|
|
(2 intermediate revisions by the same user not shown) |
Line 3: |
Line 3: |
| | issue date = 08/11/1997 | | | issue date = 08/11/1997 |
| | title = 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 | | | title = 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 |
| | author name = ST MARTIN J T | | | author name = St Martin J |
| | author affiliation = ROCHESTER GAS & ELECTRIC CORP. | | | author affiliation = ROCHESTER GAS & ELECTRIC CORP. |
| | addressee name = | | | addressee name = |
Line 17: |
Line 17: |
|
| |
|
| =Text= | | =Text= |
| {{#Wiki_filter:NRCRORM366(4-95)U.S.NUCLEARREGULATORY COMMISSIO LICENSEEEVENTREPORT(LER).(Seereverseforrequirednumberofdigits/characters foreachblock)APPROVEDBYOMBNO.3150-0104 | | {{#Wiki_filter:NRC RORM 366 U.S. NUCLEAR REGULATORY COMMISSIO APPROVED BY OMB NO. 3150-0104 (4-95) 'XPIRES 04/30/96 ESTIMATED BURDEN PER RESPONSE To COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS. |
| 'XPIRES04/30/96ESTIMATED BURDENPERRESPONSEToCOMPLYWITHTHISMANDATORY INFORMATION COLLECTION REQUEST:50.0HRS.REPORTEDLESSONSLEARNEDAREINCORPORATED INTOTHELICENSING PROCESSANDFEDBACKToINDUSTRY. | | LICENSEE EVENT REPORT (LER) REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK To INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE To THE |
| FORWARDCOMMENTSREGARDING BURDENESTIMATEToTHEINFORMATION ANDRECORDSMANAGEMENT BRANCHIT-6F33),U.S.NUCLEARREGULATORY COMMISSION, WASHINGTON, Dc20555-0001, ANDToTHEPAPERWORK REDUCTION PROJECTFACILITYNAME(1)TITLE(4)R.E.GinnaNuclearPowerPlantDOCKETNUMBER(2) 05000244PAGE(3)1OF7LeakOutsideContainment, DuetoWeldDefect,ResultsinLeakRateGreaterThanProgramLimitMONTHDAYYEAREVENTDATE(5)LERNUMBER(6)SEQUENT)AL REVISIONNUMBERNUMBERMONTHDAYYEARREPORTDATE(7)FACILITYNAMEOTHERFACILITIES INVOLVED(6)DOCKETNUMBER07239696-009-02081197FACILITYNAMEDOCKETNUMBEROPERATING MODE(9)POWERLEVEL(10)10020.2201(b) 20.2203(a)(1) 20.2203(a)
| | .(See reverse for required number of INFORMATION AND RECORDS MANAGEMENT BRANCH IT-6 F33), |
| (2)(i)20.2203(a) | | digits/characters for each block) U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, Dc 20555-0001, AND To THE PAPERWORK REDUCTION PROJECT FACILITYNAME (1) DOCKET NUMBER(2) PAGE (3) |
| (2)(v)20.2203(a) | | R.E. Ginna Nuclear Power Plant 05000244 1OF7 TITLE (4) |
| (3)(i)20.2203(a)(3)(ii) 50.73(a)(2)(i)50.73(a)(2) | | Leak Outside Containment, Due to Weld Defect, Results in Leak Rate Greater Than Program Limit EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (6) |
| (ii)50.73(a)(2)(iii)SUANTTOTHEREQUIREMENTS OF10CFRE:(ChackoneorTHISREPORTISSUBMITTED PURmore)(11)50.73(a)(2)(viii) 50.73(a)(2)(x) 73.7120.2203(a) | | FACILITYNAME DOCKET NUMBER SEQUENT)AL REVISION MONTH DAY YEAR NUMBER MONTH DAY YEAR NUMBER FACILITY NAME DOCKET NUMBER 07 23 96 96 009 02 08 11 97 OPERATING THIS REPORT IS SUBMITTED PUR SUANT TO THE REQUIREMENTS OF 10 CFR E: (Chack one or more) (11) |
| (2)(ii)20.2203(a)(2)(iii) 20.2203(a)(2)(iv) 20.2203(a)(4) 50.36(c)(1)50.36(c)(2)LICENSEECONTACTFORTHISLER(12)50.73(a)(2)(iv)50.73(a)(2)(v) 50.73(a)(2)(vii)OTHERSpecifyInAbet(ectbeloworinNRCForm365ATELEPHONE NUMBER(lne)udeAfeeCode)JohnT.St.Martin-Technical Assistant (716)771-3641COMPLETEONELINEFOREACHCOMPONENT FAILUREDESCRIBED INTHISREPORT(13)CAUSESYSTEMCOMPONENT MANUFACTURER REPORTABLE ToNPRDSCAUSESYSTEMCOMPONENT MANUFACTURER REPORTABLE ToNPRDSBBEPSF0000N:;t~g:~i'~c)NI."i?';.: | | MODE (9) 20.2201(b) 20.2203(a) (2)(v) 50.73(a) (2) (i) 50.73(a)(2)(viii) |
| SUPPLEMENTAL REPORTEXPECTED(14)YES(Ifyas,completeEXPECTEDSUBMISSION DATE).XNOEXPECTEDSUBMISSION DATE(16)MONTHDAYYEARABSTRACT(Limitto1400spaces,i.e.,approximately 15single-spaced typewritten lines)(16)OnJuly23,1996,atapproximately 0857EDST,itwasdetermined thattheleakrate(fromaleakonapipingsystemoutsideContainment) wasgreaterthantheanalyzedvalueoftwogallonsperhourforaprimarycoolantsourceoutsideContainment. | | POWER 20.2203(a)(1) 20.2203(a) (3)(i) 50.73(a)(2) (ii) 50.73(a)(2)(x) |
| Immediate corrective actionwastoconservatively enterTechnical Specification LimitingCondition forOperation 3.0.3andinitiateaplantshutdown, andtoinitiateactionstoisolatetheleakusingafreezeseal.Oncetheleakwasisolated, theplantexitedTechnical Specification 3.0.3andtheshutdownwasstopped.Theleakingpipewascutoutandreplaced. | | LEVEL (10) 100 20.2203(a) (2) (i) 20.2203(a)(3)(ii) 50.73(a) (2)(iii) 73.71 20.2203(a) (2) (ii) 20.2203(a)(4) 50.73(a) (2) (iv) OTHER 20.2203(a)(2)(iii) 50.36(c) (1) 50.73(a)(2)(v) Specify In Abet(ect below or in NRC Form 365A 20.2203(a)(2)(iv) 50.36(c) (2) 50.73(a) (2)(vii) |
| Theunderlying causeoftheleakwasawelddefectwhichdatesbacktooriginalconstruction inthe1960's.ThiseventisNUREG-1022 CauseCode(B).Corrective actiontopreventrecurrence isoutlinedinSectionV.B.Additional corrective actionsareidentified inthissupplement.
| | LICENSEE CONTACT FOR THIS LER (12) |
| 9708i80050 9708iiPDRADOCK050002448,PDR | | TELEPHONE NUMBER (lne)ude Afee Code) |
| | John T. St. Martin - Technical Assistant (716) 771-3641 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) |
| | SYSTEM COMPONENT REPORTABLE REPORTABLE CAUSE MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER To NPRDS To NPRDS B BE PSF 0000 N |
| | :;t~g:~i'~c)NI."i?';.: |
| | SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED YES SUBMISSION (If yas, complete EXPECTED SUBMISSION DATE). X NO DATE (16) |
| | ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16) |
| | On July 23, 1996, at approximately 0857 EDST, it was determined that the leak rate (from a leak on a piping system outside Containment) was greater than the analyzed value of two gallons per hour for a primary coolant source outside Containment. |
| | Immediate corrective action was to conservatively enter Technical Specification Limiting Condition for Operation 3.0.3 and initiate a plant shutdown, and to initiate actions to isolate the leak using a freeze seal. Once the leak was isolated, the plant exited Technical Specification 3.0.3 and the shutdown was stopped. The leaking pipe was cut out and replaced. |
| | The underlying cause of the leak was a weld defect which dates back to original construction in the 1960's. |
| | This event is NUREG-1022 Cause Code (B). |
| | Corrective action to prevent recurrence is outlined in Section V.B. Additional corrective actions are identified in this supplement. |
| | 9708i80050 9708ii PDR ADOCK 05000244 8 ,PDR |
|
| |
|
| NRCFORM366A(4.95)LICENSEEEVENTREPORT(LER)TEXTCONTINUATION U.S.NUCLEARREGULATORY COMMISSION FACILITYNAMEI1)R.E.GinnaNuclearPowerPlantDOCKET05000244LERNUMBERIe)YEARSEQUENTIAL REVISIONNUMBERNUMBER96-009-02PAGEI3)2OF7TEXT(IImorespaceisrequired, useadditional copiesofNRCForm366AJ(17)I.PRE-EVENT PLANTCONDITIONS:
| | NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4.95) |
| OnJuly22,1996,theplantwasinMode1atapproximately 100%steadystatereactorpower.Atapproximately 1130EDST,aRadiation Protection (RP)technician, whowasperforming aroutineweeklysurve'yoftheAuxiliary
| | LICENSEE EVENT REPORT (LER) |
| : Building, observedwaterdrippinginanareabehindtheRefueling WaterStorageTank(RWST).TheRPtechnician notifiedtheControlRoomoperators.
| | TEXT CONTINUATION FACILITY NAME I1) DOCKET LER NUMBER Ie) PAGE I3) |
| Effortstolocatethesourceofthedrippingwaterandtoquantifytheleakratewereinitiated byOperations andNuclearEngineering Services(NES).Thesourceofwaterwasdetermined tobeasmallleakonatestlinefortheContainment Spray(CS)systemintheAuxiliary
| | YEAR SEQUENTIAL REVISION NUMBER NUMBER 2 OF 7 R.E. Ginna Nuclear Power Plant 05000244 96 009 02 TEXT (II more space is required, use additional copies of NRC Form 366AJ (17) |
| : Building, whichisoutsidetheContainment (CNMT)~Theleakratewasestimated tobeapproximately one(1)gallonperhour(GPH).TheControlRoomoperators didnotidentifyanyrequirement thatwouldlimitplantoperations.
| | I. PRE-EVENT PLANT CONDITIONS: |
| TheShiftSupervisor notifiedOperations management, maintenance management, andplantandNESstaffoftheleak.Allreachedasimilarconclusion.
| | On July 22, 1996, the plant was in Mode 1 at approximately 100% steady state reactor power. At approximately 1130 EDST, a Radiation Protection (RP) technician, who was performing a routine weekly surve'y of the Auxiliary Building, observed water dripping in an area behind the Refueling Water Storage Tank (RWST). The RP technician notified the Control Room operators. Efforts to locate the source of the dripping water and to quantify the leak rate were initiated by Operations and Nuclear Engineering Services (NES). The source of water was determined to be a small leak on a test line for the Containment Spray (CS) system in the Auxiliary Building, which is outside the Containment (CNMT) The leak rate was |
| Thestaffworkedondeveloping andimplementing anactionplantoaddresstheproblem.Theissuesoflineintegrity andRWSToperability wereinvestigated byNESandLaboratory Inspection Services(LIS)personnel.
| | ~ |
| LISpersonnel measuredtheamountofpipecross-section andweldmetalremaining, andNESpersonnel determined thattherewasamplemetalintheaffectedareatoprovidesystemintegrity.
| | estimated to be approximately one (1) gallon per hour (GPH). |
| Thesourceoftheleakwasidentified asasmallpinholeleakonapipe-to-90 degreeforgedsocketelbowweldonatwo(2)inchnominalpipesizeSchedule10stainless steeltestlinefortheCSsystem.TheleakwasinapartoftheCSsystemthatcouldnotbeisolatedfromtheRWSTbyvalvemanipulations withoutmakingbothtrainsoftheemergency corecoolingsystem(ECCS)inoperable.
| | The Control Room operators did not identify any requirement that would limit plant operations. The Shift Supervisor notified Operations management, maintenance management, and plant and NES staff of the leak. All reached a similar conclusion. The staff worked on developing and implementing an action plan to address the problem. The issues of line integrity and RWST operability were investigated by NES and Laboratory Inspection Services (LIS) personnel. LIS personnel measured the amount of pipe cross-section and weld metal remaining, and NES personnel determined that there was ample metal in the affected area to provide system integrity. |
| Theleakwasmonitored throughout theremainder ofthedayandnight.Anoperability assessment hadbeenrequested bytheShiftSupervisor onJuly22,whentheexistence oftheleakhadfirstbeenidentified.
| | The source of the leak was identified as a small pinhole leak on a pipe-to-90 degree forged socket elbow weld on a two (2) inch nominal pipe size Schedule 10 stainless steel test line for the CS system. The leak was in a part of the CS system that could not be isolated from the RWST by valve manipulations without making both trains of the emergency core cooling system (ECCS) inoperable. The leak was monitored throughout the remainder of the day and night. |
| OnthemorningofJuly23,1996,thisassessment wasreviewedbythePlantOperations ReviewCommittee (PORC)~Thisassessment, whichwasbasedonconservative assumptions, concluded thattheRWSTandECCSwerestilloperablewiththeunisolated leak.Thisconclusion wassupported bythefactthattherewasstillamplemetalareainthecross-section thatcouldprovideintegrity ofthepipeduringaseismiceventandthatthetotalcombinedstresswaslessthantheyieldstrengthofthematerial.
| | An operability assessment had been requested by the Shift Supervisor on July 22, when the existence of the leak had first been identified. On the morning of July 23, 1996, this assessment was reviewed by the Plant Operations Review Committee (PORC) ~ This assessment, which was based on conservative assumptions, concluded that the RWST and ECCS were still operable with the unisolated leak. This conclusion was supported by the fact that there was still ample metal area in the cross-section that could provide integrity of the pipe during a seismic event and that the total combined stress was less than the yield strength of the material. |
| PORCrequested amoreaccurateestimateoftheleakrate.Theleakratewasmeasuredandestimated tobe2.6GPH.Thisleakrateisinexcessoftwo(2)GPH,whichisspecified inSurveillance TestProcedure PT-39,"LeakageEvaluation ofPrimaryCoolantSourcesOutsideContainment",
| | PORC requested a more accurate estimate of the leak rate. The leak rate was measured and estimated to be 2.6 GPH. This leak rate is in excess of two (2) GPH, which is specified in Surveillance Test Procedure PT-39, "Leakage Evaluation of Primary Coolant Sources Outside Containment", as the program limit for the maximum integrated leak rate outside CNMT from these sources. Since the leak rate exceeded the requirements of procedure PT-39, it was initially judged to be a violation of the "Primary Coolant Sources Outside Containment Program", as specified in the Ginna Station Improved Technical Specifications (ITS) |
| astheprogramlimitforthemaximumintegrated leakrateoutsideCNMTfromthesesources.Sincetheleakrateexceededtherequirements ofprocedure PT-39,itwasinitially judgedtobeaviolation ofthe"PrimaryCoolantSourcesOutsideContainment Program",
| | Section 5.5.2. The PORC chairman directed the Shift Supervisor to enter ITS Limiting Condition for Operation (LCO) 3.0.3, based on the conservative requirements of procedure PT-39 and Administrative Procedure A-52.4, "Control of Limiting Conditions for Operating Equipment", even though the plant was in compliance with all ITS LCOs. |
| asspecified intheGinnaStationImprovedTechnical Specifications (ITS)Section5.5.2.ThePORCchairmandirectedtheShiftSupervisor toenterITSLimitingCondition forOperation (LCO)3.0.3,basedontheconservative requirements ofprocedure PT-39andAdministrative Procedure A-52.4,"ControlofLimitingConditions forOperating Equipment",
| | NAC FOAM 356A I4-95) |
| eventhoughtheplantwasincompliance withallITSLCOs.NACFOAM356AI4-95)
| |
| NRCFORM366A(4-95)LICENSEEEVENTREPORT(LER)TEXTCONTINUATION U.S.NUCLEARREGULATORY COMMISSION FACILITYNAME(1)DOCKETLERNUMBER(6)PAGE(3)R.E.GinnaNuclearPowerPlant05000244YEARSEQUENTIAL REVISIONNUMBERNUMBER3OF796-009-02TEXTiifmorespaceisrequired, useaddiu'onal copiesofNRCForm366AJ(17)DESCRIPTION OFEVENT:A.DATESANDAPPROXIMATE TIMESOFMAJOROCCURRENCES:
| |
| oJuly22,1996,1130EDST:Aleakisidentified onalineoutsideofContainment.
| |
| July23,1996,0857EDST:EventDateandTimeandDiscovery DateandTime.July23,1996,0900EDST:Loadreduction isstarted.July23,1996,1440EDST:Integrated leakageoutsideContainment isdetermined tobelessthantwoGPH.Loadreduction isstopped.B.EVENT:OnJuly23,1996,atapproximately 0857EDST,withtheplantinMode1atapproximately 100%steadystatereactorpower,thePlantOperations ReviewCommittee (PORC)conservatively directedtheShiftSupervisor toenterITSLCO3.0.3.:TheShiftSupervisor directedtheControlRoomoperators toinitiateaplantshutdown.
| |
| Atapproximately 0900EDST,theControlRoomoperators initiated aplantshutdownperNormalOperating Procedure 0-2,1,"NormalShutdowntoHotShutdown".
| |
| TheMechanical Supportgroupinitiated actionstoprovideafreezesealtoisolatethesourceofleakagefromtheRWST.Afreezesealwasinitiated betweentheleakingsocketandtheCSpumpsuctionlinefromtheRWST.Atapproximately 1440EDSTonJuly23,1996,thefreezesealhadisolatedtheleak.ITSLCO3.0.3wasexitedandtheloadreduction wasstopped.Theaffectedpipeandsocketweldswerecutoutandanewprefabricated spoolpiece wasinstalled.
| |
| Theaffectedweldwasretainedforfailuremodeanalysis.
| |
| Aloadincreasetoreturntheplanttofullpowerwasinitiated atapproximateiy 1530EDSTonJuly23,1996.C.INOPERABLE STRUCTURES, COMPONENTS, ORSYSTEMSTHATCONTRIBUTED TOTHEEVENT:NoneD.OTHERSYSTEMSORSECONDARY FUNCTIONS AFFECTED'ue tothelocationandsizeoftheleak,anoperability assessment determined thatnoemergency safeguards systemfunctionwasmadeinoperable.
| |
| However,theleakwasnotisolablefromtheRWSTwithoutvalvemanipulations thatwouldhaveresultedintheinoperability oftheRWSTandbothtrainsoftheECCS.NACFOAM366A(4-95)
| |
|
| |
|
| NRCFORM366A(4-95)LICENSEEEVENTREPORT(LER)TEXTCONTINUATION U.S.NUCLEARREGULATORY COMMISSION FACILITYNAME(1)R.E.GinnaNuclearPowerPlantDOCKET05000244LERNUMBER(6)YEARSEQUENTIAL REVISIONNUMBERNUMBER96-009-02PAGE(3)4OF7TEXTiifmorespaceisrequired, useedditionalcopiesofNRCForm368A)(17)E.METHODOFDISCOVERY:
| | NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95) |
| Theleakwasfirstdiscovered byanRPtechnician duringaroutineweeklysurveyoftheAuxiliary Building.
| | LICENSEE EVENT REPORT (LER) |
| 'enotifiedtheControlRoomoperators, andthepreciselocationwasconfirmed byOperations andEngineering.
| | TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3) |
| Confirmation thattheleakratewasabove2GPHoccurredaftermeasurements performed atthedirection ofPORC.OPERATORACTION:WhennotifiedbyPORC,theShiftSupervisor directedtheControlRoomoperators toenterITSLCO3.0.3andtoinitiateaplantshutdown.
| | YEAR SEQUENTIAL REVISION NUMBER NUMBER 3 OF 7 R.E. Ginna Nuclear Power Plant 05000244 96 009 02 TEXT iifmore spaceis required, use addiu'onal copies of NRC Form 366A J (17) |
| 'heControlRoomoperators initiated aplantshutdownperprocedure 0-2.1.TheShiftSupervisor subsequently notifiedtheNRCper10CFR50.72(b)(1)(i)(A),non-emergency onehournotification, atapproximately 0951EDSTonJuly23,1996.Whentheleakwasisolated, ITSLCO3.0.3wasexited,andtheplantwasreturnedtofullpower.G.SAFETYSYSTEMRESPONSES:
| | DESCRIPTION OF EVENT: |
| NoneIII.CAUSEOFEVENT:A.IMMEDIATE CAUSE:Theimmediate causeofthecondition prohibited byTechnical Specifications wastheconservative determination byPORCtoenterITSLCO3.0.3duetotheteak.B.INTERMEDIATE CAUSE:Theintermediate causeoftheleakwasapinholeleakintheweldattaching thepipetothesocketelbow.C.ROOTCAUSE:Theaffectedpipe,socket,andfilletweldwereanalyzedforthefailuremode.Thepinholewasdetermined tobeanoriginalinstallation welddefect(datingfromthe1960's)thatexistedoveranareaofslaginclusion intheweld.Almostalltheslaghadbeenleachedfromthepinholechanneloverthe'ast25years.NACFOAM366A(4.95)
| | A. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES: |
| | o July 22, 1996, 1130 EDST: A leak is identified on a line outside of Containment. |
| | July 23, 1996, 0857 EDST: Event Date and Time and Discovery Date and Time. |
| | July 23, 1996, 0900 EDST: Load reduction is started. |
| | July 23, 1996, 1440 EDST: Integrated leakage outside Containment is determined to be less than two GPH. Load reduction is stopped. |
| | B. EVENT: |
| | On July 23, 1996, at approximately 0857 EDST, with the plant in Mode 1 at approximately 100% |
| | steady state reactor power, the Plant Operations Review Committee (PORC) conservatively directed the Shift Supervisor to enter ITS LCO 3.0.3.: The Shift Supervisor directed the Control Room operators to initiate a plant shutdown. At approximately 0900 EDST, the Control Room operators initiated a plant shutdown per Normal Operating Procedure 0-2,1, "Normal Shutdown to Hot Shutdown". |
| | The Mechanical Support group initiated actions to provide a freeze seal to isolate the source of leakage from the RWST. A freeze seal was initiated between the leaking socket and the CS pump suction line from the RWST. At approximately 1440 EDST on July 23, 1996, the freeze seal had isolated the leak. ITS LCO 3.0.3 was exited and the load reduction was stopped. The affected pipe and socket welds were cut out and a new prefabricated spoolpiece was installed. The affected weld was retained for failure mode analysis. |
| | A load increase to return the plant to full power was initiated at approximateiy 1530 EDST on July 23, 1996. |
| | C. INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT: |
| | None D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED'ue to the location and size of the leak, an operability assessment determined that no emergency safeguards system function was made inoperable. However, the leak was not isolable from the RWST without valve manipulations that would have resulted in the inoperability of the RWST and both trains of the ECCS. |
| | NAC FOAM 366A (4-95) |
|
| |
|
| NRCFORM366A(4-95)LXCENSEEEVENTREPORT(LER)TEXTCONTINUATION U.S.NUCLEARREGULATORY COMMISSION FACILITYNAME(1)DOCKETLERNUMBER(6)PAGE(3)R.E.GinnaNuclearPowerPlant05000244YEARSEQUENTIAL REVISIONNUMBERNUMBER5OF796-'-009-02TEXTilfmorespeceisrequired, useedditionel copiesofNRCForm366AJ(17)ThiseventisNUREG-1022 CauseCode(B),"Design,Manufacturing, Construction
| | NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95) |
| /Installation".
| | LICENSEE EVENT REPORT (LER) |
| ThesmallpinholeleakdoesnotmeettheNUMARC93-01,"Industry Guideline forMonitoring theEffectiveness ofMaintenance atNuclearPowerPlants",definition ofa"Maintenance Preventable Functional Failure"~IV.ANALYSISOFEVENT:Thiseventisreportable inaccordance with10CFR50.73,LicenseeEventReportSystem,item(a)(2)(i)(B),"Anyoperation orcondition prohibited bytheplant'sTechnical Specifications".
| | TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3) |
| Theintegrated leakrateinexcessof2GPHoutsideCNMTresultedinconservative entryintoITSLCO3.0.3.ITSLCO3.0.3isonlyapplicable torequirements contained inChapter3oftheITS.Therewasnolicenserequirement forentryintoITSLCO3.0.3.Nevertheless, sincetheplantvoluntarily enteredITSLCO3.0.3,thiscondition isreportable.
| | YEAR SEQUENTIAL REVISION NUMBER NUMBER 4 OF 7 R.E. Ginna Nuclear Power Plant 05000244 96 009 02 TEXT iifmore space is required, use eddi tional copies of NRC Form 368A) (17) |
| Anassessment wasperformed considering boththesafetyconsequences andimplications ofthisevent.Thefollowing aretheresultsandconclusions ofthisassessment:
| | E. METHOD OF DISCOVERY: |
| Therewerenooperational orsafetyconsequences orimplications attributed totheleakontheCStestlinebecause:Almostalltheslaghadbeenleachedfromthepinholechannelandtherewasnosignofcorrosion orstress(cracking) attackofthestainless steelmaterialadjacenttothepinhole.Sincethepinholedidnotdegradethecross-sectional areaoftheforgedsocketelbowandpipe,thestressesonthislinedue.totheconcurrent loadsofdeadweight,
| | Building. 'e The leak was first discovered by an RP technician during a routine weekly survey of the Auxiliary notified the Control Room operators, and the precise location was confirmed by Operations and Engineering. Confirmation that the leak rate was above 2 GPH occurred after measurements performed at the direction of PORC. |
| : pressure, seismic,andthermaleffectswerestillwithinthecodelimit.Seismicintegrity wasnotthreatened bytheleaknorbytheinstallation ofthefreezeseal.Thenatureofthisleakmechanism wouldnotleadtoanincreaseinholesizeabovethatinitially identified.
| | OPERATOR ACTION: |
| Thelinecontaining theaffectedelbowexperiences flowfromtheRefueling WaterStorageTank(RWST)totheContainment Spray(CS)eductorsuctionvalvesduringperiodicsurveillance testingandisnotsubjected toanyappreciable dynamicstressesduringtesting,norwoulditbeexpectedtoduringpost-accident operation oftheCSorsafetyinjection.
| | When notified by PORC, the Shift Supervisor directed the Control Room operators to enter ITS LCO 3.0.3 and to initiate a plant shutdown. 'he Control Room operators initiated a plant shutdown per procedure 0-2.1. |
| Theoffsiteexposureduetoalargebreaklossofcoolantaccident(LOCA)isanalyzedintheGinnaStationUpdatedFinalSafetyAnalysisReport(UFSAR).Thesubjectleaklocationisexposedtoradioactive sumpfluidfollowing 6smallorlargebreakLOCAduringrecirculation.
| | The Shift Supervisor subsequently notified the NRC per 10 CFR 50.72 (b) (1) (i) (A), non-emergency one hour notification, at approximately 0951 EDST on July 23, 1996. When the leak was isolated, ITS LCO 3.0.3 was exited, and the plant was returned to full power. |
| NRCFORM366A(4-95)
| | G. SAFETY SYSTEM RESPONSES: |
| | None III. CAUSE OF EVENT: |
| | A. IMMEDIATECAUSE: |
| | The immediate cause of the condition prohibited by Technical Specifications was the conservative determination by PORC to enter ITS LCO 3.0.3 due to the teak. |
| | B. INTERMEDIATE CAUSE: |
| | The intermediate cause of the leak was a pinhole leak in the weld attaching the pipe to the socket elbow. |
| | C. ROOT CAUSE: |
| | The affected pipe, socket, and fillet weld were analyzed for the failure mode. The pinhole was determined to be an original installation weld defect (dating from the 1960's) that existed over an area of slag inclusion in the weld. Almost all the slag had been leached from the pinhole channel over the'ast 25 years. |
| | NAC FOAM 366A (4.95) |
|
| |
|
| NRCrORM366A(4-95)LICENSEEEVENTREPORT(LER)TEXTCONTINUATION U.S.NUCLEARREGULATORY COMMISSION FACILITYNAME(1)R.E.GinnaNuclearPowerPlantDOCKET05000244LERNUMBER(6)YEARSEQUENTIAL REVISIONNUMBERNUMBER96-009-02PAGE(3)6OF7TEXTlifmorespaceisrequired, useadditional copiesofNRCForm386A/(17)Recentanalyseshavebeenperformed byWestinghouse torecalculate siteboundaryandcontrolroomdosesfromalargebreakLOCA,whichisthemostlimitingLOCAsizefordoseevaluations.
| | NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95) |
| Withanassumedrecirculation leakrateof2.75GPH,itwasdetermined thatthesiteboundaryandlowpopulation zonethyroidandwholebodydosesresulting fromthepostulated largebreakLOCAarewithinthe10CFR100doseguidelines of300REMthyroidand25REMwholebody,respectively.
| | LXCENSEE EVENT REPORT (LER) |
| TheControlRoomdosecalculations resultinaControlRoomthyroid,wholebodyandbetaskindosewithinthedoselimitsdefinedintheNRCStandardReviewPlan(SRP)Section6.4,whichare30REMthyroid,5REIVIwholebodyand30REMbetaskin.Thesecalculated dosesarealsowithinthedosevaluesstatedintheUFSAR.Basedontheabove,it'canbeconcluded thatthepublic'shealthandsafetywasassuredatalltimes.V.CORRECTIVE ACTION:A.ACTIONTAKENTORETURNAFFECTEDSYSTEMSTOPRE-EVENT NORMALSTATUS:0ThesourceoftheleakagewasisolatedfromtheRWSTbyfreezeseal,permitting exitfromITSLCO3.0.3.0Theaffectedpipeandsocketweldswerecutoutandreplacedwithanewprefabricated spoolpiece, andthefreezesealwasremoved.B.ACTIONTAKENORPLANNEDTOPREVENTRECURRENCE:
| | TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3) |
| 0FailureAnalysisoftheleakinthesocketweldhasbeenperformed.
| | YEAR SEQUENTIAL REVISION NUMBER NUMBER 5 OF 7 R.E. Ginna Nuclear Power Plant 05000244 96 -'- 009 02 TEXT ilfmore speceis required, use edditionel copies of NRC Form 366AJ (17) |
| Theadjacentoriginalexistingweldswerecutoutandreplaced, Additional corrective actionsareidentified inthissupplement andarelistedbelow:ThebasesforITSLCO3.6.1,"Containment",
| | This event is NUREG-1022 Cause Code (B), "Design, Manufacturing, Construction / Installation". |
| hasbeenrevisedtospecifythenewleakagelimitof2.75GPHforintegrated leakageoutsideCNMTfromtheresidualheatremoval(RHR),safetyinjection (Sl),andCSsystems,andtoconsidertheeffectofthisleakageonCNMToperabiity.
| | The small pinhole leak does not meet the NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants", definition of a "Maintenance Preventable Functional Failure" ~ |
| 2.TheprogramlimitforleakageoutsideCNMT,asstatedinprocedure PT-39,wasrevisedfrom2.0GPHto2.75GPH,toreflectthesenewleakagelimits.3.Procedure A-52Awasrevisedtoreflectthenewprogramlimit:theintegrated leakagemustbe<I=2.75GPHasdetermined byPT-39.Theconservative requirement toenterITSLCO3.0.3forleakageabovethislimitwaseliminated. | | IV. ANALYSIS OF EVENT: |
| NACFOAM366A(4-95)
| | This event is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (a) (2) (i) |
| NRCFORM366AI4.95)LXCENSEEEVENTREPORT(LER)TEXTCONTINUATION U.S.NUCLEARREGULATORY COMMISSION FACILITYNAME<1)R.E.GinnaNuclearPowerPlantDOCKET05000244LERNUMBERI6)YEARSEQUENTIAL REVISIONNUMBERNUMBER96-009-02PAGEI3)7OF7TEXTllfmorespaceisrequired, useadditional copiesofNRCForm866AJl17)Vl.ADDITIONAL INFORMATION:
| | (B), "Any operation or condition prohibited by the plant's Technical Specifications". The integrated leak rate in excess of 2 GPH outside CNMT resulted in conservative entry into ITS LCO 3.0.3. ITS LCO 3.0.3 is only applicable to requirements contained in Chapter 3 of the ITS. There was no license requirement for entry into ITS LCO 3.0.3. Nevertheless, since the plant voluntarily entered ITS LCO 3.0.3, this condition is reportable. |
| A.FAILEDCOMPONENTS: | | An assessment was performed considering both the safety consequences and implications of this event. |
| Thefailedcomponent wasa2inchnominalpipesizeSchedule10pipe-to-90 degreeforgedsocketelbowweld.Thepipeandsocketelbowarestainless steel,ASTMA312Type304andA182F304(forged)respectively.
| | The following are the results and conclusions of this assessment: |
| Thesocketweldwasanoriginalinstallation weldmadebefore1970.B.PREVIOUSLERsONSIIVIILAR EVENTS:AsimilarLEReventhistorical searchwasconducted withthefollowing results:Nodocumentation ofsimilarLEReventswiththesamerootcauseatGinnaNuclearPowerPlantcouldbeidentified.
| | There were no operational or safety consequences or implications attributed to the leak on the CS test line because: |
| C.SPECIALCOMMENTS: | | Almost all the slag had been leached from the pinhole channel and there was no sign of corrosion or stress (cracking) attack of the stainless steel material adjacent to the pinhole. |
| NoneNRCFORM366AI4.95)
| | Since the pinhole did not degrade the cross-sectional area of the forged socket elbow and pipe, the stresses on this line due.to the concurrent loads of deadweight, pressure, seismic, and thermal effects were still within the code limit. Seismic integrity was not threatened by the leak nor by the installation of the freeze seal. |
| I1}}
| | The nature of this leak mechanism would not lead to an increase in hole size above that initially identified. |
| | The line containing the affected elbow experiences flow from the Refueling Water Storage Tank (RWST) to the Containment Spray (CS) eductor suction valves during periodic surveillance testing and is not subjected to any appreciable dynamic stresses during testing, nor would it be expected to during post-accident operation of the CS or safety injection. |
| | The offsite exposure due to a large break loss of coolant accident (LOCA) is analyzed in the Ginna Station Updated Final Safety Analysis Report (UFSAR). The subject leak location is exposed to radioactive sump fluid following 6 small or large break LOCA during recirculation. |
| | NRC FORM 366A (4-95) |
| | |
| | NRC rORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95) |
| | LICENSEE EVENT REPORT (LER) |
| | TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3) |
| | YEAR SEQUENTIAL REVISION NUMBER NUMBER 6 OF 7 R.E. Ginna Nuclear Power Plant 05000244 96 009 02 TEXT lifmore space is required, use additional copies of NRC Form 386A/ (17) |
| | Recent analyses have been performed by Westinghouse to recalculate site boundary and control room doses from a large break LOCA, which is the most limiting LOCA size for dose evaluations. With an assumed recirculation leak rate of 2.75 GPH, it was determined that the site boundary and low population zone thyroid and whole body doses resulting from the postulated large break LOCA are within the 10 CFR 100 dose guidelines of 300 REM thyroid and 25 REM whole body, respectively. The Control Room dose calculations result in a Control Room thyroid, whole body and beta skin dose within the dose limits defined in the NRC Standard Review Plan (SRP) Section 6.4, which are 30 REM thyroid, 5 REIVI whole body and 30 REM beta skin. These calculated doses are also within the dose values stated in the UFSAR. |
| | Based on the above, it'can be concluded that the public's health and safety was assured at all times. |
| | V. CORRECTIVE ACTION: |
| | A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS: |
| | 0 The source of the leakage was isolated from the RWST by freeze seal, permitting exit from ITS LCO 3.0.3. |
| | 0 The affected pipe and socket welds were cut out and replaced with a new prefabricated spoolpiece, and the freeze seal was removed. |
| | B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE: |
| | Failure Analysis of the leak in the socket weld has been performed. |
| | The adjacent original existing welds were cut out and replaced, 0 Additional corrective actions are identified in this supplement and are listed below: |
| | The bases for ITS LCO 3.6.1, "Containment", has been revised to specify the new leakage limit of 2.75 GPH for integrated leakage outside CNMT from the residual heat removal (RHR), safety injection (Sl), and CS systems, and to consider the effect of this leakage on CNMT operabiity. |
| | : 2. The program limit for leakage outside CNMT, as stated in procedure PT-39, was revised from 2.0 GPH to 2.75 GPH, to reflect these new leakage limits. |
| | : 3. Procedure A-52A was revised to reflect the new program limit: the integrated leakage must be <I= 2.75 GPH as determined by PT-39. The conservative requirement to enter ITS LCO 3.0.3 for leakage above this limit was eliminated. |
| | NAC FOAM 366A (4-95) |
| | |
| | NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION I4.95) |
| | LXCENSEE EVENT REPORT (LER) |
| | TEXT CONTINUATION FACILITY NAME <1) DOCKET LER NUMBER I6) PAGE I3) |
| | YEAR SEQUENTIAL REVISION NUMBER NUMBER 7 OF 7 R.E. Ginna Nuclear Power Plant 05000244 96 009 02 TEXT llfmore spaceis required, use additional copies of NRC Form 866AJ l17) |
| | Vl. ADDITIONALINFORMATION: |
| | A. FAILED COMPONENTS: |
| | The failed component was a 2 inch nominal pipe size Schedule 10 pipe-to-90 degree forged socket elbow weld. The pipe and socket elbow are stainless steel, ASTM A312 Type 304 and A182 F304 (forged) respectively. The socket weld was an original installation weld made before 1970. |
| | B. PREVIOUS LERs ON SIIVIILAR 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 identified. |
| | C. SPECIAL COMMENTS: |
| | None NRC FORM 366A I4.95) |
| | |
| | I 1}} |
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 PipeML17264A991 |
Person / Time |
---|
Site: |
Ginna ![Constellation icon.png](/w/images/b/be/Constellation_icon.png) |
---|
Issue date: |
08/11/1997 |
---|
From: |
St Martin J ROCHESTER GAS & ELECTRIC CORP. |
---|
To: |
|
---|
Shared Package |
---|
ML17264A990 |
List: |
---|
References |
---|
LER-96-009, LER-96-9, NUDOCS 9708180050 |
Download: ML17264A991 (12) |
|
Similar Documents at Ginna |
---|
Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML17265A7541999-09-22022 September 1999 LER 99-011-00:on 990823,small Tears Were Discovered in Flexible Duct Work Connector at Inlet of CR HVAC Sys Return Air Fan (AKF08).Caused by in-leakage Greater than That Assumed.Implemented Temporary Mod 99-029.With 990922 Ltr ML17265A7431999-08-24024 August 1999 LER 99-004-01:on 990412,discovered That Containment Recirculation Fan Chevron Separator Vanes Were Installed Backwards.Caused by Improper Assembly by Mfg.Moisture Separator Vanes Were Dismantled & Correctly re-installed ML17265A7181999-07-23023 July 1999 LER 99-007-01:on 990423,reactor Trip Occurred Due to Instrument & Control Technicians Inadvertently Pulling Fuses from Wrong Nuclear Instrument Channel.Setpoint Adjustments Were Completed by Different Crew of Technicians ML17265A7081999-07-22022 July 1999 LER 98-003-02:on 980904,actuations of CR Emergency Air Treatment Sys Was Noted Due to Invalid Causes.Caused by Various Degraded Components in CR RM Sys.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored ML17265A7031999-07-19019 July 1999 LER 99-S01-00:on 990617,determined That Temporary Unescorted Access Had Been Granted to Contractor Employee.Caused by Incomplete Info Re Circumstances of Individual Military Separation.Individual Access Was Revoked.With 990719 Ltr ML17265A7021999-07-15015 July 1999 LER 99-010-00:on 990615,ventilation Isolation of Auxiliary Bldg Occurred When Auxiliary Bldg Gas Radiation Monitor R-14 Reached High Alarm Setpoint.Cr Operators Rest Auxiliary Bldg Ventilation Isolation Signal.With 990715 Ltr ML17265A6851999-06-21021 June 1999 LER 99-001-01:on 990222,deficiencies in NSSS Vendor steam- Line Brake Mass & Energy Release Analysis Results in Plant Being Outside Design Bases Occurred.Caused by Deficiencies in W.Temporary Administrative Replaced.With 990621 Ltr ML17265A6661999-06-0202 June 1999 LER 99-009-00:on 990503,instrumentation Declared Inoperable in Multiple Channels Resulted in Condition Prohibited by Ts. Caused by Unanticipated High Frequency AC Voltage Ripple. Entered TS LCO 3.0.3.With 990602 Ltr ML17309A6541999-05-27027 May 1999 LER 99-008-00:on 990427,overtemperature Delta T Reactor Trip Occurred Due to Faulted Bistable During Calibr of Redundant Channel.Plant Was Stabilized in Mode 3 & Faulted Bistable Was Subsequently Replaced.With 990527 Ltr ML17265A6631999-05-24024 May 1999 LER 99-007-00:on 990423,technicians Inadvertently Pulled Fuses from Wrong Nuclear Instrument Cahnnel,Causing Reactor Trip,Due to High Range Flux Trip.Caused by Personnel Error. Labeling Scheme Improved ML17265A6601999-05-21021 May 1999 LER 99-006-00:on 990421,start of turbine-driven Auxiliary Feedwater Pump Was Noted.Caused by MOV Being Left in Open Position.Closed Manual Isolation Valve to Secure Steam to Pump.With 990521 Ltr ML17265A6441999-05-13013 May 1999 LER 99-005-00:on 990413,undervoltage Signal of Safeguards Bus During Testing Resulted in Automatic Start of B Edg. Caused by Personnel Error.Blown Fuse Was Replaced & Offsite Power Was Restored to Safeguards Bus 17.With 990513 Ltr ML17265A6431999-05-12012 May 1999 LER 99-004-00:on 990412,discovered That Containment Recirculation Fan Moisture Separator Vanes Were Incorrectly Installed,Per 10CFR21.Caused by Improper Assembly by Mfg. Subject Vanes Were Dismantled & Correctly re-installed ML17265A6141999-03-31031 March 1999 LER 99-003-00:on 990301,two Main Steam non-return Check Valves Were Declared Inoperable Due to Exceedance of Acceptance Criteria.Caused by Changes in Methodology & Matls.Packing Gland Torque Will Be Adjusted.With 990331 Ltr ML17265A6131999-03-29029 March 1999 LER 99-002-00:on 990227,discovered That Surveillance Had Not Been Performed at Frequency,Per Ts.Caused by Personnel Error.Procedure O-6.13 Will Be Evaluated for Enhancement Documentation of Completion of ITS Srs.With 990329 Ltr ML17265A6061999-03-24024 March 1999 LER 99-001-00:on 990222,plant Was Noted Outside Design Basis.Caused by Deficiencies in NSSS Vendor Slb Mass & Energy Release.Placed Temporary Administrative Restriction 40 Degrees F Max on Screenhouse Bay Temp ML17265A4951998-12-21021 December 1998 LER 98-005-00:on 981120,loss of 34.5 Kv Offsite Power Circuit 751,resulted in Automatic Start of B Edg.Caused by Faulted Cable Splice.Performed Appropriate Actions of Abnormal Procedure AP-ELEC.1.With 981221 Ltr ML17265A4931998-12-17017 December 1998 LER 98-004-00:on 971030,determined That Improperly Performed Surveillance Resulted in Condition Prohibited by Ts.Caused by Procedure non-adherence.Appropriate Calibr Procedures Were Properly Performed with 24 H of Condition Discovery ML17265A4691998-11-25025 November 1998 LER 98-003-01:on 980904,actuations of CR Emergency Air Treatment Systems (Creats) Occurred.Caused by Radon build-up During Temp Inversion.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored to CR ML17265A4271998-10-0505 October 1998 LER 98-003-00:on 980904,actuations of CR Emergency Air Treatment Sys Occurred.Caused by Radon build-up During Temp Inversion.Air Samples Were Taken & Determined That Source of Radiation Was Naturally Occurring Radon.With 981005 Ltr ML17265A3671998-07-14014 July 1998 LER 98-002-00:on 971019,CR Emergency Air Treatment Sys Actuating Function Was Not Operable.Caused by Mispositioned Switch.Revised Procedure CPI-MON-R37.W/980714 Ltr ML17265A1921998-03-11011 March 1998 LER 98-001-00:on 980209,discovered That Boraflex Degradation in SPF Was Greater than Was Assumed.Caused by Dissolution of Boron on Boraflex Matrix,Per 10CFR50.21.Removed Spent Fuel Assemblies from Selected Degraded Storage Rack Cells ML17265A1641998-02-0606 February 1998 LER 97-007-01:on 971117,reactor Engineer Recognized That Neutron Flux Low Range Trip Circuitry for Channel Was Not in Tripped Condition as Required.Caused by Technical Inadequacies.Channel Defeat Will Be Identified ML17265A1601998-02-0606 February 1998 LER 97-006-01:on 971103,verification of B Concentration Was Not Performed Due to Misinterpretation of Event Sequence. Audible Count Rate Function Was Restored to Operable Status ML17264B1441997-12-17017 December 1997 LER 97-007-00:on 971117,NF Low Range Trip Circuitry for Channel N-44 Was Not Placed in Tripped Condition.Caused by Technical Inadequacies in Procedures.Implemented EWR 4862 to Resolve Design deficiency.W/971217 Ltr ML17264B1291997-12-0303 December 1997 LER 97-006-00:on 971103,NIS Audible Count Rate Function Was Inoperable.Caused by Misinterpretation of Event Sequence Due to Not Verifying Boron Concentration.B Verification Occurred Every 12 H Per ITS LCO Action 3.9.2.C.3.W/971203 Ltr ML17264B1271997-12-0101 December 1997 LER 97-005-00:on 971031,undetected Unblocking of SI Actuation Signal Occurred at Low Pressure Condition,Due to Faulty Bistable Which Resulted in Inadvertent SI Actuation Signal.Sias,Ci & CVI Signals Were Reset ML17264B1211997-11-24024 November 1997 LER 97-004-00:on 971024,radiation Monitor Alarm Were Noted Due to Higher than Normal Radioactive Gas Concentration Resulted in Cvi.New R-12 Alarm Setpoint Was Maintained for Duration of Refueling Outage ML17264B0461997-09-29029 September 1997 LER 97-003-01:on 970730,bistable Instrument Trip Setpoint Could Have Exceeded Allowable Value.Caused by Insufficient Existing Margin Between Trip Setpoint & Allowable Value. Held Switches in Tripped configuration.W/970929 Ltr ML17264B0111997-08-27027 August 1997 LER 97-003-00:on 970730,high Steam Flow Bistable Instrument Setpoint Plus Instrument Uncertainty Could Exceed Allowable Value in ITS Was Identified.Caused by Entry Into ITS LCO 3.0.3.Switches Placed in Tripped configuration.W/970827 Ltr ML17264A9941997-08-19019 August 1997 LER 97-002-00:on 970720,34.5 Kv Offsite Power Circuit 751 Was Lost.Caused by Automatic Actuation of B Emergency DG Due to Undervoltage on Safeguards Buses 16 & 17.Offsite Power Restored to Safeguards Buses 16 & 17.W/970819 Ltr ML17264A9911997-08-11011 August 1997 LER 96-009-02:on 960723,determined That Leak Rate Outside Containment Was Greater than Program Limit.Caused by Weld Defect.Isolated Leak & Cut Out & Replaced Leaking Pipe ML17264A8271997-03-0303 March 1997 LER 97-001-00:on 970131,discovered Service Water Temp Was Less than Specified Value.Caused by non-representative Method of Monitoring.Increased Water Temp in Screenhouse Bay to Greater than 35 Degrees F.W/970303 Ltr ML17264A8071997-01-22022 January 1997 LER 96-015-00:on 961223,discovered Thermally Induced Overpressure Transient Could Occur.Caused by Thermal Expansion of Fluid During Design Basis Accident Condition. Installed Relief Valve on Affected line.W/970122 Ltr ML17264A7471996-11-27027 November 1996 LER 96-013-00:on 961029,circuit Breakers Closed While in Mode 3 & Resulted in Condition Prohibited by TS Due to Personnel Error.Circuit Breakers for MOV-878B & MOV-878D Were re-opened.W/961127 Ltr ML17264A6051996-09-19019 September 1996 LER 96-012-00:on 960820,feedwater Transient Occurred,Due to Closure of Feedwater Regulating Valve,Causing Lo Lo Steam Generator Level Reactor Trip.Sgs Were Restored & Missing Screw in 1/P-476 Was replaced.W/960919 Ltr ML17264A6061996-09-19019 September 1996 LER 96-009-01:on 960723,leakage Outside Containment Occurred,Due to Weld Defect,Resulting in Leak Rate Greater than Program Limits.Source of Leakage Isolated from RWST by Freeze Seal,Allowing Exit from ITS LCO 3.0.3.W/960919 Ltr ML17264A5911996-09-0505 September 1996 LER 96-011-00:on 960807,improper Configuration of Circuit Breaker Occurred,Due to Undetected Internal Interference, Resulting in Automatic Start of Both Auxiliary Feedwater Pumps.Running AFW Pumps Were secured.W/960905 Ltr ML17264A5921996-09-0505 September 1996 LER 96-010-00:on 960806,latching of Main Turbine While in Mode 4 Occurred,Due to Defective Procedure,Resulting in Automatic Start of Auxiliary Feedwater Pump.Caused by Defective Maint Procedure.Procedure revised.W/960905 Ltr ML17264A5891996-08-22022 August 1996 LER 96-009-00:on 960723,determined Leak on Piping Sys Outside Containment Greater than Program Limit.Caused by Weld Defect.Pipe & Socket Welds Were Cut Out & Replaced. W/960822 Ltr ML17264A5781996-08-0606 August 1996 LER 96-008-00:on 960707,main Feedwater Pump Breakers Opened. Caused by Change in Seal Water Differential Pressure Occurred During Sys Realignment.Afw Flow Controlled as Desired to Maintain S/G level.W/960806 Ltr ML17264A5561996-07-12012 July 1996 LER 96-007-00:on 960612,CR Operators Identified Control Rods Misaligned & Not Moving in Proper Sequence.Caused by Faulty Firing Circuit Card in Rod Control Sys.Faulty Firing Circuit Card in 1BD Power Cabinet replaced.W/960712 Ltr ML17264A5421996-06-20020 June 1996 LER 96-006-00:on 960521,discovered Containment Penetration Not in Required Status.Caused by Personnel Error.Installed Flange Inside Containment Penetration 2.W/960620 Ltr ML17264A5411996-06-17017 June 1996 LER 96-005-00:on 960516,PORC Determined Deficient Procedures Do Not Meet SRs for Testing safety-related Logic Circuits. Caused by Inadequancies in Individual Testing Procedures. Procedures Re Improved TSs revised.W/960617 Ltr ML17264A5051996-05-17017 May 1996 LER 96-003-01:on 960308,identified That Both Pressurizer PORVs Inoperable Concurrently Due to Disconnection of Flex Hose to Both PORV Actuators to Install air-sets for Benchset & Limit Switch Activities.Hpes Completed ML17264A4481996-04-0808 April 1996 LER 96-003-00:on 960308,both Pressurizer Relief Valves Inoperable.Hpes Evaluation Is Being Conducted to Determined Cause of Event.C/As:Both PORVs restored.W/960408 Ltr ML17264A4471996-04-0808 April 1996 LER 96-002-00:on 960307,secondary Transient Occurred.Caused by Loss of B Condenser Circulating Water Pump.C/As: Thermography performed.W/960408 Ltr ML17264A4101996-03-18018 March 1996 LER 96-001-00:on 950504,inservice Test Not Performed During Refueling Outage.Caused by Inadequate Tracking of Surveillance Frequency.Valve Test Performed & Disassembled. W/960318 Ltr ML17264A2971995-12-14014 December 1995 LER 95-009-00:on 950817,surveillance Was Not Performed Due to Improper Application of TS Requirements Resulting in TS Violation.Testing of MOV-515 Was Performed on 951115.W/ 951214 Ltr ML17264A1711995-09-25025 September 1995 LER 95-008-00:on 950825,secondary Transient Occurred.Caused by Loss of B Condenser Circulating Water Pump That Resulted in Manual Rt.Returned S/G Levels to Normal Operating levels.W/950925 Ltr 1999-09-22
[Table view] Category:RO)
MONTHYEARML17265A7541999-09-22022 September 1999 LER 99-011-00:on 990823,small Tears Were Discovered in Flexible Duct Work Connector at Inlet of CR HVAC Sys Return Air Fan (AKF08).Caused by in-leakage Greater than That Assumed.Implemented Temporary Mod 99-029.With 990922 Ltr ML17265A7431999-08-24024 August 1999 LER 99-004-01:on 990412,discovered That Containment Recirculation Fan Chevron Separator Vanes Were Installed Backwards.Caused by Improper Assembly by Mfg.Moisture Separator Vanes Were Dismantled & Correctly re-installed ML17265A7181999-07-23023 July 1999 LER 99-007-01:on 990423,reactor Trip Occurred Due to Instrument & Control Technicians Inadvertently Pulling Fuses from Wrong Nuclear Instrument Channel.Setpoint Adjustments Were Completed by Different Crew of Technicians ML17265A7081999-07-22022 July 1999 LER 98-003-02:on 980904,actuations of CR Emergency Air Treatment Sys Was Noted Due to Invalid Causes.Caused by Various Degraded Components in CR RM Sys.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored ML17265A7031999-07-19019 July 1999 LER 99-S01-00:on 990617,determined That Temporary Unescorted Access Had Been Granted to Contractor Employee.Caused by Incomplete Info Re Circumstances of Individual Military Separation.Individual Access Was Revoked.With 990719 Ltr ML17265A7021999-07-15015 July 1999 LER 99-010-00:on 990615,ventilation Isolation of Auxiliary Bldg Occurred When Auxiliary Bldg Gas Radiation Monitor R-14 Reached High Alarm Setpoint.Cr Operators Rest Auxiliary Bldg Ventilation Isolation Signal.With 990715 Ltr ML17265A6851999-06-21021 June 1999 LER 99-001-01:on 990222,deficiencies in NSSS Vendor steam- Line Brake Mass & Energy Release Analysis Results in Plant Being Outside Design Bases Occurred.Caused by Deficiencies in W.Temporary Administrative Replaced.With 990621 Ltr ML17265A6661999-06-0202 June 1999 LER 99-009-00:on 990503,instrumentation Declared Inoperable in Multiple Channels Resulted in Condition Prohibited by Ts. Caused by Unanticipated High Frequency AC Voltage Ripple. Entered TS LCO 3.0.3.With 990602 Ltr ML17309A6541999-05-27027 May 1999 LER 99-008-00:on 990427,overtemperature Delta T Reactor Trip Occurred Due to Faulted Bistable During Calibr of Redundant Channel.Plant Was Stabilized in Mode 3 & Faulted Bistable Was Subsequently Replaced.With 990527 Ltr ML17265A6631999-05-24024 May 1999 LER 99-007-00:on 990423,technicians Inadvertently Pulled Fuses from Wrong Nuclear Instrument Cahnnel,Causing Reactor Trip,Due to High Range Flux Trip.Caused by Personnel Error. Labeling Scheme Improved ML17265A6601999-05-21021 May 1999 LER 99-006-00:on 990421,start of turbine-driven Auxiliary Feedwater Pump Was Noted.Caused by MOV Being Left in Open Position.Closed Manual Isolation Valve to Secure Steam to Pump.With 990521 Ltr ML17265A6441999-05-13013 May 1999 LER 99-005-00:on 990413,undervoltage Signal of Safeguards Bus During Testing Resulted in Automatic Start of B Edg. Caused by Personnel Error.Blown Fuse Was Replaced & Offsite Power Was Restored to Safeguards Bus 17.With 990513 Ltr ML17265A6431999-05-12012 May 1999 LER 99-004-00:on 990412,discovered That Containment Recirculation Fan Moisture Separator Vanes Were Incorrectly Installed,Per 10CFR21.Caused by Improper Assembly by Mfg. Subject Vanes Were Dismantled & Correctly re-installed ML17265A6141999-03-31031 March 1999 LER 99-003-00:on 990301,two Main Steam non-return Check Valves Were Declared Inoperable Due to Exceedance of Acceptance Criteria.Caused by Changes in Methodology & Matls.Packing Gland Torque Will Be Adjusted.With 990331 Ltr ML17265A6131999-03-29029 March 1999 LER 99-002-00:on 990227,discovered That Surveillance Had Not Been Performed at Frequency,Per Ts.Caused by Personnel Error.Procedure O-6.13 Will Be Evaluated for Enhancement Documentation of Completion of ITS Srs.With 990329 Ltr ML17265A6061999-03-24024 March 1999 LER 99-001-00:on 990222,plant Was Noted Outside Design Basis.Caused by Deficiencies in NSSS Vendor Slb Mass & Energy Release.Placed Temporary Administrative Restriction 40 Degrees F Max on Screenhouse Bay Temp ML17265A4951998-12-21021 December 1998 LER 98-005-00:on 981120,loss of 34.5 Kv Offsite Power Circuit 751,resulted in Automatic Start of B Edg.Caused by Faulted Cable Splice.Performed Appropriate Actions of Abnormal Procedure AP-ELEC.1.With 981221 Ltr ML17265A4931998-12-17017 December 1998 LER 98-004-00:on 971030,determined That Improperly Performed Surveillance Resulted in Condition Prohibited by Ts.Caused by Procedure non-adherence.Appropriate Calibr Procedures Were Properly Performed with 24 H of Condition Discovery ML17265A4691998-11-25025 November 1998 LER 98-003-01:on 980904,actuations of CR Emergency Air Treatment Systems (Creats) Occurred.Caused by Radon build-up During Temp Inversion.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored to CR ML17265A4271998-10-0505 October 1998 LER 98-003-00:on 980904,actuations of CR Emergency Air Treatment Sys Occurred.Caused by Radon build-up During Temp Inversion.Air Samples Were Taken & Determined That Source of Radiation Was Naturally Occurring Radon.With 981005 Ltr ML17265A3671998-07-14014 July 1998 LER 98-002-00:on 971019,CR Emergency Air Treatment Sys Actuating Function Was Not Operable.Caused by Mispositioned Switch.Revised Procedure CPI-MON-R37.W/980714 Ltr ML17265A1921998-03-11011 March 1998 LER 98-001-00:on 980209,discovered That Boraflex Degradation in SPF Was Greater than Was Assumed.Caused by Dissolution of Boron on Boraflex Matrix,Per 10CFR50.21.Removed Spent Fuel Assemblies from Selected Degraded Storage Rack Cells ML17265A1641998-02-0606 February 1998 LER 97-007-01:on 971117,reactor Engineer Recognized That Neutron Flux Low Range Trip Circuitry for Channel Was Not in Tripped Condition as Required.Caused by Technical Inadequacies.Channel Defeat Will Be Identified ML17265A1601998-02-0606 February 1998 LER 97-006-01:on 971103,verification of B Concentration Was Not Performed Due to Misinterpretation of Event Sequence. Audible Count Rate Function Was Restored to Operable Status ML17264B1441997-12-17017 December 1997 LER 97-007-00:on 971117,NF Low Range Trip Circuitry for Channel N-44 Was Not Placed in Tripped Condition.Caused by Technical Inadequacies in Procedures.Implemented EWR 4862 to Resolve Design deficiency.W/971217 Ltr ML17264B1291997-12-0303 December 1997 LER 97-006-00:on 971103,NIS Audible Count Rate Function Was Inoperable.Caused by Misinterpretation of Event Sequence Due to Not Verifying Boron Concentration.B Verification Occurred Every 12 H Per ITS LCO Action 3.9.2.C.3.W/971203 Ltr ML17264B1271997-12-0101 December 1997 LER 97-005-00:on 971031,undetected Unblocking of SI Actuation Signal Occurred at Low Pressure Condition,Due to Faulty Bistable Which Resulted in Inadvertent SI Actuation Signal.Sias,Ci & CVI Signals Were Reset ML17264B1211997-11-24024 November 1997 LER 97-004-00:on 971024,radiation Monitor Alarm Were Noted Due to Higher than Normal Radioactive Gas Concentration Resulted in Cvi.New R-12 Alarm Setpoint Was Maintained for Duration of Refueling Outage ML17264B0461997-09-29029 September 1997 LER 97-003-01:on 970730,bistable Instrument Trip Setpoint Could Have Exceeded Allowable Value.Caused by Insufficient Existing Margin Between Trip Setpoint & Allowable Value. Held Switches in Tripped configuration.W/970929 Ltr ML17264B0111997-08-27027 August 1997 LER 97-003-00:on 970730,high Steam Flow Bistable Instrument Setpoint Plus Instrument Uncertainty Could Exceed Allowable Value in ITS Was Identified.Caused by Entry Into ITS LCO 3.0.3.Switches Placed in Tripped configuration.W/970827 Ltr ML17264A9941997-08-19019 August 1997 LER 97-002-00:on 970720,34.5 Kv Offsite Power Circuit 751 Was Lost.Caused by Automatic Actuation of B Emergency DG Due to Undervoltage on Safeguards Buses 16 & 17.Offsite Power Restored to Safeguards Buses 16 & 17.W/970819 Ltr ML17264A9911997-08-11011 August 1997 LER 96-009-02:on 960723,determined That Leak Rate Outside Containment Was Greater than Program Limit.Caused by Weld Defect.Isolated Leak & Cut Out & Replaced Leaking Pipe ML17264A8271997-03-0303 March 1997 LER 97-001-00:on 970131,discovered Service Water Temp Was Less than Specified Value.Caused by non-representative Method of Monitoring.Increased Water Temp in Screenhouse Bay to Greater than 35 Degrees F.W/970303 Ltr ML17264A8071997-01-22022 January 1997 LER 96-015-00:on 961223,discovered Thermally Induced Overpressure Transient Could Occur.Caused by Thermal Expansion of Fluid During Design Basis Accident Condition. Installed Relief Valve on Affected line.W/970122 Ltr ML17264A7471996-11-27027 November 1996 LER 96-013-00:on 961029,circuit Breakers Closed While in Mode 3 & Resulted in Condition Prohibited by TS Due to Personnel Error.Circuit Breakers for MOV-878B & MOV-878D Were re-opened.W/961127 Ltr ML17264A6051996-09-19019 September 1996 LER 96-012-00:on 960820,feedwater Transient Occurred,Due to Closure of Feedwater Regulating Valve,Causing Lo Lo Steam Generator Level Reactor Trip.Sgs Were Restored & Missing Screw in 1/P-476 Was replaced.W/960919 Ltr ML17264A6061996-09-19019 September 1996 LER 96-009-01:on 960723,leakage Outside Containment Occurred,Due to Weld Defect,Resulting in Leak Rate Greater than Program Limits.Source of Leakage Isolated from RWST by Freeze Seal,Allowing Exit from ITS LCO 3.0.3.W/960919 Ltr ML17264A5911996-09-0505 September 1996 LER 96-011-00:on 960807,improper Configuration of Circuit Breaker Occurred,Due to Undetected Internal Interference, Resulting in Automatic Start of Both Auxiliary Feedwater Pumps.Running AFW Pumps Were secured.W/960905 Ltr ML17264A5921996-09-0505 September 1996 LER 96-010-00:on 960806,latching of Main Turbine While in Mode 4 Occurred,Due to Defective Procedure,Resulting in Automatic Start of Auxiliary Feedwater Pump.Caused by Defective Maint Procedure.Procedure revised.W/960905 Ltr ML17264A5891996-08-22022 August 1996 LER 96-009-00:on 960723,determined Leak on Piping Sys Outside Containment Greater than Program Limit.Caused by Weld Defect.Pipe & Socket Welds Were Cut Out & Replaced. W/960822 Ltr ML17264A5781996-08-0606 August 1996 LER 96-008-00:on 960707,main Feedwater Pump Breakers Opened. Caused by Change in Seal Water Differential Pressure Occurred During Sys Realignment.Afw Flow Controlled as Desired to Maintain S/G level.W/960806 Ltr ML17264A5561996-07-12012 July 1996 LER 96-007-00:on 960612,CR Operators Identified Control Rods Misaligned & Not Moving in Proper Sequence.Caused by Faulty Firing Circuit Card in Rod Control Sys.Faulty Firing Circuit Card in 1BD Power Cabinet replaced.W/960712 Ltr ML17264A5421996-06-20020 June 1996 LER 96-006-00:on 960521,discovered Containment Penetration Not in Required Status.Caused by Personnel Error.Installed Flange Inside Containment Penetration 2.W/960620 Ltr ML17264A5411996-06-17017 June 1996 LER 96-005-00:on 960516,PORC Determined Deficient Procedures Do Not Meet SRs for Testing safety-related Logic Circuits. Caused by Inadequancies in Individual Testing Procedures. Procedures Re Improved TSs revised.W/960617 Ltr ML17264A5051996-05-17017 May 1996 LER 96-003-01:on 960308,identified That Both Pressurizer PORVs Inoperable Concurrently Due to Disconnection of Flex Hose to Both PORV Actuators to Install air-sets for Benchset & Limit Switch Activities.Hpes Completed ML17264A4481996-04-0808 April 1996 LER 96-003-00:on 960308,both Pressurizer Relief Valves Inoperable.Hpes Evaluation Is Being Conducted to Determined Cause of Event.C/As:Both PORVs restored.W/960408 Ltr ML17264A4471996-04-0808 April 1996 LER 96-002-00:on 960307,secondary Transient Occurred.Caused by Loss of B Condenser Circulating Water Pump.C/As: Thermography performed.W/960408 Ltr ML17264A4101996-03-18018 March 1996 LER 96-001-00:on 950504,inservice Test Not Performed During Refueling Outage.Caused by Inadequate Tracking of Surveillance Frequency.Valve Test Performed & Disassembled. W/960318 Ltr ML17264A2971995-12-14014 December 1995 LER 95-009-00:on 950817,surveillance Was Not Performed Due to Improper Application of TS Requirements Resulting in TS Violation.Testing of MOV-515 Was Performed on 951115.W/ 951214 Ltr ML17264A1711995-09-25025 September 1995 LER 95-008-00:on 950825,secondary Transient Occurred.Caused by Loss of B Condenser Circulating Water Pump That Resulted in Manual Rt.Returned S/G Levels to Normal Operating levels.W/950925 Ltr 1999-09-22
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML17265A7601999-10-0505 October 1999 Part 21 Rept Re W2 Switch Supplied by W Drawn from Stock, Did Not Operate Properly After Being Installed on 990409. Switch Returned to W on 990514 for Evaluation & Root Cause Analysis ML17265A7621999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Re Ginna Npp.With 991008 Ltr ML17265A7531999-09-23023 September 1999 Part 21 Rept Re Corrective Action & Closeout of 10CFR21 Rept of Noncompliance Re Unacceptable Part for 30-4 Connector. Unacceptable Parts Removed from Stock & Scrapped ML17265A7541999-09-22022 September 1999 LER 99-011-00:on 990823,small Tears Were Discovered in Flexible Duct Work Connector at Inlet of CR HVAC Sys Return Air Fan (AKF08).Caused by in-leakage Greater than That Assumed.Implemented Temporary Mod 99-029.With 990922 Ltr ML17265A7471999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Re Ginna Npp.With 990909 Ltr ML17265A7431999-08-24024 August 1999 LER 99-004-01:on 990412,discovered That Containment Recirculation Fan Chevron Separator Vanes Were Installed Backwards.Caused by Improper Assembly by Mfg.Moisture Separator Vanes Were Dismantled & Correctly re-installed ML17265A7341999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Re Ginna Npp.With 990806 Ltr ML17265A7291999-07-29029 July 1999 Interim Part 21 Rept Re safety-related DB-25 Breaker Mechanism Procured from W Did Not Pas Degradatin Checks When Drawn from Stock to Be Installed Into BUS15/03A.Holes Did Not line-up & Tripper Pan Bent ML17265A7181999-07-23023 July 1999 LER 99-007-01:on 990423,reactor Trip Occurred Due to Instrument & Control Technicians Inadvertently Pulling Fuses from Wrong Nuclear Instrument Channel.Setpoint Adjustments Were Completed by Different Crew of Technicians ML17265A7081999-07-22022 July 1999 LER 98-003-02:on 980904,actuations of CR Emergency Air Treatment Sys Was Noted Due to Invalid Causes.Caused by Various Degraded Components in CR RM Sys.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored ML17265A7131999-07-22022 July 1999 Special Rept:On 990407,radiation Monitor RM-14A Was Declared Inoperable.Caused by Failed Communication Link from TSC to Plant Process Computer Sys.Communication Link Was re-established & RM-14A Was Declaed Operable on 990521 ML17265A7031999-07-19019 July 1999 LER 99-S01-00:on 990617,determined That Temporary Unescorted Access Had Been Granted to Contractor Employee.Caused by Incomplete Info Re Circumstances of Individual Military Separation.Individual Access Was Revoked.With 990719 Ltr ML17265A7211999-07-19019 July 1999 ISI Rept for Third Interval (1990-1999) Third Period, Second Outage (1999) at Re Ginna Npp. ML17265A7021999-07-15015 July 1999 LER 99-010-00:on 990615,ventilation Isolation of Auxiliary Bldg Occurred When Auxiliary Bldg Gas Radiation Monitor R-14 Reached High Alarm Setpoint.Cr Operators Rest Auxiliary Bldg Ventilation Isolation Signal.With 990715 Ltr ML17265A7661999-06-30030 June 1999 1999 Rept of Facility Changes,Tests & Experiments Conducted Without Prior NRC Approval for Jan 1998 Through June 1999, Per 10CFR50.59.With 991020 Ltr ML17265A7011999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Re Ginna Npp.With 990712 Ltr ML17265A6851999-06-21021 June 1999 LER 99-001-01:on 990222,deficiencies in NSSS Vendor steam- Line Brake Mass & Energy Release Analysis Results in Plant Being Outside Design Bases Occurred.Caused by Deficiencies in W.Temporary Administrative Replaced.With 990621 Ltr ML17265A6761999-06-16016 June 1999 Part 21 Rept Re Defects & noncompliances,10CFR21(d)(3)(ii), Which Requires Written Notification to NRC on Identification of Defect or Failure to Comply. Relays Were Returned to Eaton for Evaluation & Root Cause Analysis ML17265A6661999-06-0202 June 1999 LER 99-009-00:on 990503,instrumentation Declared Inoperable in Multiple Channels Resulted in Condition Prohibited by Ts. Caused by Unanticipated High Frequency AC Voltage Ripple. Entered TS LCO 3.0.3.With 990602 Ltr ML17265A6681999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Re Ginna Nuclear Power Plant.With 990608 Ltr ML17265A6651999-05-27027 May 1999 Interim Rept Re W2 Control Switch,Procured from W,Did Not Operate Satisfactorily When Drawn from Stock to Be Installed in Main Control Board for 1C2 Safety Injection Pump. Estimated That Evaluation Will Be Completed by 991001 ML17309A6541999-05-27027 May 1999 LER 99-008-00:on 990427,overtemperature Delta T Reactor Trip Occurred Due to Faulted Bistable During Calibr of Redundant Channel.Plant Was Stabilized in Mode 3 & Faulted Bistable Was Subsequently Replaced.With 990527 Ltr ML17265A6631999-05-24024 May 1999 LER 99-007-00:on 990423,technicians Inadvertently Pulled Fuses from Wrong Nuclear Instrument Cahnnel,Causing Reactor Trip,Due to High Range Flux Trip.Caused by Personnel Error. Labeling Scheme Improved ML17265A6601999-05-21021 May 1999 LER 99-006-00:on 990421,start of turbine-driven Auxiliary Feedwater Pump Was Noted.Caused by MOV Being Left in Open Position.Closed Manual Isolation Valve to Secure Steam to Pump.With 990521 Ltr ML17265A6591999-05-17017 May 1999 Part 21 Rept Re Relay Deficiency Detected During pre-installation Testing.Caused by Incorrectly Wired Relay Coil.Relays Were Returned to Eaton Corp for Investigation. Relays Were Repaired & Retested ML17265A6441999-05-13013 May 1999 LER 99-005-00:on 990413,undervoltage Signal of Safeguards Bus During Testing Resulted in Automatic Start of B Edg. Caused by Personnel Error.Blown Fuse Was Replaced & Offsite Power Was Restored to Safeguards Bus 17.With 990513 Ltr ML17265A6431999-05-12012 May 1999 LER 99-004-00:on 990412,discovered That Containment Recirculation Fan Moisture Separator Vanes Were Incorrectly Installed,Per 10CFR21.Caused by Improper Assembly by Mfg. Subject Vanes Were Dismantled & Correctly re-installed ML17265A6381999-05-0707 May 1999 Part 21 Rept Re Replacement Turbocharger Exhaust Turbine Side Drain Port Not Functioning as Design Intended.Caused by Manufacturing Deficiency.Turbocharger Was Reaasembled & Reinstalled on B EDG ML17265A6391999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Re Ginna Nuclear Power Plant.With 990510 Ltr ML17265A6361999-04-23023 April 1999 Part 21 Rept Re Power Supply That Did Not Work Properly When Drawn from Stock & Installed in -25 Vdc Slot.Power Supply Will Be Sent to Vendor to Perform Failure Mode Assessment.Evaluation Will Be Completed by 991001 ML17265A6301999-04-18018 April 1999 Rev 1 to Cycle 28 COLR for Re Ginna Npp. ML17265A6251999-04-15015 April 1999 Special Rept:On 990309,halon Systems Were Removed from Svc & Fire Door F502 Was Blocked Open.Caused by Mods Being Made to CR Emergency Air Treatment Sys.Continuous Fire Watch Was Established with Backup Fire Suppression Equipment ML17265A6551999-04-0909 April 1999 Initial Part 21 Rept Re Mfg Deficiency in Replacement Turbocharger for B EDG Supplied by Coltec Industries. Deficiency Consisted of Missing Drain Port in Intermediate Casing.Required Oil Drain Port Machined Open ML17265A6291999-03-31031 March 1999 Rev 0 to Cycle 28 COLR for Re Ginna Npp. ML17265A6241999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Ginna Station.With 990409 Ltr ML17265A6141999-03-31031 March 1999 LER 99-003-00:on 990301,two Main Steam non-return Check Valves Were Declared Inoperable Due to Exceedance of Acceptance Criteria.Caused by Changes in Methodology & Matls.Packing Gland Torque Will Be Adjusted.With 990331 Ltr ML17265A6131999-03-29029 March 1999 LER 99-002-00:on 990227,discovered That Surveillance Had Not Been Performed at Frequency,Per Ts.Caused by Personnel Error.Procedure O-6.13 Will Be Evaluated for Enhancement Documentation of Completion of ITS Srs.With 990329 Ltr ML17265A6061999-03-24024 March 1999 LER 99-001-00:on 990222,plant Was Noted Outside Design Basis.Caused by Deficiencies in NSSS Vendor Slb Mass & Energy Release.Placed Temporary Administrative Restriction 40 Degrees F Max on Screenhouse Bay Temp ML17265A5661999-03-0101 March 1999 Rev 26 to QA Program for Station Operation. ML17265A5961999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Ginna Nuclear Power Plant.With 990310 Ltr ML17265A5371999-01-31031 January 1999 Monthly Operating Rept for Jan 1999 for Re Ginna Nuclear Power Plant.With 990205 Ltr ML17265A5951998-12-31031 December 1998 Rg&E 1998 Annual Rept. ML17265A5001998-12-21021 December 1998 Rev 26 to QA Program for Station Operation. ML17265A4951998-12-21021 December 1998 LER 98-005-00:on 981120,loss of 34.5 Kv Offsite Power Circuit 751,resulted in Automatic Start of B Edg.Caused by Faulted Cable Splice.Performed Appropriate Actions of Abnormal Procedure AP-ELEC.1.With 981221 Ltr ML17265A4931998-12-17017 December 1998 LER 98-004-00:on 971030,determined That Improperly Performed Surveillance Resulted in Condition Prohibited by Ts.Caused by Procedure non-adherence.Appropriate Calibr Procedures Were Properly Performed with 24 H of Condition Discovery ML17265A4761998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Re Ginna Nuclear Power Plant.With 981210 Ltr ML17265A4691998-11-25025 November 1998 LER 98-003-01:on 980904,actuations of CR Emergency Air Treatment Systems (Creats) Occurred.Caused by Radon build-up During Temp Inversion.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored to CR ML17265A4531998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Re Ginna Nuclear Power Plant.With 981110 Ltr ML17265A4271998-10-0505 October 1998 LER 98-003-00:on 980904,actuations of CR Emergency Air Treatment Sys Occurred.Caused by Radon build-up During Temp Inversion.Air Samples Were Taken & Determined That Source of Radiation Was Naturally Occurring Radon.With 981005 Ltr ML17265A4291998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Re Ginna Nuclear Power Plant.With 981009 Ltr 1999-09-30
[Table view] |
Text
NRC RORM 366 U.S. NUCLEAR REGULATORY COMMISSIO APPROVED BY OMB NO. 3150-0104 (4-95) 'XPIRES 04/30/96 ESTIMATED BURDEN PER RESPONSE To COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS.
LICENSEE EVENT REPORT (LER) REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK To INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE To THE
.(See reverse for required number of INFORMATION AND RECORDS MANAGEMENT BRANCH IT-6 F33),
digits/characters for each block) U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, Dc 20555-0001, AND To THE PAPERWORK REDUCTION PROJECT FACILITYNAME (1) DOCKET NUMBER(2) PAGE (3)
R.E. Ginna Nuclear Power Plant 05000244 1OF7 TITLE (4)
Leak Outside Containment, Due to Weld Defect, Results in Leak Rate Greater Than Program Limit EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (6)
FACILITYNAME DOCKET NUMBER SEQUENT)AL REVISION MONTH DAY YEAR NUMBER MONTH DAY YEAR NUMBER FACILITY NAME DOCKET NUMBER 07 23 96 96 009 02 08 11 97 OPERATING THIS REPORT IS SUBMITTED PUR SUANT TO THE REQUIREMENTS OF 10 CFR E: (Chack one or more) (11)
MODE (9) 20.2201(b) 20.2203(a) (2)(v) 50.73(a) (2) (i) 50.73(a)(2)(viii)
POWER 20.2203(a)(1) 20.2203(a) (3)(i) 50.73(a)(2) (ii) 50.73(a)(2)(x)
LEVEL (10) 100 20.2203(a) (2) (i) 20.2203(a)(3)(ii) 50.73(a) (2)(iii) 73.71 20.2203(a) (2) (ii) 20.2203(a)(4) 50.73(a) (2) (iv) OTHER 20.2203(a)(2)(iii) 50.36(c) (1) 50.73(a)(2)(v) Specify In Abet(ect below or in NRC Form 365A 20.2203(a)(2)(iv) 50.36(c) (2) 50.73(a) (2)(vii)
LICENSEE CONTACT FOR THIS LER (12)
TELEPHONE NUMBER (lne)ude Afee Code)
John T. St. Martin - Technical Assistant (716) 771-3641 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
SYSTEM COMPONENT REPORTABLE REPORTABLE CAUSE MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER To NPRDS To NPRDS B BE PSF 0000 N
- t~g
- ~i'~c)NI."i?';.:
SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED YES SUBMISSION (If yas, complete EXPECTED SUBMISSION DATE). X NO DATE (16)
ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)
On July 23, 1996, at approximately 0857 EDST, it was determined that the leak rate (from a leak on a piping system outside Containment) was greater than the analyzed value of two gallons per hour for a primary coolant source outside Containment.
Immediate corrective action was to conservatively enter Technical Specification Limiting Condition for Operation 3.0.3 and initiate a plant shutdown, and to initiate actions to isolate the leak using a freeze seal. Once the leak was isolated, the plant exited Technical Specification 3.0.3 and the shutdown was stopped. The leaking pipe was cut out and replaced.
The underlying cause of the leak was a weld defect which dates back to original construction in the 1960's.
This event is NUREG-1022 Cause Code (B).
Corrective action to prevent recurrence is outlined in Section V.B. Additional corrective actions are identified in this supplement.
9708i80050 9708ii PDR ADOCK 05000244 8 ,PDR
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4.95)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME I1) DOCKET LER NUMBER Ie) PAGE I3)
YEAR SEQUENTIAL REVISION NUMBER NUMBER 2 OF 7 R.E. Ginna Nuclear Power Plant 05000244 96 009 02 TEXT (II more space is required, use additional copies of NRC Form 366AJ (17)
I. PRE-EVENT PLANT CONDITIONS:
On July 22, 1996, the plant was in Mode 1 at approximately 100% steady state reactor power. At approximately 1130 EDST, a Radiation Protection (RP) technician, who was performing a routine weekly surve'y of the Auxiliary Building, observed water dripping in an area behind the Refueling Water Storage Tank (RWST). The RP technician notified the Control Room operators. Efforts to locate the source of the dripping water and to quantify the leak rate were initiated by Operations and Nuclear Engineering Services (NES). The source of water was determined to be a small leak on a test line for the Containment Spray (CS) system in the Auxiliary Building, which is outside the Containment (CNMT) The leak rate was
~
estimated to be approximately one (1) gallon per hour (GPH).
The Control Room operators did not identify any requirement that would limit plant operations. The Shift Supervisor notified Operations management, maintenance management, and plant and NES staff of the leak. All reached a similar conclusion. The staff worked on developing and implementing an action plan to address the problem. The issues of line integrity and RWST operability were investigated by NES and Laboratory Inspection Services (LIS) personnel. LIS personnel measured the amount of pipe cross-section and weld metal remaining, and NES personnel determined that there was ample metal in the affected area to provide system integrity.
The source of the leak was identified as a small pinhole leak on a pipe-to-90 degree forged socket elbow weld on a two (2) inch nominal pipe size Schedule 10 stainless steel test line for the CS system. The leak was in a part of the CS system that could not be isolated from the RWST by valve manipulations without making both trains of the emergency core cooling system (ECCS) inoperable. The leak was monitored throughout the remainder of the day and night.
An operability assessment had been requested by the Shift Supervisor on July 22, when the existence of the leak had first been identified. On the morning of July 23, 1996, this assessment was reviewed by the Plant Operations Review Committee (PORC) ~ This assessment, which was based on conservative assumptions, concluded that the RWST and ECCS were still operable with the unisolated leak. This conclusion was supported by the fact that there was still ample metal area in the cross-section that could provide integrity of the pipe during a seismic event and that the total combined stress was less than the yield strength of the material.
PORC requested a more accurate estimate of the leak rate. The leak rate was measured and estimated to be 2.6 GPH. This leak rate is in excess of two (2) GPH, which is specified in Surveillance Test Procedure PT-39, "Leakage Evaluation of Primary Coolant Sources Outside Containment", as the program limit for the maximum integrated leak rate outside CNMT from these sources. Since the leak rate exceeded the requirements of procedure PT-39, it was initially judged to be a violation of the "Primary Coolant Sources Outside Containment Program", as specified in the Ginna Station Improved Technical Specifications (ITS)
Section 5.5.2. The PORC chairman directed the Shift Supervisor to enter ITS Limiting Condition for Operation (LCO) 3.0.3, based on the conservative requirements of procedure PT-39 and Administrative Procedure A-52.4, "Control of Limiting Conditions for Operating Equipment", even though the plant was in compliance with all ITS LCOs.
NAC FOAM 356A I4-95)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)
YEAR SEQUENTIAL REVISION NUMBER NUMBER 3 OF 7 R.E. Ginna Nuclear Power Plant 05000244 96 009 02 TEXT iifmore spaceis required, use addiu'onal copies of NRC Form 366A J (17)
DESCRIPTION OF EVENT:
A. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES:
o July 22, 1996, 1130 EDST: A leak is identified on a line outside of Containment.
July 23, 1996, 0857 EDST: Event Date and Time and Discovery Date and Time.
July 23, 1996, 0900 EDST: Load reduction is started.
July 23, 1996, 1440 EDST: Integrated leakage outside Containment is determined to be less than two GPH. Load reduction is stopped.
B. EVENT:
On July 23, 1996, at approximately 0857 EDST, with the plant in Mode 1 at approximately 100%
steady state reactor power, the Plant Operations Review Committee (PORC) conservatively directed the Shift Supervisor to enter ITS LCO 3.0.3.: The Shift Supervisor directed the Control Room operators to initiate a plant shutdown. At approximately 0900 EDST, the Control Room operators initiated a plant shutdown per Normal Operating Procedure 0-2,1, "Normal Shutdown to Hot Shutdown".
The Mechanical Support group initiated actions to provide a freeze seal to isolate the source of leakage from the RWST. A freeze seal was initiated between the leaking socket and the CS pump suction line from the RWST. At approximately 1440 EDST on July 23, 1996, the freeze seal had isolated the leak. ITS LCO 3.0.3 was exited and the load reduction was stopped. The affected pipe and socket welds were cut out and a new prefabricated spoolpiece was installed. The affected weld was retained for failure mode analysis.
A load increase to return the plant to full power was initiated at approximateiy 1530 EDST on July 23, 1996.
C. INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:
None D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED'ue to the location and size of the leak, an operability assessment determined that no emergency safeguards system function was made inoperable. However, the leak was not isolable from the RWST without valve manipulations that would have resulted in the inoperability of the RWST and both trains of the ECCS.
NAC FOAM 366A (4-95)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)
YEAR SEQUENTIAL REVISION NUMBER NUMBER 4 OF 7 R.E. Ginna Nuclear Power Plant 05000244 96 009 02 TEXT iifmore space is required, use eddi tional copies of NRC Form 368A) (17)
E. METHOD OF DISCOVERY:
Building. 'e The leak was first discovered by an RP technician during a routine weekly survey of the Auxiliary notified the Control Room operators, and the precise location was confirmed by Operations and Engineering. Confirmation that the leak rate was above 2 GPH occurred after measurements performed at the direction of PORC.
OPERATOR ACTION:
When notified by PORC, the Shift Supervisor directed the Control Room operators to enter ITS LCO 3.0.3 and to initiate a plant shutdown. 'he Control Room operators initiated a plant shutdown per procedure 0-2.1.
The Shift Supervisor subsequently notified the NRC per 10 CFR 50.72 (b) (1) (i) (A), non-emergency one hour notification, at approximately 0951 EDST on July 23, 1996. When the leak was isolated, ITS LCO 3.0.3 was exited, and the plant was returned to full power.
G. SAFETY SYSTEM RESPONSES:
None III. CAUSE OF EVENT:
A. IMMEDIATECAUSE:
The immediate cause of the condition prohibited by Technical Specifications was the conservative determination by PORC to enter ITS LCO 3.0.3 due to the teak.
B. INTERMEDIATE CAUSE:
The intermediate cause of the leak was a pinhole leak in the weld attaching the pipe to the socket elbow.
C. ROOT CAUSE:
The affected pipe, socket, and fillet weld were analyzed for the failure mode. The pinhole was determined to be an original installation weld defect (dating from the 1960's) that existed over an area of slag inclusion in the weld. Almost all the slag had been leached from the pinhole channel over the'ast 25 years.
NAC FOAM 366A (4.95)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)
LXCENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)
YEAR SEQUENTIAL REVISION NUMBER NUMBER 5 OF 7 R.E. Ginna Nuclear Power Plant 05000244 96 -'- 009 02 TEXT ilfmore speceis required, use edditionel copies of NRC Form 366AJ (17)
This event is NUREG-1022 Cause Code (B), "Design, Manufacturing, Construction / Installation".
The small pinhole leak does not meet the NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants", definition of a "Maintenance Preventable Functional Failure" ~
IV. ANALYSIS OF EVENT:
This event is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (a) (2) (i)
(B), "Any operation or condition prohibited by the plant's Technical Specifications". The integrated leak rate in excess of 2 GPH outside CNMT resulted in conservative entry into ITS LCO 3.0.3. ITS LCO 3.0.3 is only applicable to requirements contained in Chapter 3 of the ITS. There was no license requirement for entry into ITS LCO 3.0.3. Nevertheless, since the plant voluntarily entered ITS LCO 3.0.3, this condition is reportable.
An assessment was performed considering both the safety consequences and implications of this event.
The following are the results and conclusions of this assessment:
There were no operational or safety consequences or implications attributed to the leak on the CS test line because:
Almost all the slag had been leached from the pinhole channel and there was no sign of corrosion or stress (cracking) attack of the stainless steel material adjacent to the pinhole.
Since the pinhole did not degrade the cross-sectional area of the forged socket elbow and pipe, the stresses on this line due.to the concurrent loads of deadweight, pressure, seismic, and thermal effects were still within the code limit. Seismic integrity was not threatened by the leak nor by the installation of the freeze seal.
The nature of this leak mechanism would not lead to an increase in hole size above that initially identified.
The line containing the affected elbow experiences flow from the Refueling Water Storage Tank (RWST) to the Containment Spray (CS) eductor suction valves during periodic surveillance testing and is not subjected to any appreciable dynamic stresses during testing, nor would it be expected to during post-accident operation of the CS or safety injection.
The offsite exposure due to a large break loss of coolant accident (LOCA) is analyzed in the Ginna Station Updated Final Safety Analysis Report (UFSAR). The subject leak location is exposed to radioactive sump fluid following 6 small or large break LOCA during recirculation.
NRC FORM 366A (4-95)
NRC rORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)
YEAR SEQUENTIAL REVISION NUMBER NUMBER 6 OF 7 R.E. Ginna Nuclear Power Plant 05000244 96 009 02 TEXT lifmore space is required, use additional copies of NRC Form 386A/ (17)
Recent analyses have been performed by Westinghouse to recalculate site boundary and control room doses from a large break LOCA, which is the most limiting LOCA size for dose evaluations. With an assumed recirculation leak rate of 2.75 GPH, it was determined that the site boundary and low population zone thyroid and whole body doses resulting from the postulated large break LOCA are within the 10 CFR 100 dose guidelines of 300 REM thyroid and 25 REM whole body, respectively. The Control Room dose calculations result in a Control Room thyroid, whole body and beta skin dose within the dose limits defined in the NRC Standard Review Plan (SRP) Section 6.4, which are 30 REM thyroid, 5 REIVI whole body and 30 REM beta skin. These calculated doses are also within the dose values stated in the UFSAR.
Based on the above, it'can be concluded that the public's health and safety was assured at all times.
V. CORRECTIVE ACTION:
A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:
0 The source of the leakage was isolated from the RWST by freeze seal, permitting exit from ITS LCO 3.0.3.
0 The affected pipe and socket welds were cut out and replaced with a new prefabricated spoolpiece, and the freeze seal was removed.
B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:
Failure Analysis of the leak in the socket weld has been performed.
The adjacent original existing welds were cut out and replaced, 0 Additional corrective actions are identified in this supplement and are listed below:
The bases for ITS LCO 3.6.1, "Containment", has been revised to specify the new leakage limit of 2.75 GPH for integrated leakage outside CNMT from the residual heat removal (RHR), safety injection (Sl), and CS systems, and to consider the effect of this leakage on CNMT operabiity.
- 2. The program limit for leakage outside CNMT, as stated in procedure PT-39, was revised from 2.0 GPH to 2.75 GPH, to reflect these new leakage limits.
- 3. Procedure A-52A was revised to reflect the new program limit: the integrated leakage must be <I= 2.75 GPH as determined by PT-39. The conservative requirement to enter ITS LCO 3.0.3 for leakage above this limit was eliminated.
NAC FOAM 366A (4-95)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION I4.95)
LXCENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME <1) DOCKET LER NUMBER I6) PAGE I3)
YEAR SEQUENTIAL REVISION NUMBER NUMBER 7 OF 7 R.E. Ginna Nuclear Power Plant 05000244 96 009 02 TEXT llfmore spaceis required, use additional copies of NRC Form 866AJ l17)
Vl. ADDITIONALINFORMATION:
A. FAILED COMPONENTS:
The failed component was a 2 inch nominal pipe size Schedule 10 pipe-to-90 degree forged socket elbow weld. The pipe and socket elbow are stainless steel, ASTM A312 Type 304 and A182 F304 (forged) respectively. The socket weld was an original installation weld made before 1970.
B. PREVIOUS LERs ON SIIVIILAR 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 identified.
C. SPECIAL COMMENTS:
None NRC FORM 366A I4.95)
I 1