criteria, then results of the inspection shall be considered a Reportable Event pursuant to 10 CFR 50.73", the attached Licensee Event Report LER 90-004 is hereby submitted.
This event has in no way affected the public's health and safety.Ve truly yours, o ert C.M redy Division Manager Nuclear Production xco U.S.Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna USNRC Senior Resident Inspector yA555 pgiP
NAC tea)00 flOII LICENSEE EVENT REPORT{LER)VS NUCLCAA AIOVLATOARCOLNIIHION APPAOVIO OLII NO.IIIO OIOH cxf)At~~IIIIII PACILITY NAME lll R.E.Ginna Nuclear Power Plant OOCXI'1 NUMIIA Nl o 6 o o 024 co(07"'"'" Steam Generator Tube Degradation Due To IGA/SCC Causes Q.A.Manual Reportable Limits To Be Reached EVINT OATI III LtA NVMICA (4 AttOAT OATt ltl OTHER IACILITIII INVOLV EO III UOHTN OAY YCAA YEAH IEOUINZIAL
" AIvrjON NUUI I A sf HUU~~A MONTH OAY YIAA PACILITY H*ULI OOCXtT NVUIIAI4 0 5 0 0 0 0 4 9 0 9 0 004 00 05 090 0 s 0 0 0 Otl A AT I HO MOO C Ili~Owtx LlvlL Q 0 Q II II II.LOIILI II.LOI41(I I(0 IOAOI4)lll(4 IIAOIIL)llllw)IOAOI4)(1)
IHI IO.H)04)(1)l
~)10.MN(h IO JI(a)II I IOM lelNI~0.114IN)II 00.114)N)lll IO.11 4 IN)I wl IO.)14 IN)(HI~0.114)N I(h IO.1 14 IN)IMI I0.1 14 IN)I tW)IA)~O.f 14)(11(HI)Ill IO.TI(IIN)la) 0 THC AIOVIAIMENTI Ot 10CPA$: IOUct Art W re tt IAP IUNrUPI 111 THII ACPOAT II IUILIITTCO tUAIVANT 1 OT)IEA IIALtltt 4 AUYrtt~rpr tp0 M TML HAC Pprr JIIA)HALIC Paul Gorski Mechanical Maintenance Manager LICCNIII CONTACT POA TH)I LIA l11)*AIA COOt TELEPHONE HUMICA 1552-4446 COMtLCTI ONt LINt fOA EACH COLIPOHtNT I AILUAC OIICAIItO IN THII AIPOAT (11(CAUlt IYCTIM COMPO H I HT MAHUf AC TVAEA COMPOHCHT MAHVfAC TVAEA EPOATAIL TO HtADI l~4+8ggjr~eq,Sg IVttLCMlHTAL AltOAT CxtlCTCO l(4 It~I>'+%/K<"'MONTH CAY YCAA txtCCTtO lUILNCCIOH OATI (I~I'YCI l)t tw.~N CIPICTCO IUIUIIJIOH OATCI X HO LAITA ACT I@cut IP I t00 NWN, I I, tyyntrr lelt AlrrL alLPtmCt nttrrINA VM (14 During the 1990 Annual Refueling and Maintenance Outage subsequent to the eddy current examination performed on both the"A" and"B" Westinghouse Series 44 Steam Generators (S/G), 75 tubes in the"A" S/G and 211 tubes in the"B" S/G required corrective action due to tube degradation.
This defect population includes 28 tubes in the"B" S/G that had known defects plugged in prior outages.These tubes were unplugged for full length eddy current examination and were returned to service with a sleeve repair in the degraded region.The immediate cause of the event was that the"A" and"B" S/G tube degradation was in excess of the Ginna Quality Assurance Manual reportability limits.The underlying cause of the tube degradation is a common S/G problem of a partially rolled tube sheet crevice with recurring Zntergranular Attack/Stress Corrosion Cracking (EGA/SCC)and Primary Water Stress Corrosion Cracking (PWSCC)attack on S/G tubing.Corrective action taken was to either sleeve or plug the affected tubes with accepted industry repair methods.NAC ters III IIIII
NRC Form 944A l943I LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S, NUCLEAR REOULATORY COMMISSION APPROVEO OME NO 3140&l04 EXPIRES'ISI/SS FACILITY NAME>>I OOCKET NUMEER Ql LER NUMEER (41 wE::, SEOVENTIAL NVM EA AEVOIOH MVM ER~AOE IS)R.E.Ginna Nuclear Power Plant TEXT IIF uuuu<<MPI<<newt'.uup<<PPppA<<IYIIC Avw SOSASI I I TI 0 5 0 0 0 2 4 4 90-0 04-000 2 OF 0 7 PRE-EVENT PLANT CONDITIONS The unit was in cold/refueling shutdown for the Annual Refueling Maintenance Outage.All fuel had been removed from the Reactor Vessel.Steam Generator eddy current inspection was in progress.DESCRIPTION OP EVENT A.DATES AND APPROXIMATE TIMES FOR MAJOR OCCURRENCES:
o April 16, 1990, 1711 EDST: Event date and time.~t>o.':pril 16, 1990, 1711 EDST: Discovery date and".;.me.o:.ril 18, 1990, 1000 EDST: Oral notification made to the NRC Office of Nuclear Reactor."Regulation (NRR).o A--.ril 21, 1990, 1800 EDST: Steam Generator rap'rs completed.
o May 1, 1990: Followup report sent to NRC Office of Nuclear Reactor Regulation (NRR).B.EVENT: During the 1990 Annual Refueling and Maintenance Outage, an eddy current examination was performed in both the>>A>>and>>B>>Westinghouse Series 44 Design recirculating steam generators.
The purpose of the eddy current examination was to assess any corrosion or mechanical damage that may have occurred during the cycle since May 1989.'AC FORM 944A 1943 I NRC Perm SSBA (9451 LICENSEE EVENT REPORT (LERI TEXTCONTINUATION V.S.NUCLEAR REOULATORY COMMISSION APPROVED OMB NO 5150&(95 ExPIRES: BISI/BS PACILITY NAME (11 DOCKET NUMBER (l(LER NUMBER (51 55OVENTIAL NVM REVISION I4vM eo~AOE (SI R.E..Ginna Nuclear Power Plant TEXT lll more 5oece r reeeeerf.Pee~Arr(C forrrr JRLI Sl (Ill o s o o o 2 4 4 90-004 00 03 oF 07 The examination was performed by personnel from Rochester Gas and Electric Corporation (RG&E)f and Allen Nuclear Associates.
All personnel had been trained and qualified in the eddy current examination method and had been certified to a minimum of Level I for data acquisition and Level II for data analysis.The eddy current examination of the"A" and"B" steam generators was performed utilizing the Zetec MIZ-18 Digital Data Acquisition System.The frequencies selected were 400, 200, 100, and 25 KHZ.The inlet or hot leg examination program plan was generated to provide the examination of 1004 of each open (not sleeved, or plugged)steam generator tube from the tube end to the first tube support.Zn addition, 204 of these tubes were selected and examined for their full length as recommended in the EPRZ guidelines.
All tubes with previous indications greater than 20%through wall (TW)depth were examined at a minimum to the location of their degradation.
Approximately 20%of all open Row 1 U-bend regions were examined with the Motorized Rotating Pancake Coil (MRPC)between the g6 Tube Support Plate Hot (TSPH)and the 56 Tube Support Plant Cold (TSPC)from the cold leg side.Results of the above inspections indicated that 75 tubes in the"A" steam generator and 211 tubes in the"B" steam generator (183 new repairs plus 28 previously plugged tubes)required corrective action.C.INOPERABLE STRUCTURES f COMPONENTS f OR SYSTEMS THAT CONTRIBUTED TO THE EVENT: D.None.OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED: None.E.METHOD OF DISCOVERY:
The event was apparent after the review of the eddy current examination results.NRC POIIM 555A (9451 NRC Form 844A 19')5 I D LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S.NUCLEAR REOULATORY COMMISSION APPROVED OM8 NO S)50&104 EXPIRES.8181/85 FACILITY NAME 111 R.E.Ginna Nuclear Power Plant TEXT lll mcpp N>>ce N>>EMcpd.p>>cHtcNanV HRC%%dmc JR)A 9)I IT)DOCKET NUMSER Ill o 5 o o o 244 I.ER NUMSER 14)54OVSNTIAL 45.EVIS o NVMSSR-.I" MVM 90-004 00 PACE (8)4 oF 0 7 F.OPERATOR ACTION: None.G.SAFETY SYSTEM RESPONSES:
None.III.CAUSE OF EVENT A.IMMEDIATE CAUSE: The Immediate Cause of the event was that the<<A<<and'B<<steam generator tube degradation was in excess of the Ginna Quality Assurance Manual Reportable Limits.B.ROOT CAUSE: The results of the examination indicate that the ZGA and SCC continue to be active within the tube sheet crevice region on the inlet side of each steam generator.
As in the past, the IGA/SCC is much more prevalent in the<<B<<steam generator with 108 ZGA indications and 49 SCC indications reported.Zn the<<A<<steam generator 16 ZGA indications and 22 SCC indications were reported.The majority of the inlet tube sheet crevice corrosion indications are IGA/SCC of the Mil-annealed Inconel 600 Tube Material.This form of corrosion is believed to be the result of the tube sheet crevices forming an alkaline environment.
This environment has developed over the years as deposits and active species have reacted with sodium and phosphate, changing a neutral or inhibited crevice into the aggressive environment that presently exists.In addition to the IGA/SCC in the crevices PWSCC indications continue to be found at the roll transi-tion.This year there were 23 PWSCC indications in the<<B<<steam generator and 37 PWSCC indications in the<<A<<.steam generator.
NAC FOIIM 545A 19851 HRC form 559A (9451 PACILITY NAME 111UCENSEE EVENT REPORT ILERI TEXT CONTINUATION I/.9.NI/CLEAR REGULATORY COMMIEEION APPROVEO OME I/O 515OPOIOA EXP/RES: 9/91/85 LER NUMEER 191 R.E.Ginna Nuclear Power Plant TEXT/ll mIVO Opoco IP Olvrror/, vro~/V/IC fo/III PS@'p/IITI o s o o o 2 4 4 vEAR 9 0 9 9 O I/C/IT I A L II I/II C R 0 0 4 R 9 V If IO Ir Ir I/II 9 A 000 5 OF 07 Ivo ANALYSIS OF EVENT The event is reportable in accordance with 10 CFR 50.73, Licensee Event Report Item (other)and the Ginna Station Quality Assurance Manual Appendix B which requires that,"If the number of tubes in a generator falling into'ategories (a)or (b)exceeds the criteria, then results of the inspection shall be considered a Reportable Event pursuant to 10 CFR 50.73." The tube degradation in the"A" and"B" steam generators exceeded the criterion of (b)which states,"More than 1%of the total tubes inspected are degraded, (imperfections greater than the repair limit)." This repair limit is defined as,"steam generator tubes that have imperfections greater than 40%through wall, as'ndicated by eddy current, shall be repaired by-,@ging or sleeving." An assessment was performed considering.the safety con-sequences and implications of this event with the following results and conclusions:
There were no operational or safety consequences or safety im",.'ations resulting from the steam generator tube dec;dation in excess of the Q.A.Manual reportable limits bee..use: o The degraded tubes were identified and repaired prior to any significant leakage or S/G tube rupture occurring.
o Even assuming a complete severance of a steam generator tube at full power, as stated in the R.E.Ginna Nuclear Power Plant Updated Final Safety Analysis Report (Ginna/UFSAR)
Section 15.6.3, (Steam-Generator Tube Rupture)the sequence, of recovery actions ensures early termination of primary to secondary leakage with or without offsite power available thus limiting offsite radiation doses to within the guidelines of 10 CFR 100.Based on the above, it can be concluded that the public's health and safety were assured at all times.NRC PORM POOA 19 EE I NRC Form SSSA 19431 IEACILITY NAME III LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.5.NUCLEAR REGULATORY COMMISSION APPROVEO OMS HO 5150WI05 EXPIRES SISI/95 LER NUMSER IS)R.E.Ginna Nuclear Power Plant TEXT Ill more opec@rp orrrrerF.1>>poooronolHlIC Form PxWSI IITI o s o o o2 44 90 55GUSNTrAL
~rUM 51 004 rr 5 Y l5 lO 1~rUrr 51-0 0 06oF 7 CORRECTIVE ACTION A.ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS: o Of the 75 degraded tubes in the"A" Steam Generator, 51 tubes were repaired using a Combustion Engineering welded 27" sleeve in the hot leg and these tubes will remain in service.The remaining 24 tubes were removed from active service by plugging both the hot and cold leg tube ends.o Of the 211 degraded tubes in the"B" steam'enerator, 191 tubes were repaired using a Combustion Engineering welded 27" sleeve in the hot leg and these tubes will remain in service.The remaining 20 tubes were removed from active service by plugging both the hot and cold leg tube ends.B.ACTION TAKEN OR PLANNED TO PREVENT RECUSANCE:
The occurrence/presence of IGA, SCC and PWSCC is a common PHR Steam Generator problem.Utile~"ies with susceptible tubing and partially rolled crevices must deal with this recurring attack on steam generator tubing.R.E.Ginna Station will continue careful monitoring of both primary RCS and secondary side water chemistry parameters.
These water chemistry parameters will be evaluated against accepted industry guidelines in order to minimize harmful primary and/or secondary side environments.
Degraded S/G tubes shall be sleeved or plugged in accordance with the inservice inspection program and accepted industry repair methods.~roc po1M 555k 19M I NRC FoIm 344A (943)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION V.S.NUCLEAR REGULATORY COMMISSION APPROVEO OMB NO, 3)50M)04 EXPIRESI 813)185 FACILITY NAME 111 OOCKET NUMBER Ill YEAR LER NUMBER IS)44OUENTIAL NUM SR REVISION NUM 4 II~AGE IS)R.E.Ginna Nuclear Power Plant TEXT llf mCVP g>>44 N fPqviNL MW~ff)IC Ann 3RLISl)IT)0 5 0 0 0 2 4 4 9 p 004 0 0 7 QF p ADDITIONAL INFORMATION A.FAILED COMPONENTS:
None.B.PREVIOUS LERs ON SIMILAR EVENTS: A similar LER event historical search was conducted with the following results: The crevice indications are similar to those reported in A0-74-02, A0-75-07, R0-75-013, and LER's76-008, 77-008,78-003, 79-006,79-022, 80-003,81-009, 82-003,82-022, 83-013, and 89-001.C.SPECIAL COMMENTS: None.NRC FORM 344A 1983) 0 E, J r
