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| {{#Wiki_filter:ACCELERATED DISTRIBUTION DEMONSTRATION SYSTEMREGULATORY INFORMATION DISTRIBUTION-SYSTEM(RIDS)ACCESSION NBR:9406030044 DOC.DATE: | | {{#Wiki_filter:ACCELERATED DISTRIBUTION DEMONSTRATION SYSTEM REGULATORY INFORMATION DISTRIBUTION-SYSTEM (RIDS)ACCESSION NBR:9406030044 DOC.DATE: 94/05/27 NOTARIZED: |
| 94/05/27NOTARIZED: | | NO FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G AUTH.NAME AUTHOR AFFILIATION ST.MARTIN,J.T. |
| NOFACIL:50-244 RobertEmmetGinnaNuclearPlant,Unit1,Rochester GAUTH.NAMEAUTHORAFFILIATION ST.MARTIN,J.T.
| | Rochester Gas&Electric Corp.MECREDY,R.C. |
| Rochester Gas&ElectricCorp.MECREDY,R.C. | | Rochester Gas&Electric Corp.RECIP.NAME RECIPIENT AFFILIATION I |
| Rochester Gas&ElectricCorp.RECIP.NAME RECIPIENT AFFILIATION I | |
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| ==SUBJECT:== | | ==SUBJECT:== |
| LER94-007-00:on 940427,feedwater transient occurredduetolossofabilitytocontrolfeedwater regulatinq valve,causingloloSGlevelreactortrip.Causedbyimproperly securedstrokeadjustsetscrew.W/940527 ltr.DISTRIBUTION CODE:IE22TCOPIESRECEIVED:LTR jENCL/SIZE:TITLE:50.73/50.9 LicenseeEventReport(LER),IncidentRpt,etc.DOCKETg05000244.RD8NOTES:License Expdateinaccordance with10CFR2,2.109(9/19/72).
| | LER 94-007-00:on 940427,feedwater transient occurred due to loss of ability to control feedwater regulatinq valve, causing lo lo SG level reactor trip.Caused by improperly secured stroke adjust set screw.W/940527 ltr.DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR j ENCL/SIZE: TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.DOCKET g 05000244.R D 8 NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). |
| 05000244ARECIPIENT IDCODE/NAME PD1-3PDINTERNAL:
| | 05000244 A RECIPIENT ID CODE/NAME PD1-3 PD INTERNAL: AEOD/DOA AEOD/ROAB/DSP NRR/DE/EMEB NRR/DRCH/HHFB NRR/DRCH/HOLB NRR/DRSS/PRPB NRR/DSSA/SRXB RES/DSIR/EIB I EXTERNAL: EG&G BRYCE,J.H NRC PDR NSIC POOREiW~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 JOHNSON,A AEOD/DSP/TPAB NRR/DE/EELB NRR/DORS/OEAB NRR/DRCH/HICB NRR/DRIL/RPEB NRR/~SSA/S PLB REGFFL~02 G 1 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 D D D NOTE TO ALL"RIDS" RECIPIENTS: |
| AEOD/DOAAEOD/ROAB/DSP NRR/DE/EMEB NRR/DRCH/HHFB NRR/DRCH/HOLB NRR/DRSS/PRPB NRR/DSSA/SRXB RES/DSIR/EIB IEXTERNAL: | | A D D PLEASE HELP US TO REDUCE WASTE!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 27 ENCL 27 A sr ROCHESTER GAS AND ELECTRIC CORPORATION ROBERT C.MECREOY Vice Preridr nr Cinna Crurlesr Producuon rr C~~~~v r 4'r roe rr Scare 89 EAST AVENUE, ROCHESTER N.Y.14649-0001 TELEPHONE AicEA coDE 716 546'2700 May 27, 1994 U.S.Nuclear Regulatory Commission Attn: Allen R.Johnson PWR Project Directorate I-3 Document Control Desk Washington, DC 20555 |
| EG&GBRYCE,J.H NRCPDRNSICPOOREiW~COPIESLTTRENCL111122111111221111221111RECIPIENT IDCODE/NAME JOHNSON,A AEOD/DSP/TPAB NRR/DE/EELB NRR/DORS/OEAB NRR/DRCH/HICB NRR/DRIL/RPEB NRR/~SSA/S PLBREGFFL~02G1FILE01LSTLOBBYWARDNSICMURPHY,G.A NUDOCSFULLTXTCOPIESLTTRENCL11111111111111111.11,11'11DDDNOTETOALL"RIDS"RECIPIENTS: | |
| ADDPLEASEHELPUSTOREDUCEWASTE!CONTACTTHEDOCUMENTCONTROLDESK,ROOMPl-37(EXT.20079)TOELIMINATE YOURNAMEFROMDISTRIBUTION LISTSFORDOCUMENTS YOUDON'TNEED!FULLTEXTCONVERSION REQUIREDTOTALNUMBEROFCOPIESREQUIRED:
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| LTTR27ENCL27 AsrROCHESTER GASANDELECTRICCORPORATION ROBERTC.MECREOYVicePreridrnrCinnaCrurlesrProducuon rrC~~~~vr4'rroerrScare89EASTAVENUE,ROCHESTER N.Y.14649-0001 TELEPHONE AicEAcoDE716546'2700May27,1994U.S.NuclearRegulatory Commission Attn:AllenR.JohnsonPWRProjectDirectorate I-3DocumentControlDeskWashington, DC20555
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| ==Subject:== | | ==Subject:== |
| LER94-007,Feedwater Transient, DuetoLossofAbilitytoControlFeedwater Regulating Valve,CausesaLoLoSteamGenerator LevelReactorTllpR.E.GinnaNuclearPowerPlantDocketNo.50-244Inaccordance with10CFR50.73,LicenseeEventReportSystem,item(a)(2)(iv),whichrequiresareportof,"anyeventorcondition thatresultedinamanualorautomatic actuation ofanyengineered safetyfeature(ESF),including thereactorprotection system(RPS)",theattachedLicenseeEventReportLER94-007isherebysubmitted.
| | LER 94-007, Feedwater Transient, Due to Loss of Ability to Control Feedwater Regulating Valve, Causes a Lo Lo Steam Generator Level Reactor Tl lp R.E.Ginna Nuclear Power Plant Docket No.50-244 In accordance with 10 CFR 50.73, Licensee Event Report System, item (a)(2)(iv), which requires a report of,"any event or condition that resulted in a manual or automatic actuation of any engineered safety feature (ESF), including the reactor protection system (RPS)", the attached Licensee Event Report LER 94-007 is hereby submitted. |
| Thiseventhasinnowayaffectedthepublic'shealthandsafety.Verytrulyyours,g.P~g'~/r X.C~RobertC.Mecredyxc:U.S.NuclearRegulatory Commission RegionI475Allendale RoadKingofPrussia,PA19406GinnaUSNRCSeniorResidentInspector 9406030044
| | This event has in no way affected the public's health and safety.Very truly yours, g.P~g'~/r X.C~Robert C.Mecredy xc: U.S.Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna USNRC Senior Resident Inspector 9406030044 |
| 'ss40527PDRADOCK05000244C~PDR NRCFORH366(5-92)U.S.NUCLEARREGULATORY C(NHISSIOHPROVEDBYMSNO.3150-0104 EXPIRES5/31/95LICENSEEEVENTREPORT(LER)(Seereverseforrequirednunberofdigits/characters foreachblock)ESTIHATED BURDENPERRESPONSETOCOMPLYWITHTHISINFORMATION COLLECTION REQUEST:50.0HRS.FORWARDCOMMENTSREGARDING BURDENESTIMA'IE TOTHEINFORMATION ANDRECORDSMANAGEMENT BRANCH(HNBB7714),U.S.NUCLEARREGULATORY COMMISSION, WASHINGTON, DC20555-000'I ANDTOTHEPAPERWORK REDUCTION PROJECT(31400104),OFFICEOFMANAGEHENT ANDBUDGETWASHINGTON DC20503.FAGILITYNAME(1)R.E~GinnaNuclearPowerPlantDOCKETHNNIER(2)05000244PAGE(3)10F9TITLE(4)Feedwater Transient, DuetoLossofAbilitytoControlFeedwater Regulating Valve,CausesaLoLoSteamGenerator LevelReactorTripHONTHDAYYEAREVENTDATE5YEARLERNINBER6SEQUENTIAL NUMBERREVISIONNUMBERMONTHDAYYEARREPORTDATE7OTHERFACILITIES INVOLVED8DOCKETNUMBERFACILITYNAME04279494--007--000527FACILITYHAHEDOCKETNUMBEROPERATING MODE(9)POWERLEVEL(10)N045THISREPORTISSUBMITTED PURSUANT20.402(b) 20.405(a)(1)(i)20.405(a)(1)(ii) 20.405(a)(1)(iii) 20.405(a)(1)(iv) 20.405(a)(1)(v) 20.405(c) 50.36(c)(1) 50.36(c)(2) 50.73(a)(2)(i) 50.73(a)(2)(ii) 50.73(a)(2)(iii) | | 'ss40527 PDR ADOCK 05000244 C~PDR NRC FORH 366 (5-92)U.S.NUCLEAR REGULATORY C(NHI SSIOH PROVED BY MS NO.3150-0104 EXPIRES 5/31/95 LICENSEE EVENT REPORT (LER)(See reverse for required nunber of digits/characters for each block)ESTIHATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMA'IE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (HNBB 7714), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-000'I AND TO THE PAPERWORK REDUCTION PROJECT (3140 0104), OFFICE OF MANAGEHENT AND BUDGET WASHINGTON DC 20503.FAGILITY NAME (1)R.E~Ginna Nuclear Power Plant DOCKET HNNIER (2)05000244 PAGE (3)10F9 TITLE (4)Feedwater Transient, Due to Loss of Ability to Control Feedwater Regulating Valve, Causes a Lo Lo Steam Generator Level Reactor Trip HONTH DAY YEAR EVENT DATE 5 YEAR LER NINBER 6 SEQUENTIAL NUMBER REVISION NUMBER MONTH DAY YEAR REPORT DATE 7 OTHER FACILITIES INVOLVED 8 DOCKET NUMBER FACILITY NAME 04 27 94 94--007--00 05 27 FACILITY HAHE DOCKET NUMBER OPERATING MODE (9)POWER LEVEL (10)N 045 THIS REPORT IS SUBMITTED PURSUANT 20.402(b)20.405(a)(1)(i)20.405(a)(1)(ii) 20.405(a)(1)(iii) 20.405(a)(1)(iv) 20.405(a)(1)(v) 20.405(c)50.36(c)(1) 50.36(c)(2) 50.73(a)(2)(i) 50.73(a)(2)(ii) 50.73(a)(2)(iii) |
| X50.73(a)(2)(iv) 50.73(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(b)73~71(c)OTHER(SpecifyinAbstractbelo~andinText,NRCForm366ATOTHEREQUIREMENTS OF10CFR:(Checkoneormore11LICENSEECONTACTFORTHISLER12NAMEJohnT.St.Martin-Director, Operating Experience TELEPHONE NUMBER(IncludeAreaCode)(315)524.4446C(NPLETEONELINEFOREACHCOHPOHENT FAILUREDESCRIBED INTHISREPORT13CAUSESYSTEHBJBLCVB042COMPONENT MANUFACTURER REPORTABLE TONPRDSPj's)kj~';.CAUSESYSTEMCOMPONENT MANUFACTURER REPORTABLE TONPRDSSUPPLEMENTAL REPORTEXPECTED14YES(Ifyes,coapleteEXPECTEDSUBMISSION DATE).XNOEXPECTEDSUBHISSIONDATE(15)MONTHDAYYEARABSTRACT(Limitto1400spaces,i.e.,approximately 15single-spaced typewritten lines)(16)OnApril27,1994,atapproximately 1407EDST,withthereactoratapproximately 454reactorpower,theabilitytocontrolthe"A"mainfeedwater regulating valvewaslost.At1410EDST,thereactortrippedonLoLolevel((/=17%)inthe"A"SteamGenerator.
| | X 50.73(a)(2)(iv) 50.73(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(b)73~71(c)OTHER (Specify in Abstract belo~and in Text, NRC Form 366A TO THE REQUIREMENTS OF 10 CFR: (Check one or more 11 LICENSEE CONTACT FOR THIS LER 12 NAME John T.St.Martin-Director, Operating Experience TELEPHONE NUMBER (Include Area Code)(315)524.4446 C(NPLETE ONE LINE FOR EACH COHPOHENT FAILURE DESCRIBED IN THIS REPORT 13 CAUSE SYSTEH B JB LCV B042 COMPONENT MANUFACTURER REPORTABLE TO NPRDS Pj's)kj~';.CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NPRDS SUPPLEMENTAL REPORT EXPECTED 14 YES (If yes, coaplete EXPECTED SUBMISSION DATE).X NO EXPECTED SUBHI SSI ON DATE (15)MONTH DAY YEAR ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)(16)On April 27, 1994, at approximately 1407 EDST, with the reactor at approximately 454 reactor power, the ability to control the"A" main feedwater regulating valve was lost.At 1410 EDST, the reactor tripped on Lo Lo level ((/=17%)in the"A" Steam Generator. |
| TheControlRoomoperators performed theactionsofprocedures E-0andES-0.1.Theunderlying causewasdetermined tobeanimproperly securedstrokeadjustsetscrewforthevalvepositioner signaldiaphragm assemblyforthe"A"mainfeedwater regulating valve.ThiseventisNUREG-1022 CauseCode(B),"Design,Manufacturing, Construction/Installation."
| | The Control Room operators performed the actions of procedures E-0 and ES-0.1.The underlying cause was determined to be an improperly secured stroke adjust set screw for the valve positioner signal diaphragm assembly for the"A" main feedwater regulating valve.This event is NUREG-1022 Cause Code (B),"Design, Manufacturing, Construction/Installation." Immediate corrective action was to install a new valve positioner of a previous design.Corrective action to preclude repetition is outlined in Section V (B).NRC FORM 366 (5-92) |
| Immediate corrective actionwastoinstallanewvalvepositioner ofapreviousdesign.Corrective actiontoprecluderepetition isoutlinedinSectionV(B).NRCFORM366(5-92) | | NRC FORM 366A (5.92)U.S.NUCLEAR REGULATORY COMMISSIOH LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PROVED BY QHI NO.3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY MITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.FORNARD COMMENTS REGARDIHG BURDEN ESTIMATE TO THE INFORMATION AHD RECORDS MANAGEHENT BRANCH (MHBB 7714), U.S.NUCLEAR REGULATORY COMMISSION, NASHINGTON, DC 20555-0001 AND TO THE PAPERHORK REDUCTION PROJECT (3140.0104), OFFICE OF MANAGEMENT AHD BUDGET llASHINGTON DC 20503.FACILITY NAME 1 R.E.Ginna Nuclear Power Plant DOCKET NMBER 2 05000244 YEAR 94 LER NUMBER 6 SEQUENTIAL |
| NRCFORM366A(5.92)U.S.NUCLEARREGULATORY COMMISSIOH LICENSEEEVENTREPORT(LER)TEXTCONTINUATION PROVEDBYQHINO.3150-0104 EXPIRES5/31/95ESTIMATED BURDENPERRESPONSETOCOMPLYMITHTHISINFORMATION COLLECTION REQUEST:50.0HRS.FORNARDCOMMENTSREGARDIHG BURDENESTIMATETOTHEINFORMATION AHDRECORDSMANAGEHENT BRANCH(MHBB7714),U.S.NUCLEARREGULATORY COMMISSION, NASHINGTON, DC20555-0001 ANDTOTHEPAPERHORK REDUCTION PROJECT(3140.0104),
| | --007--REVISION 00 PAGE 3 2 OF 9 TEXT (If more space is required, use additional copies of HRC Form 366A)(17)PRE-EVENT PLANT CONDITIONS The plant achieved full power operation on April 23, 1994, after completion of the 1994 annual refueling and maintenance outage.On April 26, 1994, the plant was manually shut down to repair a small steam leak on an instrument fitting on the high pressure (HP)turbine.Due to stability problems with control, of feedwater flow to the"A" Steam Generator (S/G)at steady state conditions, the valve positioner for the"A" main feedwater regulating valve (MFRV)was replaced during this brief shutdown.On April 27, 1994, a load increase was in progress, controlled by Plant Operating Procedure 0-1.2,"Plant-Startup from Hot Shutdown to Full Load".The plant was at approximately 45%reactor power.Preparations were being made to start a second main feedwater pump, per System Operating Procedure T-4.F,"Restoring 1B Feedwater Pump to Service After Maintenance or Power Reduction". |
| OFFICEOFMANAGEMENT AHDBUDGETllASHINGTON DC20503.FACILITYNAME1R.E.GinnaNuclearPowerPlantDOCKETNMBER205000244YEAR94LERNUMBER6SEQUENTIAL
| | II.DESCRIPTION OF EVENT A.DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES: |
| --007--REVISION00PAGE32OF9TEXT(Ifmorespaceisrequired, useadditional copiesofHRCForm366A)(17)PRE-EVENT PLANTCONDITIONS Theplantachievedfullpoweroperation onApril23,1994,aftercompletion ofthe1994annualrefueling andmaintenance outage.OnApril26,1994,theplantwasmanuallyshutdowntorepairasmallsteamleakonaninstrument fittingonthehighpressure(HP)turbine.Duetostability problemswithcontrol,offeedwater flowtothe"A"SteamGenerator (S/G)atsteadystateconditions, thevalvepositioner forthe"A"mainfeedwater regulating valve(MFRV)wasreplacedduringthisbriefshutdown. | | o April 27, 1994, 1410 EDST: Event date and time.o April 27, 1994, 1410 EDST: Discovery date and time.o April 27, 1994, 1410 EDST: Control Room operators verify both reactor trip breakers open, and all control and shutdown rods inserted.o April 27, 1994, 1412 EDST: Control Room operators manually stop the operating main feedwater pump to limit a reactor coolant system cooldown.o April 27, 1994, 1416 EDST: Control Room operators manually close both main steam isolation valves to limit a reactor coolant system cooldown.o April 27, 1994, 1440 EDST: Plant stabilized at hot shutdown condition. |
| OnApril27,1994,aloadincreasewasinprogress, controlled byPlantOperating Procedure 0-1.2,"Plant-StartupfromHotShutdowntoFullLoad".Theplantwasatapproximately 45%reactorpower.Preparations werebeingmadetostartasecondmainfeedwater pump,perSystemOperating Procedure T-4.F,"Restoring 1BFeedwater PumptoServiceAfterMaintenance orPowerReduction".
| | NRC FORM 366A (5-92) |
| II.DESCRIPTION OFEVENTA.DATESANDAPPROXIMATE TIMESOFMAJOROCCURRENCES: | | NRC FORM 366A (5-92)U.S.NUCLEAR REGULATORY CQOIISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION APPROVED BY Q(B HO.3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS IHFORMATION COLLECTIOH REQUEST: 50.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (31/0-0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.FACILITY NAME 1 R.E.Ginna Nuclear Power Plant DOCKET HQIBER 2 YEAR 05000244 94 LER NUMBER 6 SEOUENTIAL M--007--REVISION 00 PAGE 3 3 OF 9 TEXT (If more space is required, use additional copies of HRC Form 366A)(17)B.EVENT: On April 27, 1994, a load increase was in progress, following the brief plant shutdown on April 26, 1994.Plant Operating Procedure 0-1.2 was being followed to control the load increase.Only one main feedwater pump was in operation, and preparations were being made to start the second main feedwater pump and to continue with the load increase.On April 27, 1994, at approximately 1407 EDST, the Control Room operators noticed a slight decrease in feedwater flow to the"A" S/G.They attempted to manually increase feedwater flow, but the"A" MFRV did not respond to the demand signal to open the valve from the Main Control Board.At approximately 1408 EDST, Main Control Board annunciator G-22,"ADFCS System Trouble" alarmed, due to a deviation between the demand signal and actual position of the"A" MFRV.Level continued to decrease in the"A" S/G.The Shift Supervisor ordered a rapid load reduction, in an attempt to decrease the need for feedwater flow.Within two minutes, power had been decreased by approximately 15%, and actual feedwater flow being delivered to the"A" S/G exceeded steam flow.However, level in the"A" S/G decrea'sed to (17oI resulting in a reactor trip on S/G Lo Lo level, at 1410 EDST.The Control Room operators performed the immediate actions of Emergency Operating Procedure E-O,"Reactor Trip or Safety Injection", and transitioned to Emergency Operating Procedure ES-0.1',"Reactor Trip Response", when it was verified that both reactor trip breakers were open, all control and shutdown rods were inserted, and safety injection was not actuated or required.During performance of ES-0.1, the Control Room operators noted that an anticipated reactor coolant system (RCS)cooldown was occurring, and manually stopped the operating main feedwater pump.In addition, both main steam isolation valves (MSIVs)were manually closed by the Control Room operators. |
| oApril27,1994,1410EDST:Eventdateandtime.oApril27,1994,1410EDST:Discovery dateandtime.oApril27,1994,1410EDST:ControlRoomoperators verifybothreactortripbreakersopen,andallcontrolandshutdownrodsinserted.
| | These actions mitigated the RCS cooldown.During this event, pressurizer (PRZR)level decreased below the setpoint for letdown isolation, closing the letdown isolation valves and deenergizing the PRZR heaters.After PRZR level was restored above the setpoint, the Control Room Foreman directed that letdown and PRZR heaters be restored to service.The plant was subsequently stabilized in hot shutdown (at approximately 1440 EDST)using Plant Operating Procedures 0-3,"Hot Shutdown with Xenon Present", and 0-2,"Plant Shutdown". |
| oApril27,1994,1412EDST:ControlRoomoperators manuallystoptheoperating mainfeedwater pumptolimitareactorcoolantsystemcooldown.
| | HRC FORM 366A (5-92) |
| oApril27,1994,1416EDST:ControlRoomoperators manuallyclosebothmainsteamisolation valvestolimitareactorcoolantsystemcooldown.
| | HRC FORM 366A (5-92).S.NUCLEAR REGUULTORY COMMISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PROVED BY OIB NO.3150-0104 EXP I RES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTIOH REQUEST: 50.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATIOH AHD RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140-0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.FACILITY NAME 1 R.E.Ginna Nuclear Power Plant DOCKET NMBER 2 05000244 i LER NMBER 6 YEAR SEQUEH'TIAL 94--007 REVISION 00 PAGE 3 4 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)C~I NOPERABLE STRUCTURES I COMPONENTS I OR SYSTEMS THAT CONTRIBUTED TO THE EVENT: None D.OTHER'YSTEMS OR SECONDARY FUNCTIONS AFFECTED: None E.METHOD OF DISCOVERY: |
| oApril27,1994,1440EDST:Plantstabilized athotshutdowncondition.
| | This event was apparent due to Main Control Board indications of the loss of ability to control feedwater flow to the"A" S/G.The reactor trip was immediately apparent due to alarms and indications in the Control Room.F.OPERATOR ACTION: After the reactor trip, the Control Room operators performed the actions of Emergency Operating Procedures E-O,"Reactor Trip or Safety Injection" and ES-0.1,"Reactor Trip Response". |
| NRCFORM366A(5-92)
| | The operating main feedwater pump was manually stopped, and the MSIVs were manually closed to limit further RCS cooldown.The plant was stabilized at hot shutdown.Subsequently, the Control Room operators notified higher supervision and the Nuclear Regulatory Commission per 10CFR50.72, Non-Emergency, 4 Hour Notification. |
| NRCFORM366A(5-92)U.S.NUCLEARREGULATORY CQOIISSION LICENSEEEVENTREPORT(LER)TEXTCONTINUATION APPROVEDBYQ(BHO.3150-0104 EXPIRES5/31/95ESTIMATED BURDENPERRESPONSETOCOMPLYWITHTHISIHFORMATION COLLECTIOH REQUEST:50.0HRS.FORWARDCOMMENTSREGARDING BURDENESTIMATETOTHEINFORMATION ANDRECORDSMANAGEMENT BRANCH(MNBB7714),U.S.NUCLEARREGULATORY COMMISSION, WASHINGTON, DC20555-0001 ANDTOTHEPAPERWORK REDUCTION PROJECT(31/0-0104),
| | G.SAFETY SYSTEM RESPONSES: |
| OFFICEOFMANAGEMENT ANDBUDGETWASHINGTON DC20503.FACILITYNAME1R.E.GinnaNuclearPowerPlantDOCKETHQIBER2YEAR0500024494LERNUMBER6SEOUENTIAL M--007--REVISION00PAGE33OF9TEXT(Ifmorespaceisrequired, useadditional copiesofHRCForm366A)(17)B.EVENT:OnApril27,1994,aloadincreasewasinprogress, following thebriefplantshutdownonApril26,1994.PlantOperating Procedure 0-1.2wasbeingfollowedtocontroltheloadincrease.
| | None III.CAUSE OF EVENT A.IMMEDIATE CAUSE: The reactor trip was due to"A" S/G Lo Lo level ((/=17%), caused by decreased feedwater flow to the"A" S/G.B.INTERMEDIATE CAUSE: The decreased feedwater flow to the"A" S/G was due to loss of ability to control the"A" MFRV, caused by the valve positioner for the"A" MFRV not responding to the demand open signal.HRC FORM 366A (5-92) |
| Onlyonemainfeedwater pumpwasinoperation, andpreparations werebeingmadetostartthesecondmainfeedwater pumpandtocontinuewiththeloadincrease.
| | HRC FORM 366A (5-92)U.S.NUCLEAR REGULATORY COMMISSION LICENSEE.EVENT REPORT (LER)TEXT CONTINUATION PROVED BY OMB NO.3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.'ORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S.NUCLEAR REGULATORY COMMISSIOH, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150 0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.FACILITY NAME 1'R.E.Ginna Nuclear Power Plant DOCKET NWBER 2 05000244 LER NIÃBER 6)YEAR SEQUENTIAL 94--007--REVISION ,00 PACE 3 5 OF 9 TEXT (lf more space is required, use additional copies of HRC Form 366A)(17)C.ROOT CAUSE: The underlying cause of the valve positioner for the"A" MFRV not properly responding to a change in input, demand signal from the controller on the Main Control Board was an improperly secured stroke adjust set screw.The set screw was found in a backed out condition in the valve position signal diaphragm assembly for the"A" MFRV.It is postulated that this stroke adjust allen head set screw did not have adequate thread sealant to prevent it from backing out of the signal diaphragm assembly cover when subjected-to vibration. |
| OnApril27,1994,atapproximately 1407EDST,theControlRoomoperators noticedaslightdecreaseinfeedwater flowtothe"A"S/G.Theyattempted tomanuallyincreasefeedwater flow,butthe"A"MFRVdidnotrespondtothedemandsignaltoopenthevalvefromtheMainControlBoard.Atapproximately 1408EDST,MainControlBoardannunciator G-22,"ADFCSSystemTrouble"alarmed,duetoadeviation betweenthedemandsignalandactualpositionofthe"A"MFRV.Levelcontinued todecreaseinthe"A"S/G.TheShiftSupervisor orderedarapidloadreduction, inanattempttodecreasetheneedforfeedwater flow.Withintwominutes,powerhadbeendecreased byapproximately 15%,andactualfeedwater flowbeingdelivered tothe"A"S/Gexceededsteamflow.However,levelinthe"A"S/Gdecrea'sed to(17oIresulting inareactortriponS/GLoLolevel,at1410EDST.TheControlRoomoperators performed theimmediate actionsofEmergency Operating Procedure E-O,"ReactorTriporSafetyInjection",
| | With the set screw backed out, the signal diaphragm was restricted from responding to an input demand signal to open the MFRV.Note that the set screw is a factory-set adjustment. |
| andtransitioned toEmergency Operating Procedure ES-0.1',"ReactorTripResponse",
| | This particular set screw backed out when subjected to less than twelve hours of operation, and this failure mode has not occurred in other valve positioners that have'accumulated thousands of hours of operation. |
| whenitwasverifiedthatbothreactortripbreakerswereopen,allcontrolandshutdownrodswereinserted, andsafetyinjection wasnotactuatedorrequired.
| | This event is NUREG-1022 Cause Code (B),"Design, Manufacturing, Construction/Installation." IV.ANALYSIS OF EVENT This event is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (a)(2)(iv), which requires a report of,"any event or condition that resulted in manual or automatic actuation of any engineered safety feature (ESF)including the reactor protection system (RPS)".The"A" S/G Lo Lo level reactor trip was an automatic actuation of the RPS.An assessment was performed considering both the safety consequen-ces and implications of this event with the following results and conclusions: |
| Duringperformance ofES-0.1,theControlRoomoperators notedthatananticipated reactorcoolantsystem(RCS)cooldownwasoccurring, andmanuallystoppedtheoperating mainfeedwater pump.Inaddition, bothmainsteamisolation valves(MSIVs)weremanuallyclosedbytheControlRoomoperators.
| | o There were no safety consequences or implications attributed to the reactor trip because:*The two reactor trip breakers opened as required.*All control and shutdown rods inserted as designed.*The plant was stabilized at hot shutdown.HRC FORM 366A (5-92) |
| Theseactionsmitigated theRCScooldown.
| | NRC FORH 366A (5-92)U.S NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION APPROVED BY (HGI NO.3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLEC'IIOH REQUEST: 50.0 HRS~FORNARD COMMENTS REGARDING BURDEN ESTIMATE TO tHE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S.NUCLEAR REGULATORY COMMISSION, MASHIHGTON, DC 20555-0001, AND TO THE PAPERNORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET MASHINGTON DC 20503.FACILITY NAME 1 R.E.Ginna Nuclear Power Plant DOCKET HINSER 2 I 05000244 YEAR 94--007 00 LER NIMBER 6 SEQUENTIAL REVISION PAGE 3 6 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)o The Ginna Updated Final Safety Analysis Report (UFSAR)transient, as described in Chapter 15.2.6,"Loss of Normal Feedwater", describes a condition where the reactor trips on Lo Lo S/G level.This UFSAR transient was reviewed and compared to the plant response for this event.The UFSAR transient is a complete loss of Main Feedwater (MFW)at full power, with only one auxiliary feedwater (AFW)pump available one (1)minute after the loss of MFW, and secondary steam relief (i.e., decay heat removal)through the safety valves only.The protection against a loss of MFW includes the reactor trip on Lo Lo S/G level and the start of the AFW pumps.These protection features operated as designed.Based on the above evaluation, the plant transient of April 27, 1994, is bounded by the UFSAR Safety Analysis assumptions. |
| Duringthisevent,pressurizer (PRZR)leveldecreased belowthesetpointforletdownisolation, closingtheletdownisolation valvesanddeenergizing thePRZRheaters.AfterPRZRlevelwasrestoredabovethesetpoint, theControlRoomForemandirectedthatletdownandPRZRheatersberestoredtoservice.Theplantwassubsequently stabilized inhotshutdown(atapproximately 1440EDST)usingPlantOperating Procedures 0-3,"HotShutdownwithXenonPresent",
| | o Technical Specifications (TS)were reviewed with respect to the post trip review data.'he following are the results of that review:*Following the reactor trip, PRZR water level decreased'o approximately 12.5%due to a moderate RCS cooldown.This cooldown occurred during the post trip recovery period.This cooldown was bounded by the plant accident analysis, and did not exceed the TS limit of 100 degrees F per hour.Additional mitigation was provided by closing the MSIVs and stopping-the main feedwater pump.TS 3.1.1.5 states, in part, that when the RCS temperature is at or above 350 degrees F, at least 100 KW of PRZR heaters will be operable.TS 3.1.1.5 also states, in part, that if the PRZR,is inoperable due to heaters, restore the PRZR'to operable status within six (6)hours.PRZR water level was restored above the setpoint for letdown isolation within two (2)minutes, restoring the PRZR heaters to operable status, well before the six'(6)hour action statement. |
| and0-2,"PlantShutdown".
| | NRC FORM 366A (5-92) |
| HRCFORM366A(5-92)
| | NRC FORH 366A (5.92)U.S.NUCLEAR REGULATORY COMHISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION APPROVED BY QGI NO.3150-0104 EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REOUEST: 50.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEHENT BRANCH (MNBB 7714), U.ST NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140 0104), OFFICE OF HANAGEMENT AND BUDGET WASHINGTON DC 20503.FACILITY NAHE 1 R.E.Ginna Nuclear Power Plant DOCKET NINBER 2 05000244 LER NUMBER 6 YEAR SEOUENTIAL 94--007--REVISION 00 PAGE 3 7 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)*Both S/G levels decreased following the reactor trip."A" S/G level decreased to<04, and"B" S/G level decreased to<14'.This is an expected transient. |
| HRCFORM366A(5-92).S.NUCLEARREGUULTORY COMMISSION LICENSEEEVENTREPORT(LER)TEXTCONTINUATION PROVEDBYOIBNO.3150-0104 EXPIRES5/31/95ESTIMATED BURDENPERRESPONSETOCOMPLYWITHTHISINFORMATION COLLECTIOH REQUEST:50.0HRS.FORWARDCOMMENTSREGARDING BURDENESTIMATETOTHEINFORMATIOH AHDRECORDSMANAGEMENT BRANCH(MNBB7714),U.S.NUCLEARREGULATORY COMMISSION, WASHINGTON, DC20555-0001 ANDTOTHEPAPERWORK REDUCTION PROJECT(3140-0104),
| | TS 4.3.5.5 states that in order to demonstrate that a reactor coolant loop is operable, the S/G water level shall be>/=169.Thus, both coolant loops were.inoperable, even though both loops were still in operation and performing their intended function of decay heat removal.Both S/Gs were available as a heat sink, and sufficient AFW flow was maintained for adequate steam release from both S/Gs.TS 3.1.1.1(c) states, in part, that except for special tests, when the RCS temperature is at or above 350 degrees F with the reactor power less than or equal to 130 MWT (8.5%), at least one reactor coolant loop and its associated S/G and reactor coolant pump shall be in operation. |
| OFFICEOFMANAGEMENT ANDBUDGETWASHINGTON DC20503.FACILITYNAME1R.E.GinnaNuclearPowerPlantDOCKETNMBER205000244iLERNMBER6YEARSEQUEH'TIAL 94--007REVISION00PAGE34OF9TEXT(Ifmorespaceisrequired, useadditional copiesofNRCForm366A)(17)C~INOPERABLE STRUCTURES ICOMPONENTS IORSYSTEMSTHATCONTRIBUTED TOTHEEVENT:NoneD.OTHER'YSTEMS ORSECONDARY FUNCTIONS AFFECTED:
| | Both reactor coolant loops were in operation, but the S/Gs were inoperable due to level indication. |
| NoneE.METHODOFDISCOVERY:
| | Both loops were restored to operable status when S/G levels were restored to>/=164 ("B" S/G level in less than four (4)minutes, and"A" S/G level in approximately ten (10)minutes).Based on the above and a review of post trip data and past plant transients, it can be concluded that the plant operated as designed, that there were no unreviewed safety questions, and that the public's health and safety was assured at all times.NRC FORM 366A (5-92) |
| ThiseventwasapparentduetoMainControlBoardindications ofthelossofabilitytocontrolfeedwater flowtothe"A"S/G.Thereactortripwasimmediately apparentduetoalarmsandindications intheControlRoom.F.OPERATORACTION:Afterthereactortrip,theControlRoomoperators performed theactionsofEmergency Operating Procedures E-O,"ReactorTriporSafetyInjection" andES-0.1,"ReactorTripResponse".
| | NRC FORH 366A (5-92)U.S.NUCLEAR REGULATORY C(SOIISSIOH LICENSEE EVENT REPORT (LER)TEXT CONTINUATION APPROVED BY MB NO.3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.FORWARD COMMENTS REGARDIHG BURDEN'STIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MHBB 7714), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140-0104), OFFICE OF MANAGEMENT AND BUDGET'WASHINGTON DC 20503.FACILITY NAME 1 R.E.Ginna Nuclear Power Plant DOCKET NQIBER 2 05000244 YEAR 94--007--00 LER HINBER 6 SEQUENTIAL REVISION M PAGE 3 8 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)V.CORRECTIVE ACTION A.ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS: o The valve positioner for the"A" MFRV (Bailey Model AV112100I or"Type AV1")had been replaced on April 26, 1994, with another Type AV1 positioner. |
| Theoperating mainfeedwater pumpwasmanuallystopped,andtheMSIVsweremanuallyclosedtolimitfurtherRCScooldown.
| | This specific positioner failed in less than twelve hours of use and was replaced after this event (on April 28, 1994)with a Bailey Model 5321030A10 (or"5321030") |
| Theplantwasstabilized athotshutdown.
| | valve positioner. |
| Subsequently, theControlRoomoperators notifiedhighersupervision andtheNuclearRegulatory Commission per10CFR50.72, Non-Emergency, 4HourNotification. | | The Bailey Model 5321030 is the original model of valve positioner for the MFRV application and had operated successfully since plant startup in 1969.This positioner model was changed out in 1991 as part of EWR 4773, which installed the Advanced Digital Feedwater Control System (ADFCS).The Type AV1 positioners have had a history of reliability and stability concerns in this MFRV application. |
| G.SAFETYSYSTEMRESPONSES: | | o The valve positioner for the"B" MFRV was also replaced on April 28, 1994, for the reasons discussed above.o Valve.ramp and step change diagnostic testing was performed for both MFRVs to verify proper valve positioning and response.B.ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE: |
| NoneIII.CAUSEOFEVENTA.IMMEDIATE CAUSE:Thereactortripwasdueto"A"S/GLoLolevel((/=17%),causedbydecreased feedwater flowtothe"A"S/G.B.INTERMEDIATE CAUSE:Thedecreased feedwater flowtothe"A"S/Gwasduetolossofabilitytocontrolthe"A"MFRV,causedbythevalvepositioner forthe"A"MFRVnotresponding tothedemandopensignal.HRCFORM366A(5-92)
| | o The positioners for both MFRVs were replaced with Model 5321030 positioners as discussed above.HRC FORM 366A (5-92) |
| HRCFORM366A(5-92)U.S.NUCLEARREGULATORY COMMISSION LICENSEE.
| | NRC FORM 366A (5-92)I.S.NUCLEAR REGULATORY CQOIISSIOH LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PROVED BY OMB NO.3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTIOH REDUEST: 50.0 HRS.FORWARD COMMENTS REGARDIHG BURDEH ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MHBB 7714), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001 AHD TO THE PAPERWORK REDUCTION PROJECT (3140.0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.FACILITY NAME 1 R.E.Ginna Nuclear Power Plant DOCKET NUMBER 2 05000244 YEAR 94 LER NUMBER 6 SEOUENTIAL M--007--REVISION 00 PAGE 3 9 OF 9 TEXT (If more space is required, use additional copies of HRC Form 366A)(17)VI.ADDITIONAL INFORMATION A.FAILED COMPONENTS: |
| EVENTREPORT(LER)TEXTCONTINUATION PROVEDBYOMBNO.3150-0104 EXPIRES5/31/95ESTIMATED BURDENPERRESPONSETOCOMPLYWITHTHISINFORMATION COLLECTION REQUEST:50.0HRS.'ORWARD COMMENTSREGARDING BURDENESTIMATETOTHEINFORMATION ANDRECORDSMANAGEMENT BRANCH(MNBB7714),U.S.NUCLEARREGULATORY COMMISSIOH, WASHINGTON, DC20555-0001, ANDTOTHEPAPERWORK REDUCTION PROJECT(31500104),OFFICEOFMANAGEMENT ANDBUDGETWASHINGTON DC20503.FACILITYNAME1'R.E.GinnaNuclearPowerPlantDOCKETNWBER205000244LERNIÃBER6)YEARSEQUENTIAL 94--007--REVISION,00PACE35OF9TEXT(lfmorespaceisrequired, useadditional copiesofHRCForm366A)(17)C.ROOTCAUSE:Theunderlying causeofthevalvepositioner forthe"A"MFRVnotproperlyresponding toachangeininput,demandsignalfromthecontroller ontheMainControlBoardwasanimproperly securedstrokeadjustsetscrew.Thesetscrewwasfoundinabackedoutcondition inthevalvepositionsignaldiaphragm assemblyforthe"A"MFRV.Itispostulated thatthisstrokeadjustallenheadsetscrewdidnothaveadequatethreadsealanttopreventitfrombackingoutofthesignaldiaphragm assemblycoverwhensubjected-tovibration.
| | The failed component was the stroke adjust allen head set screw for the valve positioner signal diaphragm assembly for the"A" MFRV.The positioner is a Model AV112100 positioner, manufactured by Bailey Controls Inc.B.PREVIOUS LERs ON SIMILAR EVENTS: A similar LER event historical search was conducted with the following results: no documentation of similar LER events with the same root cause at Ginna Nuclear Power Plant could be iden-tified;however, LER 93-006 was a similar event, in that there was loss of ability to control the"A" MFRV;and LERs 88-005, 90 007 I and 90-0 1 0 were simi lar events with dif f erent root causes.C.SPECIAL COMMENTS: None HRC FORM 366A (5-92)}} |
| Withthesetscrewbackedout,thesignaldiaphragm wasrestricted fromresponding toaninputdemandsignaltoopentheMFRV.Notethatthesetscrewisafactory-set adjustment.
| |
| Thisparticular setscrewbackedoutwhensubjected tolessthantwelvehoursofoperation, andthisfailuremodehasnotoccurredinothervalvepositioners thathave'accumulated thousands ofhoursofoperation.
| |
| ThiseventisNUREG-1022 CauseCode(B),"Design,Manufacturing, Construction/Installation."
| |
| IV.ANALYSISOFEVENTThiseventisreportable inaccordance with10CFR50.73,LicenseeEventReportSystem,item(a)(2)(iv),whichrequiresareportof,"anyeventorcondition thatresultedinmanualorautomatic actuation ofanyengineered safetyfeature(ESF)including thereactorprotection system(RPS)".The"A"S/GLoLolevelreactortripwasanautomatic actuation oftheRPS.Anassessment wasperformed considering boththesafetyconsequen-cesandimplications ofthiseventwiththefollowing resultsandconclusions: | |
| oTherewerenosafetyconsequences orimplications attributed tothereactortripbecause:*Thetworeactortripbreakersopenedasrequired.
| |
| *Allcontrolandshutdownrodsinsertedasdesigned. | |
| *Theplantwasstabilized athotshutdown. | |
| HRCFORM366A(5-92)
| |
| NRCFORH366A(5-92)U.SNUCLEARREGULATORY COMMISSION LICENSEEEVENTREPORT(LER)TEXTCONTINUATION APPROVEDBY(HGINO.3150-0104 EXPIRES5/31/95ESTIMATED BURDENPERRESPONSETOCOMPLYWITHTHISINFORMATION COLLEC'IIOH REQUEST:50.0HRS~FORNARDCOMMENTSREGARDING BURDENESTIMATETOtHEINFORMATION ANDRECORDSMANAGEMENT BRANCH(MNBB7714),U.S.NUCLEARREGULATORY COMMISSION, MASHIHGTON, DC20555-0001, ANDTOTHEPAPERNORK REDUCTION PROJECT(3150-0104),
| |
| OFFICEOFMANAGEMENT ANDBUDGETMASHINGTON DC20503.FACILITYNAME1R.E.GinnaNuclearPowerPlantDOCKETHINSER2I05000244YEAR94--00700LERNIMBER6SEQUENTIAL REVISIONPAGE36OF9TEXT(Ifmorespaceisrequired, useadditional copiesofNRCForm366A)(17)oTheGinnaUpdatedFinalSafetyAnalysisReport(UFSAR)transient, asdescribed inChapter15.2.6,"LossofNormalFeedwater",
| |
| describes acondition wherethereactortripsonLoLoS/Glevel.ThisUFSARtransient wasreviewedandcomparedtotheplantresponseforthisevent.TheUFSARtransient isacompletelossofMainFeedwater (MFW)atfullpower,withonlyoneauxiliary feedwater (AFW)pumpavailable one(1)minuteafterthelossofMFW,andsecondary steamrelief(i.e.,decayheatremoval)throughthesafetyvalvesonly.Theprotection againstalossofMFWincludesthereactortriponLoLoS/GlevelandthestartoftheAFWpumps.Theseprotection featuresoperatedasdesigned. | |
| Basedontheaboveevaluation, theplanttransient ofApril27,1994,isboundedbytheUFSARSafetyAnalysisassumptions.
| |
| oTechnical Specifications (TS)werereviewedwithrespecttotheposttripreviewdata.'hefollowing aretheresultsofthatreview:*Following thereactortrip,PRZRwaterleveldecreased
| |
| 'oapproximately 12.5%duetoamoderateRCScooldown. | |
| Thiscooldownoccurredduringtheposttriprecoveryperiod.Thiscooldownwasboundedbytheplantaccidentanalysis, anddidnotexceedtheTSlimitof100degreesFperhour.Additional mitigation wasprovidedbyclosingtheMSIVsandstopping-themainfeedwater pump.TS3.1.1.5states,inpart,thatwhentheRCStemperature isatorabove350degreesF,atleast100KWofPRZRheaterswillbeoperable.
| |
| TS3.1.1.5alsostates,inpart,thatifthePRZR,isinoperable duetoheaters,restorethePRZR'tooperablestatuswithinsix(6)hours.PRZRwaterlevelwasrestoredabovethesetpointforletdownisolation withintwo(2)minutes,restoring thePRZRheaterstooperablestatus,wellbeforethesix'(6)houractionstatement.
| |
| NRCFORM366A(5-92)
| |
| NRCFORH366A(5.92)U.S.NUCLEARREGULATORY COMHISSION LICENSEEEVENTREPORT(LER)TEXTCONTINUATION APPROVEDBYQGINO.3150-0104 EXPIRES5/31/95ESTIHATED BURDENPERRESPONSETOCOMPLYWITHTHISINFORMATION COLLECTION REOUEST:50.0HRS.FORWARDCOMMENTSREGARDING BURDENESTIMATETOTHEINFORMATION ANDRECORDSMANAGEHENT BRANCH(MNBB7714),U.STNUCLEARREGULATORY COMMISSION, WASHINGTON, DC20555-0001 ANDTOTHEPAPERWORK REDUCTION PROJECT(31400104),OFFICEOFHANAGEMENT ANDBUDGETWASHINGTON DC20503.FACILITYNAHE1R.E.GinnaNuclearPowerPlantDOCKETNINBER205000244LERNUMBER6YEARSEOUENTIAL 94--007--REVISION00PAGE37OF9TEXT(Ifmorespaceisrequired, useadditional copiesofNRCForm366A)(17)*BothS/Glevelsdecreased following thereactortrip."A"S/Gleveldecreased to<04,and"B"S/Gleveldecreased to<14'.Thisisanexpectedtransient.
| |
| TS4.3.5.5statesthatinordertodemonstrate thatareactorcoolantloopisoperable, theS/Gwaterlevelshallbe>/=169.Thus,bothcoolantloopswere.inoperable, eventhoughbothloopswerestillinoperation andperforming theirintendedfunctionofdecayheatremoval.BothS/Gswereavailable asaheatsink,andsufficient AFWflowwasmaintained foradequatesteamreleasefrombothS/Gs.TS3.1.1.1(c) states,inpart,thatexceptforspecialtests,whentheRCStemperature isatorabove350degreesFwiththereactorpowerlessthanorequalto130MWT(8.5%),atleastonereactorcoolantloopanditsassociated S/Gandreactorcoolantpumpshallbeinoperation.
| |
| Bothreactorcoolantloopswereinoperation, buttheS/Gswereinoperable duetolevelindication.
| |
| BothloopswererestoredtooperablestatuswhenS/Glevelswererestoredto>/=164("B"S/Glevelinlessthanfour(4)minutes,and"A"S/Glevelinapproximately ten(10)minutes).
| |
| Basedontheaboveandareviewofposttripdataandpastplanttransients, itcanbeconcluded thattheplantoperatedasdesigned, thattherewerenounreviewed safetyquestions, andthatthepublic'shealthandsafetywasassuredatalltimes.NRCFORM366A(5-92)
| |
| NRCFORH366A(5-92)U.S.NUCLEARREGULATORY C(SOIISSIOH LICENSEEEVENTREPORT(LER)TEXTCONTINUATION APPROVEDBYMBNO.3150-0104 EXPIRES5/31/95ESTIMATED BURDENPERRESPONSETOCOMPLYWITHTHISINFORMATION COLLECTION REQUEST:50.0HRS.FORWARDCOMMENTSREGARDIHG BURDEN'STIMATE TOTHEINFORMATION ANDRECORDSMANAGEMENT BRANCH(MHBB7714),U.S.NUCLEARREGULATORY COMMISSION, WASHINGTON, DC20555-0001 ANDTOTHEPAPERWORK REDUCTION PROJECT(3140-0104),
| |
| OFFICEOFMANAGEMENT ANDBUDGET'WASHINGTON DC20503.FACILITYNAME1R.E.GinnaNuclearPowerPlantDOCKETNQIBER205000244YEAR94--007--00LERHINBER6SEQUENTIAL REVISIONMPAGE38OF9TEXT(Ifmorespaceisrequired, useadditional copiesofNRCForm366A)(17)V.CORRECTIVE ACTIONA.ACTIONTAKENTORETURNAFFECTEDSYSTEMSTOPRE-EVENT NORMALSTATUS:oThevalvepositioner forthe"A"MFRV(BaileyModelAV112100I or"TypeAV1")hadbeenreplacedonApril26,1994,withanotherTypeAV1positioner.
| |
| Thisspecificpositioner failedinlessthantwelvehoursofuseandwasreplacedafterthisevent(onApril28,1994)withaBaileyModel5321030A10 (or"5321030")
| |
| valvepositioner.
| |
| TheBaileyModel5321030istheoriginalmodelofvalvepositioner fortheMFRVapplication andhadoperatedsuccessfully sinceplantstartupin1969.Thispositioner modelwaschangedoutin1991aspartofEWR4773,whichinstalled theAdvancedDigitalFeedwater ControlSystem(ADFCS).TheTypeAV1positioners havehadahistoryofreliability andstability concernsinthisMFRVapplication.
| |
| oThevalvepositioner forthe"B"MFRVwasalsoreplacedonApril28,1994,forthereasonsdiscussed above.oValve.rampandstepchangediagnostic testingwasperformed forbothMFRVstoverifypropervalvepositioning andresponse.
| |
| B.ACTIONTAKENORPLANNEDTOPREVENTRECURRENCE: | |
| oThepositioners forbothMFRVswerereplacedwithModel5321030positioners asdiscussed above.HRCFORM366A(5-92)
| |
| NRCFORM366A(5-92)I.S.NUCLEARREGULATORY CQOIISSIOH LICENSEEEVENTREPORT(LER)TEXTCONTINUATION PROVEDBYOMBNO.3150-0104 EXPIRES5/31/95ESTIMATED BURDENPERRESPONSETOCOMPLYWITHTHISINFORMATION COLLECTIOH REDUEST:50.0HRS.FORWARDCOMMENTSREGARDIHG BURDEHESTIMATETOTHEINFORMATION ANDRECORDSMANAGEMENT BRANCH(MHBB7714),U.S.NUCLEARREGULATORY COMMISSION, WASHINGTON, DC20555-0001 AHDTOTHEPAPERWORK REDUCTION PROJECT(3140.0104),
| |
| OFFICEOFMANAGEMENT ANDBUDGETWASHINGTON DC20503.FACILITYNAME1R.E.GinnaNuclearPowerPlantDOCKETNUMBER205000244YEAR94LERNUMBER6SEOUENTIAL M--007--REVISION00PAGE39OF9TEXT(Ifmorespaceisrequired, useadditional copiesofHRCForm366A)(17)VI.ADDITIONAL INFORMATION A.FAILEDCOMPONENTS:
| |
| Thefailedcomponent wasthestrokeadjustallenheadsetscrewforthevalvepositioner signaldiaphragm assemblyforthe"A"MFRV.Thepositioner isaModelAV112100positioner, manufactured byBaileyControlsInc.B.PREVIOUSLERsONSIMILAREVENTS:AsimilarLEReventhistorical searchwasconducted withthefollowing results:nodocumentation ofsimilarLEReventswiththesamerootcauseatGinnaNuclearPowerPlantcouldbeiden-tified;however,LER93-006wasasimilarevent,inthattherewaslossofabilitytocontrolthe"A"MFRV;andLERs88-005,90007Iand90-010weresimilareventswithdifferentrootcauses.C.SPECIALCOMMENTS:
| |
| NoneHRCFORM366A(5-92)}}
| |
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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
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ACCELERATED DISTRIBUTION DEMONSTRATION SYSTEM REGULATORY INFORMATION DISTRIBUTION-SYSTEM (RIDS)ACCESSION NBR:9406030044 DOC.DATE: 94/05/27 NOTARIZED:
NO FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G AUTH.NAME AUTHOR AFFILIATION ST.MARTIN,J.T.
Rochester Gas&Electric Corp.MECREDY,R.C.
Rochester Gas&Electric Corp.RECIP.NAME RECIPIENT AFFILIATION I
SUBJECT:
LER 94-007-00:on 940427,feedwater transient occurred due to loss of ability to control feedwater regulatinq valve, causing lo lo SG level reactor trip.Caused by improperly secured stroke adjust set screw.W/940527 ltr.DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR j ENCL/SIZE: TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.DOCKET g 05000244.R D 8 NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).
05000244 A RECIPIENT ID CODE/NAME PD1-3 PD INTERNAL: AEOD/DOA AEOD/ROAB/DSP NRR/DE/EMEB NRR/DRCH/HHFB NRR/DRCH/HOLB NRR/DRSS/PRPB NRR/DSSA/SRXB RES/DSIR/EIB I EXTERNAL: EG&G BRYCE,J.H NRC PDR NSIC POOREiW~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 JOHNSON,A AEOD/DSP/TPAB NRR/DE/EELB NRR/DORS/OEAB NRR/DRCH/HICB NRR/DRIL/RPEB NRR/~SSA/S PLB REGFFL~02 G 1 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 D D D NOTE TO ALL"RIDS" RECIPIENTS:
A D D PLEASE HELP US TO REDUCE WASTE!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 27 ENCL 27 A sr ROCHESTER GAS AND ELECTRIC CORPORATION ROBERT C.MECREOY Vice Preridr nr Cinna Crurlesr Producuon rr C~~~~v r 4'r roe rr Scare 89 EAST AVENUE, ROCHESTER N.Y.14649-0001 TELEPHONE AicEA coDE 716 546'2700 May 27, 1994 U.S.Nuclear Regulatory Commission Attn: Allen R.Johnson PWR Project Directorate I-3 Document Control Desk Washington, DC 20555
Subject:
LER 94-007, Feedwater Transient, Due to Loss of Ability to Control Feedwater Regulating Valve, Causes a Lo Lo Steam Generator Level Reactor Tl lp R.E.Ginna Nuclear Power Plant Docket No.50-244 In accordance with 10 CFR 50.73, Licensee Event Report System, item (a)(2)(iv), which requires a report of,"any event or condition that resulted in a manual or automatic actuation of any engineered safety feature (ESF), including the reactor protection system (RPS)", the attached Licensee Event Report LER 94-007 is hereby submitted.
This event has in no way affected the public's health and safety.Very truly yours, g.P~g'~/r X.C~Robert C.Mecredy xc: U.S.Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna USNRC Senior Resident Inspector 9406030044
'ss40527 PDR ADOCK 05000244 C~PDR NRC FORH 366 (5-92)U.S.NUCLEAR REGULATORY C(NHI SSIOH PROVED BY MS NO.3150-0104 EXPIRES 5/31/95 LICENSEE EVENT REPORT (LER)(See reverse for required nunber of digits/characters for each block)ESTIHATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMA'IE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (HNBB 7714), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-000'I AND TO THE PAPERWORK REDUCTION PROJECT (3140 0104), OFFICE OF MANAGEHENT AND BUDGET WASHINGTON DC 20503.FAGILITY NAME (1)R.E~Ginna Nuclear Power Plant DOCKET HNNIER (2)05000244 PAGE (3)10F9 TITLE (4)Feedwater Transient, Due to Loss of Ability to Control Feedwater Regulating Valve, Causes a Lo Lo Steam Generator Level Reactor Trip HONTH DAY YEAR EVENT DATE 5 YEAR LER NINBER 6 SEQUENTIAL NUMBER REVISION NUMBER MONTH DAY YEAR REPORT DATE 7 OTHER FACILITIES INVOLVED 8 DOCKET NUMBER FACILITY NAME 04 27 94 94--007--00 05 27 FACILITY HAHE DOCKET NUMBER OPERATING MODE (9)POWER LEVEL (10)N 045 THIS REPORT IS SUBMITTED PURSUANT 20.402(b)20.405(a)(1)(i)20.405(a)(1)(ii) 20.405(a)(1)(iii) 20.405(a)(1)(iv) 20.405(a)(1)(v) 20.405(c)50.36(c)(1) 50.36(c)(2) 50.73(a)(2)(i) 50.73(a)(2)(ii) 50.73(a)(2)(iii)
X 50.73(a)(2)(iv) 50.73(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(b)73~71(c)OTHER (Specify in Abstract belo~and in Text, NRC Form 366A TO THE REQUIREMENTS OF 10 CFR: (Check one or more 11 LICENSEE CONTACT FOR THIS LER 12 NAME John T.St.Martin-Director, Operating Experience TELEPHONE NUMBER (Include Area Code)(315)524.4446 C(NPLETE ONE LINE FOR EACH COHPOHENT FAILURE DESCRIBED IN THIS REPORT 13 CAUSE SYSTEH B JB LCV B042 COMPONENT MANUFACTURER REPORTABLE TO NPRDS Pj's)kj~';.CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NPRDS SUPPLEMENTAL REPORT EXPECTED 14 YES (If yes, coaplete EXPECTED SUBMISSION DATE).X NO EXPECTED SUBHI SSI ON DATE (15)MONTH DAY YEAR ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)(16)On April 27, 1994, at approximately 1407 EDST, with the reactor at approximately 454 reactor power, the ability to control the"A" main feedwater regulating valve was lost.At 1410 EDST, the reactor tripped on Lo Lo level ((/=17%)in the"A" Steam Generator.
The Control Room operators performed the actions of procedures E-0 and ES-0.1.The underlying cause was determined to be an improperly secured stroke adjust set screw for the valve positioner signal diaphragm assembly for the"A" main feedwater regulating valve.This event is NUREG-1022 Cause Code (B),"Design, Manufacturing, Construction/Installation." Immediate corrective action was to install a new valve positioner of a previous design.Corrective action to preclude repetition is outlined in Section V (B).NRC FORM 366 (5-92)
NRC FORM 366A (5.92)U.S.NUCLEAR REGULATORY COMMISSIOH LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PROVED BY QHI NO.3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY MITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.FORNARD COMMENTS REGARDIHG BURDEN ESTIMATE TO THE INFORMATION AHD RECORDS MANAGEHENT BRANCH (MHBB 7714), U.S.NUCLEAR REGULATORY COMMISSION, NASHINGTON, DC 20555-0001 AND TO THE PAPERHORK REDUCTION PROJECT (3140.0104), OFFICE OF MANAGEMENT AHD BUDGET llASHINGTON DC 20503.FACILITY NAME 1 R.E.Ginna Nuclear Power Plant DOCKET NMBER 2 05000244 YEAR 94 LER NUMBER 6 SEQUENTIAL
--007--REVISION 00 PAGE 3 2 OF 9 TEXT (If more space is required, use additional copies of HRC Form 366A)(17)PRE-EVENT PLANT CONDITIONS The plant achieved full power operation on April 23, 1994, after completion of the 1994 annual refueling and maintenance outage.On April 26, 1994, the plant was manually shut down to repair a small steam leak on an instrument fitting on the high pressure (HP)turbine.Due to stability problems with control, of feedwater flow to the"A" Steam Generator (S/G)at steady state conditions, the valve positioner for the"A" main feedwater regulating valve (MFRV)was replaced during this brief shutdown.On April 27, 1994, a load increase was in progress, controlled by Plant Operating Procedure 0-1.2,"Plant-Startup from Hot Shutdown to Full Load".The plant was at approximately 45%reactor power.Preparations were being made to start a second main feedwater pump, per System Operating Procedure T-4.F,"Restoring 1B Feedwater Pump to Service After Maintenance or Power Reduction".
II.DESCRIPTION OF EVENT A.DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES:
o April 27, 1994, 1410 EDST: Event date and time.o April 27, 1994, 1410 EDST: Discovery date and time.o April 27, 1994, 1410 EDST: Control Room operators verify both reactor trip breakers open, and all control and shutdown rods inserted.o April 27, 1994, 1412 EDST: Control Room operators manually stop the operating main feedwater pump to limit a reactor coolant system cooldown.o April 27, 1994, 1416 EDST: Control Room operators manually close both main steam isolation valves to limit a reactor coolant system cooldown.o April 27, 1994, 1440 EDST: Plant stabilized at hot shutdown condition.
NRC FORM 366A (5-92)
NRC FORM 366A (5-92)U.S.NUCLEAR REGULATORY CQOIISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION APPROVED BY Q(B HO.3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS IHFORMATION COLLECTIOH REQUEST: 50.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (31/0-0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.FACILITY NAME 1 R.E.Ginna Nuclear Power Plant DOCKET HQIBER 2 YEAR 05000244 94 LER NUMBER 6 SEOUENTIAL M--007--REVISION 00 PAGE 3 3 OF 9 TEXT (If more space is required, use additional copies of HRC Form 366A)(17)B.EVENT: On April 27, 1994, a load increase was in progress, following the brief plant shutdown on April 26, 1994.Plant Operating Procedure 0-1.2 was being followed to control the load increase.Only one main feedwater pump was in operation, and preparations were being made to start the second main feedwater pump and to continue with the load increase.On April 27, 1994, at approximately 1407 EDST, the Control Room operators noticed a slight decrease in feedwater flow to the"A" S/G.They attempted to manually increase feedwater flow, but the"A" MFRV did not respond to the demand signal to open the valve from the Main Control Board.At approximately 1408 EDST, Main Control Board annunciator G-22,"ADFCS System Trouble" alarmed, due to a deviation between the demand signal and actual position of the"A" MFRV.Level continued to decrease in the"A" S/G.The Shift Supervisor ordered a rapid load reduction, in an attempt to decrease the need for feedwater flow.Within two minutes, power had been decreased by approximately 15%, and actual feedwater flow being delivered to the"A" S/G exceeded steam flow.However, level in the"A" S/G decrea'sed to (17oI resulting in a reactor trip on S/G Lo Lo level, at 1410 EDST.The Control Room operators performed the immediate actions of Emergency Operating Procedure E-O,"Reactor Trip or Safety Injection", and transitioned to Emergency Operating Procedure ES-0.1',"Reactor Trip Response", when it was verified that both reactor trip breakers were open, all control and shutdown rods were inserted, and safety injection was not actuated or required.During performance of ES-0.1, the Control Room operators noted that an anticipated reactor coolant system (RCS)cooldown was occurring, and manually stopped the operating main feedwater pump.In addition, both main steam isolation valves (MSIVs)were manually closed by the Control Room operators.
These actions mitigated the RCS cooldown.During this event, pressurizer (PRZR)level decreased below the setpoint for letdown isolation, closing the letdown isolation valves and deenergizing the PRZR heaters.After PRZR level was restored above the setpoint, the Control Room Foreman directed that letdown and PRZR heaters be restored to service.The plant was subsequently stabilized in hot shutdown (at approximately 1440 EDST)using Plant Operating Procedures 0-3,"Hot Shutdown with Xenon Present", and 0-2,"Plant Shutdown".
HRC FORM 366A (5-92)
HRC FORM 366A (5-92).S.NUCLEAR REGUULTORY COMMISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PROVED BY OIB NO.3150-0104 EXP I RES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTIOH REQUEST: 50.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATIOH AHD RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140-0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.FACILITY NAME 1 R.E.Ginna Nuclear Power Plant DOCKET NMBER 2 05000244 i LER NMBER 6 YEAR SEQUEH'TIAL 94--007 REVISION 00 PAGE 3 4 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)C~I NOPERABLE STRUCTURES I COMPONENTS I OR SYSTEMS THAT CONTRIBUTED TO THE EVENT: None D.OTHER'YSTEMS OR SECONDARY FUNCTIONS AFFECTED: None E.METHOD OF DISCOVERY:
This event was apparent due to Main Control Board indications of the loss of ability to control feedwater flow to the"A" S/G.The reactor trip was immediately apparent due to alarms and indications in the Control Room.F.OPERATOR ACTION: After the reactor trip, the Control Room operators performed the actions of Emergency Operating Procedures E-O,"Reactor Trip or Safety Injection" and ES-0.1,"Reactor Trip Response".
The operating main feedwater pump was manually stopped, and the MSIVs were manually closed to limit further RCS cooldown.The plant was stabilized at hot shutdown.Subsequently, the Control Room operators notified higher supervision and the Nuclear Regulatory Commission per 10CFR50.72, Non-Emergency, 4 Hour Notification.
G.SAFETY SYSTEM RESPONSES:
None III.CAUSE OF EVENT A.IMMEDIATE CAUSE: The reactor trip was due to"A" S/G Lo Lo level ((/=17%), caused by decreased feedwater flow to the"A" S/G.B.INTERMEDIATE CAUSE: The decreased feedwater flow to the"A" S/G was due to loss of ability to control the"A" MFRV, caused by the valve positioner for the"A" MFRV not responding to the demand open signal.HRC FORM 366A (5-92)
HRC FORM 366A (5-92)U.S.NUCLEAR REGULATORY COMMISSION LICENSEE.EVENT REPORT (LER)TEXT CONTINUATION PROVED BY OMB NO.3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.'ORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S.NUCLEAR REGULATORY COMMISSIOH, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150 0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.FACILITY NAME 1'R.E.Ginna Nuclear Power Plant DOCKET NWBER 2 05000244 LER NIÃBER 6)YEAR SEQUENTIAL 94--007--REVISION ,00 PACE 3 5 OF 9 TEXT (lf more space is required, use additional copies of HRC Form 366A)(17)C.ROOT CAUSE: The underlying cause of the valve positioner for the"A" MFRV not properly responding to a change in input, demand signal from the controller on the Main Control Board was an improperly secured stroke adjust set screw.The set screw was found in a backed out condition in the valve position signal diaphragm assembly for the"A" MFRV.It is postulated that this stroke adjust allen head set screw did not have adequate thread sealant to prevent it from backing out of the signal diaphragm assembly cover when subjected-to vibration.
With the set screw backed out, the signal diaphragm was restricted from responding to an input demand signal to open the MFRV.Note that the set screw is a factory-set adjustment.
This particular set screw backed out when subjected to less than twelve hours of operation, and this failure mode has not occurred in other valve positioners that have'accumulated thousands of hours of operation.
This event is NUREG-1022 Cause Code (B),"Design, Manufacturing, Construction/Installation." IV.ANALYSIS OF EVENT This event is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (a)(2)(iv), which requires a report of,"any event or condition that resulted in manual or automatic actuation of any engineered safety feature (ESF)including the reactor protection system (RPS)".The"A" S/G Lo Lo level reactor trip was an automatic actuation of the RPS.An assessment was performed considering both the safety consequen-ces and implications of this event with the following results and conclusions:
o There were no safety consequences or implications attributed to the reactor trip because:*The two reactor trip breakers opened as required.*All control and shutdown rods inserted as designed.*The plant was stabilized at hot shutdown.HRC FORM 366A (5-92)
NRC FORH 366A (5-92)U.S NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION APPROVED BY (HGI NO.3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLEC'IIOH REQUEST: 50.0 HRS~FORNARD COMMENTS REGARDING BURDEN ESTIMATE TO tHE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S.NUCLEAR REGULATORY COMMISSION, MASHIHGTON, DC 20555-0001, AND TO THE PAPERNORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET MASHINGTON DC 20503.FACILITY NAME 1 R.E.Ginna Nuclear Power Plant DOCKET HINSER 2 I 05000244 YEAR 94--007 00 LER NIMBER 6 SEQUENTIAL REVISION PAGE 3 6 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)o The Ginna Updated Final Safety Analysis Report (UFSAR)transient, as described in Chapter 15.2.6,"Loss of Normal Feedwater", describes a condition where the reactor trips on Lo Lo S/G level.This UFSAR transient was reviewed and compared to the plant response for this event.The UFSAR transient is a complete loss of Main Feedwater (MFW)at full power, with only one auxiliary feedwater (AFW)pump available one (1)minute after the loss of MFW, and secondary steam relief (i.e., decay heat removal)through the safety valves only.The protection against a loss of MFW includes the reactor trip on Lo Lo S/G level and the start of the AFW pumps.These protection features operated as designed.Based on the above evaluation, the plant transient of April 27, 1994, is bounded by the UFSAR Safety Analysis assumptions.
o Technical Specifications (TS)were reviewed with respect to the post trip review data.'he following are the results of that review:*Following the reactor trip, PRZR water level decreased'o approximately 12.5%due to a moderate RCS cooldown.This cooldown occurred during the post trip recovery period.This cooldown was bounded by the plant accident analysis, and did not exceed the TS limit of 100 degrees F per hour.Additional mitigation was provided by closing the MSIVs and stopping-the main feedwater pump.TS 3.1.1.5 states, in part, that when the RCS temperature is at or above 350 degrees F, at least 100 KW of PRZR heaters will be operable.TS 3.1.1.5 also states, in part, that if the PRZR,is inoperable due to heaters, restore the PRZR'to operable status within six (6)hours.PRZR water level was restored above the setpoint for letdown isolation within two (2)minutes, restoring the PRZR heaters to operable status, well before the six'(6)hour action statement.
NRC FORM 366A (5-92)
NRC FORH 366A (5.92)U.S.NUCLEAR REGULATORY COMHISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION APPROVED BY QGI NO.3150-0104 EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REOUEST: 50.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEHENT BRANCH (MNBB 7714), U.ST NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140 0104), OFFICE OF HANAGEMENT AND BUDGET WASHINGTON DC 20503.FACILITY NAHE 1 R.E.Ginna Nuclear Power Plant DOCKET NINBER 2 05000244 LER NUMBER 6 YEAR SEOUENTIAL 94--007--REVISION 00 PAGE 3 7 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)*Both S/G levels decreased following the reactor trip."A" S/G level decreased to<04, and"B" S/G level decreased to<14'.This is an expected transient.
TS 4.3.5.5 states that in order to demonstrate that a reactor coolant loop is operable, the S/G water level shall be>/=169.Thus, both coolant loops were.inoperable, even though both loops were still in operation and performing their intended function of decay heat removal.Both S/Gs were available as a heat sink, and sufficient AFW flow was maintained for adequate steam release from both S/Gs.TS 3.1.1.1(c) states, in part, that except for special tests, when the RCS temperature is at or above 350 degrees F with the reactor power less than or equal to 130 MWT (8.5%), at least one reactor coolant loop and its associated S/G and reactor coolant pump shall be in operation.
Both reactor coolant loops were in operation, but the S/Gs were inoperable due to level indication.
Both loops were restored to operable status when S/G levels were restored to>/=164 ("B" S/G level in less than four (4)minutes, and"A" S/G level in approximately ten (10)minutes).Based on the above and a review of post trip data and past plant transients, it can be concluded that the plant operated as designed, that there were no unreviewed safety questions, and that the public's health and safety was assured at all times.NRC FORM 366A (5-92)
NRC FORH 366A (5-92)U.S.NUCLEAR REGULATORY C(SOIISSIOH LICENSEE EVENT REPORT (LER)TEXT CONTINUATION APPROVED BY MB NO.3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.FORWARD COMMENTS REGARDIHG BURDEN'STIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MHBB 7714), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140-0104), OFFICE OF MANAGEMENT AND BUDGET'WASHINGTON DC 20503.FACILITY NAME 1 R.E.Ginna Nuclear Power Plant DOCKET NQIBER 2 05000244 YEAR 94--007--00 LER HINBER 6 SEQUENTIAL REVISION M PAGE 3 8 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)V.CORRECTIVE ACTION A.ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS: o The valve positioner for the"A" MFRV (Bailey Model AV112100I or"Type AV1")had been replaced on April 26, 1994, with another Type AV1 positioner.
This specific positioner failed in less than twelve hours of use and was replaced after this event (on April 28, 1994)with a Bailey Model 5321030A10 (or"5321030")
valve positioner.
The Bailey Model 5321030 is the original model of valve positioner for the MFRV application and had operated successfully since plant startup in 1969.This positioner model was changed out in 1991 as part of EWR 4773, which installed the Advanced Digital Feedwater Control System (ADFCS).The Type AV1 positioners have had a history of reliability and stability concerns in this MFRV application.
o The valve positioner for the"B" MFRV was also replaced on April 28, 1994, for the reasons discussed above.o Valve.ramp and step change diagnostic testing was performed for both MFRVs to verify proper valve positioning and response.B.ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:
o The positioners for both MFRVs were replaced with Model 5321030 positioners as discussed above.HRC FORM 366A (5-92)
NRC FORM 366A (5-92)I.S.NUCLEAR REGULATORY CQOIISSIOH LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PROVED BY OMB NO.3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTIOH REDUEST: 50.0 HRS.FORWARD COMMENTS REGARDIHG BURDEH ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MHBB 7714), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001 AHD TO THE PAPERWORK REDUCTION PROJECT (3140.0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.FACILITY NAME 1 R.E.Ginna Nuclear Power Plant DOCKET NUMBER 2 05000244 YEAR 94 LER NUMBER 6 SEOUENTIAL M--007--REVISION 00 PAGE 3 9 OF 9 TEXT (If more space is required, use additional copies of HRC Form 366A)(17)VI.ADDITIONAL INFORMATION A.FAILED COMPONENTS:
The failed component was the stroke adjust allen head set screw for the valve positioner signal diaphragm assembly for the"A" MFRV.The positioner is a Model AV112100 positioner, manufactured by Bailey Controls Inc.B.PREVIOUS LERs ON SIMILAR EVENTS: A similar LER event historical search was conducted with the following results: no documentation of similar LER events with the same root cause at Ginna Nuclear Power Plant could be iden-tified;however, LER 93-006 was a similar event, in that there was loss of ability to control the"A" MFRV;and LERs88-005, 90 007 I and 90-0 1 0 were simi lar events with dif f erent root causes.C.SPECIAL COMMENTS: None HRC FORM 366A (5-92)