|
|
| Line 17: |
Line 17: |
|
| |
|
| =Text= | | =Text= |
| {{#Wiki_filter:NRCFORH366(5-92).S.NUCLEARREGULATORY COHHISSION PPROVEDBYOHBNO.3150-0104 EXPIRES5/31/95LICENSEEEVENTREPORT(LER)(Seereverseforrequirednumberofdigits/characters foreachblock)ESTIMATED BURDENPERRESPONSETOCOHPLYWITHTHISINFORMATION COLLECTION REQUEST:50.0HRS.FORWARDCOHMENTSREGARDING BURDENESTIMATETOTHEINFORMATION ANDRECORDSMANAGEMENT BRANCH(HNBB7714),U.STNUCLEARREGULATORY COHHISSION, WASHINGTON, DC20555-0001, ANDTOTHEPAPERWORK REDUCTION PROJECT(3150-0104), | | {{#Wiki_filter:NRC FORH 366 (5-92).S.NUCLEAR REGULATORY COHHISSION PPROVED BY OHB NO.3150-0104 EXPIRES 5/31/95 LICENSEE EVENT REPORT (LER)(See reverse for required number of digits/characters for each block)ESTIMATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.FORWARD COHMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (HNBB 7714), U.ST NUCLEAR REGULATORY COHHISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF HANAGEHENT AND BUDGET WASHINGTON DC 20503.FACILITY NAME (1)R.E.Ginna Nuclear Power Plant DOCKET NUMBER (2)05000244 PAGE (3)10F9 TITLE (4)Pressurizer Safety Valves Lift Settings Found Above Technical Specifications Tolerance During Post-service Test, Due to Setpoint Shifts, Results in Independent Trains Being Considered Inoperable HONTH 02 DAY YEAR 03 95 EVENT DATE (5)YEAR 95 LER NUHBER (6)SEQUENTIAL NUHBER--001--REVISION NUHBER 00 MONTH DAY 03 06 YEAR REPORT DATE (7)FACILITY NAHE DOCKET NUMBER FACILITY NAME DOCKET NUMBER OTHER FACILITIES INVOLVED (8)OPERATING HODE (9)POWER LEVEL (10)N 098 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) 50'6(c)(2)50.73(a)(2)(i) 50.73(a)(2)(ii) 50.73(a)(2)(iii) 50.73(a)(2)(iv) 50.73(a)(2)(v) |
| OFFICEOFHANAGEHENT ANDBUDGETWASHINGTON DC20503.FACILITYNAME(1)R.E.GinnaNuclearPowerPlantDOCKETNUMBER(2)05000244PAGE(3)10F9TITLE(4)Pressurizer SafetyValvesLiftSettingsFoundAboveTechnical Specifications Tolerance DuringPost-service Test,DuetoSetpointShifts,ResultsinIndependent TrainsBeingConsidered Inoperable HONTH02DAYYEAR0395EVENTDATE(5)YEAR95LERNUHBER(6)SEQUENTIAL NUHBER--001--REVISIONNUHBER00MONTHDAY0306YEARREPORTDATE(7)FACILITYNAHEDOCKETNUMBERFACILITYNAMEDOCKETNUMBEROTHERFACILITIES INVOLVED(8)OPERATING HODE(9)POWERLEVEL(10)N098THISREPORTISSUBMITTED 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) 50'6(c)(2)50.73(a)(2)(i) 50.73(a)(2)(ii) 50.73(a)(2)(iii) 50.73(a)(2)(iv) 50.73(a)(2)(v)
| | X 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 below and in Text, NRC Form 366A)TO THE REQUIREMENTS OF 10 CFR 5: (Check one or more)(11)LICENSEE CONTACT FOR THIS LER (12)NAME John T.St.Hartin-Technical Assistant TELEPHONE NUMBER (Include Area Code)(315)524-4446 COHPLETE ONE LINE FOR EACH COHPONENT FAILURE DESCRIBED IN THIS REPORT (13)CAUSE SYSTEH COMPONENT RV HANUFACTURER C170 REPORTABLE TO NPRDS CAUSE SYSTEH COMPONENT MANUFACTURER REPORTABLE TO NPRDS SUPPLEHENTAL REPORT EXPECTED (14)YES (If yes, complete EXPECTED SUBHISSION DATE).X NO EXPECTED SUB HISS ION DATE (15)MONTH DAY YEAR ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)(16)On February 3, 1995, at approximately 1824 EST, with the reactor at approximately 98'.steady state power, both pressurizer safety valves, which had been previously installed and then removed for testing, were considered inoperable. |
| X50.73(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(SpecifyinAbstractbelowandinText,NRCForm366A)TOTHEREQUIREMENTS OF10CFR5:(Checkoneormore)(11)LICENSEECONTACTFORTHISLER(12)NAMEJohnT.St.Hartin-Technical Assistant TELEPHONE NUMBER(IncludeAreaCode)(315)524-4446COHPLETEONELINEFOREACHCOHPONENT FAILUREDESCRIBED INTHISREPORT(13)CAUSESYSTEHCOMPONENT RVHANUFACTURER C170REPORTABLE TONPRDSCAUSESYSTEHCOMPONENT MANUFACTURER REPORTABLE TONPRDSSUPPLEHENTAL REPORTEXPECTED(14)YES(Ifyes,completeEXPECTEDSUBHISSION DATE).XNOEXPECTEDSUBHISSIONDATE(15)MONTHDAYYEARABSTRACT(Limitto1400spaces,i.e.,approximately 15single-spaced typewritten lines)(16)OnFebruary3,1995,atapproximately 1824EST,withthereactoratapproximately 98'.steadystatepower,bothpressurizer safetyvalves,whichhadbeenpreviously installed andthenremovedfortesting,wereconsidered inoperable.
| | Recent test results discovered that the"as-found" set pressure for the lift settings had shifted above the tolerance in the Technical Specifications. |
| Recenttestresultsdiscovered thatthe"as-found" setpressurefortheliftsettingshadshiftedabovethetolerance intheTechnical Specifications.
| | Immediate corrective.action was not required, since the valves were not installed. |
| Immediate corrective | | The underlying cause of the setpoint shift has been attributed to a combination of factors, including long-term operation, removal and shipping to an off-site facility for testing, as well as a restrictive tolerance in the Technical Specifications. |
| .actionwasnotrequired, sincethevalveswerenotinstalled. | | This event is NUREG-1022 Cause Code (B).Corrective action to preclude repetition is outlined in Section V.B.9503160241 950306 PDR ADOCK 05000244 8 PDR NRC FORH 366 (5-92) |
| Theunderlying causeofthesetpointshifthasbeenattributed toacombination offactors,including long-term operation, removalandshippingtoanoff-sitefacilityfortesting,aswellasarestrictive tolerance intheTechnical Specifications.
| | NRC FORM 366A (5-92).S.NUCLEAR REGULATORY COMHISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PPROVED BY OHB NO.3150.0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.FORWARD COHMENTS REGARDING BURDEN ESTIHATE TO THE INFORHATION AHD RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF HANAGEHEHT AND BUDGET WASHINGTON DC 20503.FACILITY NAHE (1)R.E.Ginna Nuclear Power Plant DOCKET NUHBER (2)05000244 YEAR 95 LER NUHBER (6)SEQUENTIAL 001 REVISION UM 00 PAGE (3)2 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)I.PRE-EVENT PLANT CONDITIONS: |
| ThiseventisNUREG-1022 CauseCode(B).Corrective actiontoprecluderepetition isoutlinedinSectionV.B.9503160241 950306PDRADOCK050002448PDRNRCFORH366(5-92)
| | The plant was at approximately 98'.steady state reactor power with no major activities in progress.Two new pressurizer (PRZR)code safety valves had been purchased, and were tested in the spring of 1993 at the valve manufacturer's test facilities. |
| NRCFORM366A(5-92).S.NUCLEARREGULATORY COMHISSION LICENSEEEVENTREPORT(LER)TEXTCONTINUATION PPROVEDBYOHBNO.3150.0104 EXPIRES5/31/95ESTIMATED BURDENPERRESPONSETOCOMPLYWITHTHISINFORMATION COLLECTION REQUEST:50.0HRS.FORWARDCOHMENTSREGARDING BURDENESTIHATETOTHEINFORHATION AHDRECORDSMANAGEMENT BRANCH(MNBB7714),U.S.NUCLEARREGULATORY COMMISSION, WASHINGTON, DC20555-0001, ANDTOTHEPAPERWORK REDUCTION PROJECT(3150-0104),
| | The valves were shipped to Rochester Gas and Electric (RGEE)with an"as-left" set pressure of 2485 psig (+/-1'.The original safety valves at Ginna Station were removed during the 1993 outage for annual testing, and these two new safety valves were installed. |
| OFFICEOFHANAGEHEHT ANDBUDGETWASHINGTON DC20503.FACILITYNAHE(1)R.E.GinnaNuclearPowerPlantDOCKETNUHBER(2)05000244YEAR95LERNUHBER(6)SEQUENTIAL 001REVISIONUM00PAGE(3)2OF9TEXT(Ifmorespaceisrequired, useadditional copiesofNRCForm366A)(17)I.PRE-EVENT PLANTCONDITIONS:
| | These valves (V-434 and V-435)were then considered operable during the 1993/1994 operating cycle (cycle 23)~These two valves were then removed for annual lift testing during the 1994 outage, and the original pair of safety valves (which had been tested in 1994)were installed for the 1994/1995 operating cycle (cycle 24).The removed valves were shipped to a test facility in Huntsville, Alabama, for testing, as per RGEE purchase order NQ-14349-C-JW. |
| Theplantwasatapproximately 98'.steadystatereactorpowerwithnomajoractivities inprogress.
| | The valves were tested to the requirements of RGGE Test Specification MET-049,"Pressurizer Safety Relief Valve Setpoint Testing", with steam as the test medium, on January 10, 1995 (for V-434)and January 11, 1995 (for V-435).RGEE Quality Assurance (QA)witnessed the tests.The test results showed that the"as-found" setpoints were 2525 psig (for V-434)and 2543 psig (for V-435), which exceeded the 1.lift setting tolerance of Technical Specifications. |
| Twonewpressurizer (PRZR)codesafetyvalveshadbeenpurchased, andweretestedinthespringof1993atthevalvemanufacturer's testfacilities.
| | These results were recognized as nonconforming, and a Nonconformance Report (NCR 95-005)was initiated to document this condition. |
| ThevalveswereshippedtoRochester GasandElectric(RGEE)withan"as-left" setpressureof2485psig(+/-1'.TheoriginalsafetyvalvesatGinnaStationwereremovedduringthe1993outageforannualtesting,andthesetwonewsafetyvalveswereinstalled.
| | On February 3, 1995, during review of NCR 95-005 by System Engineering and Nuclear Engineering Services (NES), it was determined that this represented a potentially reportable condition. |
| Thesevalves(V-434andV-435)werethenconsidered operableduringthe1993/1994 operating cycle(cycle23)~Thesetwovalveswerethenremovedforannuallifttestingduringthe1994outage,andtheoriginalpairofsafetyvalves(whichhadbeentestedin1994)wereinstalled forthe1994/1995 operating cycle(cycle24).TheremovedvalveswereshippedtoatestfacilityinHuntsville, Alabama,fortesting,asperRGEEpurchaseorderNQ-14349-C-JW.
| | NRC FORM 366A (5-92) |
| Thevalvesweretestedtotherequirements ofRGGETestSpecification MET-049,"Pressurizer SafetyReliefValveSetpointTesting",
| | NRC FORM 366A (5-92).S.NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT'(LER) |
| withsteamasthetestmedium,onJanuary10,1995(forV-434)andJanuary11,1995(forV-435).RGEEQualityAssurance (QA)witnessed thetests.Thetestresultsshowedthatthe"as-found" setpoints were2525psig(forV-434)and2543psig(forV-435),whichexceededthe1.liftsettingtolerance ofTechnical Specifications.
| | TEXT CONTINUATION PPROVED BY OHB NO.3150~0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPOHSE TO.COMPLY WITH THIS IHFORHATION COLLECTIOH REQUEST: 50.0 HRS.FORWARD COHHENTS REGARDIHG BURDEN ESTIHATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MHBB 7714), UPS.NUCLEAR REGULATORY COHHISSION,'WASHINGTON, DC 20555-0001 AHD-TO THE PAPERWORK REDUCTION PROJECT (31i0-0104), OFFICE OF HAHAGEMENT AHD BUDGET WASHINGTON DC 20503.FACILITY NAME (1)R.E.Ginna Nuclear Power Plant DOCKET NUHBER (2)05000244 YEAR 95 001 00 LER NUHBER (6)SEQUENTIAL REVISION PAGE (3)3 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)A.DESCRIPTION OF EVENT: DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES: |
| Theseresultswererecognized asnonconforming, andaNonconformance Report(NCR95-005)wasinitiated todocumentthiscondition.
| | 1.March, 1993: Newly procured PRZR safety valves are satisfactorily tested at manufacturer's test facility.2.April, 1993: Newly procured PRZR safety valves are installed for the 1993/1994 operating cycle (cycle 23).3.January 11, 1995: Testing of safety valves completed at off-site testing facility.Test results show that the lift pressure exceeded the lift setting tolerance. |
| OnFebruary3,1995,duringreviewofNCR95-005bySystemEngineering andNuclearEngineering Services(NES),itwasdetermined thatthisrepresented apotentially reportable condition.
| | Event date and time.4.February 3, 1995, 1824 EST: Test results are reviewed with the System Engineer.Discovery date and time.5.February 3, 1995, 2011 EST: Shift Supervisor notifies NRC per 10 CFR 50.72.B.EVENT: On February 3, 1995, at approximately 1824 EST, the reactor was at approximately 98'.steady state reactor power, and no major activities were in progress.NES personnel, from Mechanical Engineering and Nuclear Safety and Licensing (NSEL), were reviewing the status of NCR 95-005 with the System Engineer.Review of the NCR suggested'an operability question involving these previously, installed safety valves.Since both valves were previously installed during cycle'23, it was conservatively assumed that the valves had shifted out of tolerance during cycle 23.HRC FORH 366A (5-92) |
| NRCFORM366A(5-92)
| | NRC FORH 366A (5-92).S.NUCLEAR REGULATORY COHHISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PROVED BY OHB NO.3150~0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLEC'TION REOUEST: 50.0 HRS.FORWARD COMHENTS REGARDING BURDEN ESTIHATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S.NUCLEAR REGULATORY COHHISSION, WASHINGTON, DC 20555-0001, AHD TO THE PAPERWORK REDUCTION PROJECT (3150.0104), OFFICE OF MANAGEHENT AND BUDGET WASHINGTON DC 20503.FACILITY NAHE (1)R.E.Ginna Nuclear Power Plant DOCKET NUMBER (2)05000244 LER NUMBER (6)SEOUENTIAL 95--001--REVISION 00 PAGE (3)4 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)The"as-found" setpoints were 1.6%(for V-434)and 2.3%(for V-435)above 2485 psig.This is contrary to Ginna Technical Specification 3.1.1.3.c, which states,"Whenever the reactor is at or above an RCS temperature of 350 degrees F, both pressurizer code safety valves shall be operable with a lift setting of 2485 psig+/-1%." A conservative decision was made to report this event under the criteria of 10 CFR 50.72 (b)(2)(iii)(D), based on input from NS&L that a+/-1%tolerance for safety valve actuation is an assumption for several design basis events.The NRC was notified at approximately 2011 EST on February 3.Subsequent evaluations have not been able to conclusively determine the time that the setpoint shift occurred, nor even if the shift occurred during cycle 23.Review of design basis events has confirmed that this condition does not meet the reporting criteria of 10 CFR 50.72.INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT: None D.OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED: None METHOD OF DISCOVERY: |
| NRCFORM366A(5-92).S.NUCLEARREGULATORY COMMISSION LICENSEEEVENTREPORT'(LER)
| | RG&E QA Surveillance of the test data identified that the test results were unacceptable. |
| TEXTCONTINUATION PPROVEDBYOHBNO.3150~0104EXPIRES5/31/95ESTIMATED BURDENPERRESPOHSETO.COMPLYWITHTHISIHFORHATION COLLECTIOH REQUEST:50.0HRS.FORWARDCOHHENTSREGARDIHG BURDENESTIHATETOTHEINFORMATION ANDRECORDSMANAGEMENT BRANCH(MHBB7714),UPS.NUCLEARREGULATORY COHHISSION,
| | This information |
| 'WASHINGTON, DC20555-0001 AHD-TOTHEPAPERWORK REDUCTION PROJECT(31i0-0104), | | 'was forwarded to NES, and NCR 95-005 was initiated. |
| OFFICEOFHAHAGEMENT AHDBUDGETWASHINGTON DC20503.FACILITYNAME(1)R.E.GinnaNuclearPowerPlantDOCKETNUHBER(2)05000244YEAR9500100LERNUHBER(6)SEQUENTIAL REVISIONPAGE(3)3OF9TEXT(Ifmorespaceisrequired, useadditional copiesofNRCForm366A)(17)A.DESCRIPTION OFEVENT:DATESANDAPPROXIMATE TIMESOFMAJOROCCURRENCES:
| | During a review of this NCR between NES and System Engineering, this condition was evaluated as potentially reportable. |
| 1.March,1993:NewlyprocuredPRZRsafetyvalvesaresatisfactorily testedatmanufacturer's testfacility. | | HRC FORH 366A (5-92) |
| 2.April,1993:NewlyprocuredPRZRsafetyvalvesareinstalled forthe1993/1994 operating cycle(cycle23).3.January11,1995:Testingofsafetyvalvescompleted atoff-sitetestingfacility. | | NRC FORM 366A (5-92>U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME (1)DOCKET NUMBER (2)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PPROVED BY OHB NO.3150.0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REOUEST: 50'NRS.FORWARD COHHENTS REGARDIHG BURDEN ESTIHATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (HNBB 7714), U.S.NUCLEAR REGULATORY COMHISSION, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140 0104>, OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.PAGE (3)LER NUHBER (6)R.E.Ginna Nuclear Power Plant 05000244 YEAR 95 001 00 SEOUENTIAL REVISION 5 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)F.OPERATOR ACTION: The System Engineer notified the Shift Supervisor of the test results that affected both PRZR safety valves that were previously installed and considered operable during cycle 23, and that these results did not affect currently installed equipment. |
| Testresultsshowthattheliftpressureexceededtheliftsettingtolerance.
| | A decision was made to notify the NRC per 10 CFR 50.72 (b)(2)(iii)(D).This notification was made at approximately 2011 EST on February 3, 1995.Since this event did not affect installed plant equipment, no other operator, actions were necessary. |
| Eventdateandtime.4.February3,1995,1824EST:TestresultsarereviewedwiththeSystemEngineer.
| | SAFETY SYSTEM RESPONSES: |
| Discovery dateandtime.5.February3,1995,2011EST:ShiftSupervisor notifiesNRCper10CFR50.72.B.EVENT:OnFebruary3,1995,atapproximately 1824EST,thereactorwasatapproximately 98'.steadystatereactorpower,andnomajoractivities wereinprogress. | | None III.CAUSE OF EVENT: IMMEDIATE CAUSE: The immediate cause for both PRZR safety valves being considered inoperable was that the"as-found" lift settings for these valves were above the setpoint tolerance of Technical Specification 3.1.1.3.c. |
| NESpersonnel, fromMechanical Engineering andNuclearSafetyandLicensing (NSEL),werereviewing thestatusofNCR95-005withtheSystemEngineer.
| | B.INTERMED1ATE CAUSE: The intermediate cause for the"as-found" lift settings above the'etpoint tolerance of Technical Specifications was a shift in the setpoints from the"as-left" conditions of March, 1993, to the"as-found" conditions of January, 1995.NRC FORH 366A (5-92> |
| ReviewoftheNCRsuggested
| | NRC FORM 366A.(5-92).S.NUCLEAR REGULATORY COHHISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PROVED BY OHB NO.3150.0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORHATION COLLECTION REQUEST: 50.0 NRS.FORWARD COHHENTS REGARDING BURDEH ESTIHATE TO THE INFORHATION AND RECORDS HANAGEHENT BRANCH (HNBB 7714), U.S.NUCLEAR REGULATORY COHHISSIONg WASHINGTON, DC 20555-0001, AHD TO THE PAPERWORK REDUCTIOH PROJECT (3150-0104), OFFICE OF HANAGEHENT AND BUDGET WASHINGTON DC 20503.FACILITY NAHE (1)R.E.Ginna Nuclear Power Plant DOCKET NUHBER (2)05000244 YEAR 95 001--00 LER NUHBER (6)SEQUENTIAL REVISION PAGE (3)6 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)C.ROOT CAUSE: The underlying cause of the shift in the setpoints is attributed to a combination of factors, including attendant variables affecting the long-term operation of the valves, and the subsequent removal, decontamination, handling, and shipping of the valves to an off-site facility for testing..The Ginna Technical Specification requirements of+/-1-'.may not be appropriate with respect to the allowances for normal setpoint shifts during operation, removal, and shipping.This event is NUREG-1022 Cause Code (B),"Design Manufacturing, Construction |
| 'anoperability questioninvolving thesepreviously, installed safetyvalves.Sincebothvalveswerepreviously installed duringcycle'23,itwasconservatively assumedthatthevalveshadshiftedoutoftolerance duringcycle23.HRCFORH366A(5-92) | | '/Installation." IV.ANALYSIS OF EVENT: This event is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (a)(2)(vii)(D), which requires a report of,"any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to...mitigate the consequences of an accident." Both independent trains of pressure relief for the PRZR were considered inoperable due to the"as-found" lift settings above the tolerance of the Technical Specifications. |
| NRCFORH366A(5-92).S.NUCLEARREGULATORY COHHISSION LICENSEEEVENTREPORT(LER)TEXTCONTINUATION PROVEDBYOHBNO.3150~0104EXPIRES5/31/95ESTIMATED BURDENPERRESPONSETOCOMPLYWITHTHISINFORMATION COLLEC'TION REOUEST:50.0HRS.FORWARDCOMHENTSREGARDING BURDENESTIHATETOTHEINFORMATION ANDRECORDSMANAGEMENT BRANCH(MNBB7714),U.S.NUCLEARREGULATORY COHHISSION, WASHINGTON, DC20555-0001, AHDTOTHEPAPERWORK REDUCTION PROJECT(3150.0104),
| | NRC FORH 366A (5-92) |
| OFFICEOFMANAGEHENT ANDBUDGETWASHINGTON DC20503.FACILITYNAHE(1)R.E.GinnaNuclearPowerPlantDOCKETNUMBER(2)05000244LERNUMBER(6)SEOUENTIAL 95--001--REVISION00PAGE(3)4OF9TEXT(Ifmorespaceisrequired, useadditional copiesofNRCForm366A)(17)The"as-found" setpoints were1.6%(forV-434)and2.3%(forV-435)above2485psig.ThisiscontrarytoGinnaTechnical Specification 3.1.1.3.c, whichstates,"Whenever thereactorisatoraboveanRCStemperature of350degreesF,bothpressurizer codesafetyvalvesshallbeoperablewithaliftsettingof2485psig+/-1%."Aconservative decisionwasmadetoreportthiseventunderthecriteriaof10CFR50.72(b)(2)(iii)(D),basedoninputfromNS&Lthata+/-1%tolerance forsafetyvalveactuation isanassumption forseveraldesignbasisevents.TheNRCwasnotifiedatapproximately 2011ESTonFebruary3.Subsequent evaluations havenotbeenabletoconclusively determine thetimethatthesetpointshiftoccurred, noreveniftheshiftoccurredduringcycle23.Reviewofdesignbasiseventshasconfirmed thatthiscondition doesnotmeetthereporting criteriaof10CFR50.72.INOPERABLE STRUCTURES, COMPONENTS, ORSYSTEMSTHATCONTRIBUTED TOTHEEVENT:NoneD.OTHERSYSTEMSORSECONDARY FUNCTIONS AFFECTED:
| | NRC FORM 366A (5.92).S.NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PROVED BY OHB NO.3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORHATIOH COLLECTION REOUEST: 50.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORHATION AHD RECORDS MANAGEMENT BRANCH (HNBB 7714), U.S.NUCLEAR REGULATORY COMHISSION, WASHINGTON, DC 20555-0001, AHD 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 NUHBER (2)05000244 LER NUMBER (6)SEOUENT I AL 95--001--REVISION 00 PAGE (3)7 OF 9 TEXT (If more space is required, use additional copies of HRC Form 366A)(17)The two limiting design basis events that challenge reactor coolant system (RCS)integrity and rely on the PRZR safety valves to mitigate their consequences are the Locked Rotor transient and the Loss of Load transients These transients were reanalyzed.by Westinghouse on behalf of RG&E.k An assessment was performed considering both the safety consequences and implications of this event with the following results and conclusions: |
| NoneMETHODOFDISCOVERY:
| | ~The setpoints for the PRZR safety valves shifted at some unknown time between March, 1993, and January, 1995.This condition did not create a significant safety hazard for the following reasons: 1.While the valves would have been declared inoperable (had the condition been known)based on the Technical Specification tolerance, the reanalyses of the two limiting transients shows that if the valves had lifted at the"as-found" pressure during a design basis event, they would have performed their design function with acceptable results.Thus, the acceptance criteria of all UFSAR Chapter 15 design basis events would still be satisfied. |
| RG&EQASurveillance ofthetestdataidentified thatthetestresultswereunacceptable. | | 2.The design basis conditions bound the actual conditions that existed during cycle 23.Factors that would have made the limiting events less severe are: (a)Steady-state reactor power was approximately 98%during cycle 23, versus the design condition of 100'%b)The"as-found" setpoints were bounded by the Westinghouse reanalysis assumptions for the setpoint tolerance. |
| Thisinformation
| | ~No event that would have required PRZR safety valve actuation occurred during cycle 23.There were no operational or safety consequences or implications attributed to the shift in lift setpoints, because all the required RCS pressure limitations were met, using the"as-found" setpoints. |
| 'wasforwarded toNES,andNCR95-005wasinitiated. | | Based on the above, it can be concluded that the public's health and safety was assured at all times.HRC FORH 366A (5-92) |
| DuringareviewofthisNCRbetweenNESandSystemEngineering, thiscondition wasevaluated aspotentially reportable.
| | NRC FORH 366A (5-92).S.NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PROVEO BY OHB NO.3150~0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORHATION COLLECTION REQUEST: 50.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE IHFORHATIOH AND RECORDS HANAGEHENT BRANCH (MNBB 7714), U.S.NUCLEAR REGULATORY COHMISSION, WASHINGTON, DC 20555-0001, AHD 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 NUHBER (2)05000244 YEAR 95 LER NUMBER (6)SEQUENTIAL |
| HRCFORH366A(5-92)
| | --001--REVISION M 00 PAGE (3)8 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)V~A.CORRECTIVE ACTION: ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:~The valve seats were lapped and the valves adjusted as necessary to bring the lift settings into conformance with the Technical Specifications tolerance. |
| NRCFORM366A(5-92>U.S.NUCLEARREGULATORY COMMISSION FACILITYNAME(1)DOCKETNUMBER(2)LICENSEEEVENTREPORT(LER)TEXTCONTINUATION PPROVEDBYOHBNO.3150.0104 EXPIRES5/31/95ESTIMATED BURDENPERRESPONSETOCOMPLYWITHTHISINFORMATION COLLECTION REOUEST:50'NRS.FORWARDCOHHENTSREGARDIHG BURDENESTIHATETOTHEINFORMATION ANDRECORDSMANAGEMENT BRANCH(HNBB7714),U.S.NUCLEARREGULATORY COMHISSION, WASHINGTON, DC20555-0001 ANDTOTHEPAPERWORK REDUCTION PROJECT(31400104>,OFFICEOFMANAGEMENT ANDBUDGETWASHINGTON DC20503.PAGE(3)LERNUHBER(6)R.E.GinnaNuclearPowerPlant05000244YEAR9500100SEOUENTIAL REVISION5OF9TEXT(Ifmorespaceisrequired, useadditional copiesofNRCForm366A)(17)F.OPERATORACTION:TheSystemEngineernotifiedtheShiftSupervisor ofthetestresultsthataffectedbothPRZRsafetyvalvesthatwerepreviously installed andconsidered operableduringcycle23,andthattheseresultsdidnotaffectcurrently installed equipment.
| | ~The accident analyses that are affected by a PRZR safety valve with a larger tolerance than required by Technical Specification 3.1.1.3.c have been reanalyzed by Westinghouse. |
| AdecisionwasmadetonotifytheNRCper10CFR50.72(b)(2)(iii)(D).Thisnotification wasmadeatapproximately 2011ESTonFebruary3,1995.Sincethiseventdidnotaffectinstalled plantequipment, nootheroperator, actionswerenecessary.
| | The results are that the functions of the PRZR safety valves were never unacceptable with the"as-found" lift settings.B.ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE: |
| SAFETYSYSTEMRESPONSES:
| | Based on the results of the accident reanalysis, a revision to the Technical Specifications to provide more realistic and achievable lift setting tolerances and acceptance criteria for operability will be pursued on a priority basis, as part of the RG&EiNRC effort to implement the Improved Technical Specifications (ITS).To minimize the chance of a shift in the setpoint due to activities associated with on-site removal, handling, decontamination, packaging, shipping, storing, and reinstallation, the administrative controls for removal, shipping, testing, and reinstallation of these valves will be evaluated and enhanced, as appropriate, to ensure that proper controls are in place for key activities that could inadvertently affect the lift settings.To increase the repeatability of test results, NES will evaluate the adequacy of the test requirements of MET-049, and revise the test specification, as appropriate. |
| NoneIII.CAUSEOFEVENT:IMMEDIATE CAUSE:Theimmediate causeforbothPRZRsafetyvalvesbeingconsidered inoperable wasthatthe"as-found" liftsettingsforthesevalveswereabovethesetpointtolerance ofTechnical Specification 3.1.1.3.c.
| | NRC FORH 366A (5-92) |
| B.INTERMED1ATE CAUSE:Theintermediate causeforthe"as-found" liftsettingsabovethe'etpointtolerance ofTechnical Specifications wasashiftinthesetpoints fromthe"as-left" conditions ofMarch,1993,tothe"as-found"conditions ofJanuary,1995.NRCFORH366A(5-92> | | NRC FORM 366A (5-92).S.NUCLEAR REGULATORY COHHISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PROVED BY OHB NO.3150~0104 EXP I RES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORHATION 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 MANAGEHENT AND BUDGET WASHINGTON DC 20503.FACILITY NAME (1)R.E.Ginna Nuclear Power Plant DOCKET NUMBER (2)05000244 95 001 00 LER NUHBER (6)YEAR SEQUENTIAL REVISION PAGE (3)9 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)VI.ADDITIONAL INFORMATION: |
| NRCFORM366A.(5-92).S.NUCLEARREGULATORY COHHISSION LICENSEEEVENTREPORT(LER)TEXTCONTINUATION PROVEDBYOHBNO.3150.0104 EXPIRES5/31/95ESTIMATED BURDENPERRESPONSETOCOHPLYWITHTHISINFORHATION COLLECTION REQUEST:50.0NRS.FORWARDCOHHENTSREGARDING BURDEHESTIHATETOTHEINFORHATION ANDRECORDSHANAGEHENT BRANCH(HNBB7714),U.S.NUCLEARREGULATORY COHHISSIONg WASHINGTON, DC20555-0001, AHDTOTHEPAPERWORK REDUCTIOH PROJECT(3150-0104),
| | A.FAILED COMPONENTS: |
| OFFICEOFHANAGEHENT ANDBUDGETWASHINGTON DC20503.FACILITYNAHE(1)R.E.GinnaNuclearPowerPlantDOCKETNUHBER(2)05000244YEAR95001--00LERNUHBER(6)SEQUENTIAL REVISIONPAGE(3)6OF9TEXT(Ifmorespaceisrequired, useadditional copiesofNRCForm366A)(17)C.ROOTCAUSE:Theunderlying causeoftheshiftinthesetpoints isattributed toacombination offactors,including attendant variables affecting thelong-term operation ofthevalves,andthesubsequent removal,decontamination,
| | The failed components are Crosby Valve and Gage Co.safety valves, Model HB-BP-86E, serial numbers N69877-00-0006 and N69877-00-0007. |
| : handling, andshippingofthevalvestoanoff-sitefacilityfortesting..
| | 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 identified. |
| TheGinnaTechnical Specification requirements of+/-1-'.maynotbeappropriate withrespecttotheallowances fornormalsetpointshiftsduringoperation, removal,andshipping.
| | C.SPECIAL COMMENTS: None NRC FORH 366A (5-92)}} |
| ThiseventisNUREG-1022 CauseCode(B),"DesignManufacturing, Construction
| |
| '/Installation." | |
| IV.ANALYSISOFEVENT:Thiseventisreportable inaccordance with10CFR50.73,LicenseeEventReportSystem,item(a)(2)(vii)(D),whichrequiresareportof,"anyeventwhereasinglecauseorcondition causedatleastoneindependent trainorchanneltobecomeinoperable inmultiplesystemsortwoindependent trainsorchannelstobecomeinoperable inasinglesystemdesignedto...mitigatetheconsequences ofanaccident." | |
| Bothindependent trainsofpressurereliefforthePRZRwereconsidered inoperable duetothe"as-found" liftsettingsabovethetolerance oftheTechnical Specifications.
| |
| NRCFORH366A(5-92)
| |
| NRCFORM366A(5.92).S.NUCLEARREGULATORY COMMISSION LICENSEEEVENTREPORT(LER)TEXTCONTINUATION PROVEDBYOHBNO.3150-0104 EXPIRES5/31/95ESTIMATED BURDENPERRESPONSETOCOHPLYWITHTHISINFORHATIOH COLLECTION REOUEST:50.0HRS.FORWARDCOMMENTSREGARDING BURDENESTIMATETOTHEINFORHATION AHDRECORDSMANAGEMENT BRANCH(HNBB7714),U.S.NUCLEARREGULATORY COMHISSION, WASHINGTON, DC20555-0001, AHDTOTHEPAPERWORK REDUCTION PROJECT(31500104),OFFICEOFMANAGEMENT ANDBUDGETWASHINGTON DC20503.FACILITYNAME(1)R.E.GinnaNuclearPowerPlantDOCKETNUHBER(2)05000244LERNUMBER(6)SEOUENTIAL95--001--REVISION00PAGE(3)7OF9TEXT(Ifmorespaceisrequired, useadditional copiesofHRCForm366A)(17)Thetwolimitingdesignbasiseventsthatchallenge reactorcoolantsystem(RCS)integrity andrelyonthePRZRsafetyvalvestomitigatetheirconsequences aretheLockedRotortransient andtheLossofLoadtransients Thesetransients werereanalyzed
| |
| .byWestinghouse onbehalfofRG&E.kAnassessment wasperformed considering boththesafetyconsequences andimplications ofthiseventwiththefollowing resultsandconclusions: | |
| ~Thesetpoints forthePRZRsafetyvalvesshiftedatsomeunknowntimebetweenMarch,1993,andJanuary,1995.Thiscondition didnotcreateasignificant safetyhazardforthefollowing reasons:1.Whilethevalveswouldhavebeendeclaredinoperable (hadthecondition beenknown)basedontheTechnical Specification tolerance, thereanalyses ofthetwolimitingtransients showsthatifthevalveshadliftedatthe"as-found" pressureduringadesignbasisevent,theywouldhaveperformed theirdesignfunctionwithacceptable results.Thus,theacceptance criteriaofallUFSARChapter15designbasiseventswouldstillbesatisfied. | |
| 2.Thedesignbasisconditions boundtheactualconditions thatexistedduringcycle23.Factorsthatwouldhavemadethelimitingeventslesssevereare:(a)Steady-state reactorpowerwasapproximately 98%duringcycle23,versusthedesigncondition of100'%b)The"as-found" setpoints wereboundedbytheWestinghouse reanalysis assumptions forthesetpointtolerance. | |
| ~NoeventthatwouldhaverequiredPRZRsafetyvalveactuation occurredduringcycle23.Therewerenooperational orsafetyconsequences orimplications attributed totheshiftinliftsetpoints, becausealltherequiredRCSpressurelimitations weremet,usingthe"as-found" setpoints. | |
| Basedontheabove,itcanbeconcluded thatthepublic'shealthandsafetywasassuredatalltimes.HRCFORH366A(5-92)
| |
| NRCFORH366A(5-92).S.NUCLEARREGULATORY COMMISSION LICENSEEEVENTREPORT(LER)TEXTCONTINUATION PROVEOBYOHBNO.3150~0104EXPIRES5/31/95ESTIMATED BURDENPERRESPONSETOCOMPLYWITHTHISINFORHATION COLLECTION REQUEST:50.0HRS.FORWARDCOMMENTSREGARDING BURDENESTIMATETOTHEIHFORHATIOH ANDRECORDSHANAGEHENT BRANCH(MNBB7714),U.S.NUCLEARREGULATORY COHMISSION, WASHINGTON, DC20555-0001, AHDTOTHEPAPERWORK REDUCTION PROJECT(3150-0104),
| |
| OFFICEOFMANAGEMENT ANDBUDGETWASHINGTON DC20503.FACILITYNAME(1)R.E.GinnaNuclearPowerPlantDOCKETNUHBER(2)05000244YEAR95LERNUMBER(6)SEQUENTIAL
| |
| --001--REVISIONM00PAGE(3)8OF9TEXT(Ifmorespaceisrequired, useadditional copiesofNRCForm366A)(17)V~A.CORRECTIVE ACTION:ACTIONTAKENTORETURNAFFECTEDSYSTEMSTOPRE-EVENT NORMALSTATUS:~Thevalveseatswerelappedandthevalvesadjustedasnecessary tobringtheliftsettingsintoconformance withtheTechnical Specifications tolerance. | |
| ~TheaccidentanalysesthatareaffectedbyaPRZRsafetyvalvewithalargertolerance thanrequiredbyTechnical Specification 3.1.1.3.c havebeenreanalyzed byWestinghouse. | |
| Theresultsarethatthefunctions ofthePRZRsafetyvalveswereneverunacceptable withthe"as-found" liftsettings.
| |
| B.ACTIONTAKENORPLANNEDTOPREVENTRECURRENCE: | |
| Basedontheresultsoftheaccidentreanalysis, arevisiontotheTechnical Specifications toprovidemorerealistic andachievable liftsettingtolerances andacceptance criteriaforoperability willbepursuedonaprioritybasis,aspartoftheRG&EiNRCefforttoimplement theImprovedTechnical Specifications (ITS).Tominimizethechanceofashiftinthesetpointduetoactivities associated withon-siteremoval,handling, decontamination, packaging,
| |
| : shipping, storing,andreinstallation, theadministrative controlsforremoval,shipping, testing,andreinstallation ofthesevalveswillbeevaluated andenhanced, asappropriate, toensurethatpropercontrolsareinplaceforkeyactivities thatcouldinadvertently affecttheliftsettings.
| |
| Toincreasetherepeatability oftestresults,NESwillevaluatetheadequacyofthetestrequirements ofMET-049,andrevisethetestspecification, asappropriate.
| |
| NRCFORH366A(5-92)
| |
| NRCFORM366A(5-92).S.NUCLEARREGULATORY COHHISSION LICENSEEEVENTREPORT(LER)TEXTCONTINUATION PROVEDBYOHBNO.3150~0104EXPIRES5/31/95ESTIHATED BURDENPERRESPONSETOCOMPLYWITHTHISINFORMATION COLLECTION REQUEST:50.0HRS.FORWARDCOMMENTSREGARDING BURDENESTIMATETOTHEINFORHATION ANDRECORDSMANAGEMENT BRANCH(MNBB7714),U.S.NUCLEARREGULATORY COMMISSION, WASHINGTON, DC20555-0001 ANDTOTHEPAPERWORK REDUCTION PROJECT(31/0-0104),
| |
| OFFICEOFMANAGEHENT ANDBUDGETWASHINGTON DC20503.FACILITYNAME(1)R.E.GinnaNuclearPowerPlantDOCKETNUMBER(2)050002449500100LERNUHBER(6)YEARSEQUENTIAL REVISIONPAGE(3)9OF9TEXT(Ifmorespaceisrequired, useadditional copiesofNRCForm366A)(17)VI.ADDITIONAL INFORMATION:
| |
| A.FAILEDCOMPONENTS: | |
| Thefailedcomponents areCrosbyValveandGageCo.safetyvalves,ModelHB-BP-86E, serialnumbersN69877-00-0006 andN69877-00-0007.
| |
| PREVIOUSLERsONSIMILAREVENTS:AsimilarLEReventhistorical searchwasconducted withthefollowing results:Nodocumentation ofsimilarLEReventswiththesamerootcauseatGinnaNuclearPowerPlantcouldbeidentified.
| |
| C.SPECIALCOMMENTS: | |
| NoneNRCFORH366A(5-92)}}
| |
LER 95-001-00:on 950203,pressurizer Safety Valves Lift Settings Found Above TS Tolerance During post-svc Test,Due to Setpoint Shifts That Resulted in Independent Trains Being Considered Inoperable| ML17263A983 |
| Person / Time |
|---|
| Site: |
Ginna  |
|---|
| Issue date: |
03/06/1995 |
|---|
| From: |
ST MARTIN J T
ROCHESTER GAS & ELECTRIC CORP. |
|---|
| To: |
|
|---|
| Shared Package |
|---|
| ML17263A982 |
List: |
|---|
| References |
|---|
| LER-95-001, LER-95-1, NUDOCS 9503160241 |
| Download: ML17263A983 (9) |
|
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 FORH 366 (5-92).S.NUCLEAR REGULATORY COHHISSION PPROVED BY OHB NO.3150-0104 EXPIRES 5/31/95 LICENSEE EVENT REPORT (LER)(See reverse for required number of digits/characters for each block)ESTIMATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.FORWARD COHMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (HNBB 7714), U.ST NUCLEAR REGULATORY COHHISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF HANAGEHENT AND BUDGET WASHINGTON DC 20503.FACILITY NAME (1)R.E.Ginna Nuclear Power Plant DOCKET NUMBER (2)05000244 PAGE (3)10F9 TITLE (4)Pressurizer Safety Valves Lift Settings Found Above Technical Specifications Tolerance During Post-service Test, Due to Setpoint Shifts, Results in Independent Trains Being Considered Inoperable HONTH 02 DAY YEAR 03 95 EVENT DATE (5)YEAR 95 LER NUHBER (6)SEQUENTIAL NUHBER--001--REVISION NUHBER 00 MONTH DAY 03 06 YEAR REPORT DATE (7)FACILITY NAHE DOCKET NUMBER FACILITY NAME DOCKET NUMBER OTHER FACILITIES INVOLVED (8)OPERATING HODE (9)POWER LEVEL (10)N 098 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) 50'6(c)(2)50.73(a)(2)(i) 50.73(a)(2)(ii) 50.73(a)(2)(iii) 50.73(a)(2)(iv) 50.73(a)(2)(v)
X 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 below and in Text, NRC Form 366A)TO THE REQUIREMENTS OF 10 CFR 5: (Check one or more)(11)LICENSEE CONTACT FOR THIS LER (12)NAME John T.St.Hartin-Technical Assistant TELEPHONE NUMBER (Include Area Code)(315)524-4446 COHPLETE ONE LINE FOR EACH COHPONENT FAILURE DESCRIBED IN THIS REPORT (13)CAUSE SYSTEH COMPONENT RV HANUFACTURER C170 REPORTABLE TO NPRDS CAUSE SYSTEH COMPONENT MANUFACTURER REPORTABLE TO NPRDS SUPPLEHENTAL REPORT EXPECTED (14)YES (If yes, complete EXPECTED SUBHISSION DATE).X NO EXPECTED SUB HISS ION DATE (15)MONTH DAY YEAR ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)(16)On February 3, 1995, at approximately 1824 EST, with the reactor at approximately 98'.steady state power, both pressurizer safety valves, which had been previously installed and then removed for testing, were considered inoperable.
Recent test results discovered that the"as-found" set pressure for the lift settings had shifted above the tolerance in the Technical Specifications.
Immediate corrective.action was not required, since the valves were not installed.
The underlying cause of the setpoint shift has been attributed to a combination of factors, including long-term operation, removal and shipping to an off-site facility for testing, as well as a restrictive tolerance in the Technical Specifications.
This event is NUREG-1022 Cause Code (B).Corrective action to preclude repetition is outlined in Section V.B.9503160241 950306 PDR ADOCK 05000244 8 PDR NRC FORH 366 (5-92)
NRC FORM 366A (5-92).S.NUCLEAR REGULATORY COMHISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PPROVED BY OHB NO.3150.0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.FORWARD COHMENTS REGARDING BURDEN ESTIHATE TO THE INFORHATION AHD RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF HANAGEHEHT AND BUDGET WASHINGTON DC 20503.FACILITY NAHE (1)R.E.Ginna Nuclear Power Plant DOCKET NUHBER (2)05000244 YEAR 95 LER NUHBER (6)SEQUENTIAL 001 REVISION UM 00 PAGE (3)2 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)I.PRE-EVENT PLANT CONDITIONS:
The plant was at approximately 98'.steady state reactor power with no major activities in progress.Two new pressurizer (PRZR)code safety valves had been purchased, and were tested in the spring of 1993 at the valve manufacturer's test facilities.
The valves were shipped to Rochester Gas and Electric (RGEE)with an"as-left" set pressure of 2485 psig (+/-1'.The original safety valves at Ginna Station were removed during the 1993 outage for annual testing, and these two new safety valves were installed.
These valves (V-434 and V-435)were then considered operable during the 1993/1994 operating cycle (cycle 23)~These two valves were then removed for annual lift testing during the 1994 outage, and the original pair of safety valves (which had been tested in 1994)were installed for the 1994/1995 operating cycle (cycle 24).The removed valves were shipped to a test facility in Huntsville, Alabama, for testing, as per RGEE purchase order NQ-14349-C-JW.
The valves were tested to the requirements of RGGE Test Specification MET-049,"Pressurizer Safety Relief Valve Setpoint Testing", with steam as the test medium, on January 10, 1995 (for V-434)and January 11, 1995 (for V-435).RGEE Quality Assurance (QA)witnessed the tests.The test results showed that the"as-found" setpoints were 2525 psig (for V-434)and 2543 psig (for V-435), which exceeded the 1.lift setting tolerance of Technical Specifications.
These results were recognized as nonconforming, and a Nonconformance Report (NCR 95-005)was initiated to document this condition.
On February 3, 1995, during review of NCR 95-005 by System Engineering and Nuclear Engineering Services (NES), it was determined that this represented a potentially reportable condition.
NRC FORM 366A (5-92)
NRC FORM 366A (5-92).S.NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT'(LER)
TEXT CONTINUATION PPROVED BY OHB NO.3150~0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPOHSE TO.COMPLY WITH THIS IHFORHATION COLLECTIOH REQUEST: 50.0 HRS.FORWARD COHHENTS REGARDIHG BURDEN ESTIHATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MHBB 7714), UPS.NUCLEAR REGULATORY COHHISSION,'WASHINGTON, DC 20555-0001 AHD-TO THE PAPERWORK REDUCTION PROJECT (31i0-0104), OFFICE OF HAHAGEMENT AHD BUDGET WASHINGTON DC 20503.FACILITY NAME (1)R.E.Ginna Nuclear Power Plant DOCKET NUHBER (2)05000244 YEAR 95 001 00 LER NUHBER (6)SEQUENTIAL REVISION PAGE (3)3 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)A.DESCRIPTION OF EVENT: DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES:
1.March, 1993: Newly procured PRZR safety valves are satisfactorily tested at manufacturer's test facility.2.April, 1993: Newly procured PRZR safety valves are installed for the 1993/1994 operating cycle (cycle 23).3.January 11, 1995: Testing of safety valves completed at off-site testing facility.Test results show that the lift pressure exceeded the lift setting tolerance.
Event date and time.4.February 3, 1995, 1824 EST: Test results are reviewed with the System Engineer.Discovery date and time.5.February 3, 1995, 2011 EST: Shift Supervisor notifies NRC per 10 CFR 50.72.B.EVENT: On February 3, 1995, at approximately 1824 EST, the reactor was at approximately 98'.steady state reactor power, and no major activities were in progress.NES personnel, from Mechanical Engineering and Nuclear Safety and Licensing (NSEL), were reviewing the status of NCR 95-005 with the System Engineer.Review of the NCR suggested'an operability question involving these previously, installed safety valves.Since both valves were previously installed during cycle'23, it was conservatively assumed that the valves had shifted out of tolerance during cycle 23.HRC FORH 366A (5-92)
NRC FORH 366A (5-92).S.NUCLEAR REGULATORY COHHISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PROVED BY OHB NO.3150~0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLEC'TION REOUEST: 50.0 HRS.FORWARD COMHENTS REGARDING BURDEN ESTIHATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S.NUCLEAR REGULATORY COHHISSION, WASHINGTON, DC 20555-0001, AHD TO THE PAPERWORK REDUCTION PROJECT (3150.0104), OFFICE OF MANAGEHENT AND BUDGET WASHINGTON DC 20503.FACILITY NAHE (1)R.E.Ginna Nuclear Power Plant DOCKET NUMBER (2)05000244 LER NUMBER (6)SEOUENTIAL 95--001--REVISION 00 PAGE (3)4 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)The"as-found" setpoints were 1.6%(for V-434)and 2.3%(for V-435)above 2485 psig.This is contrary to Ginna Technical Specification 3.1.1.3.c, which states,"Whenever the reactor is at or above an RCS temperature of 350 degrees F, both pressurizer code safety valves shall be operable with a lift setting of 2485 psig+/-1%." A conservative decision was made to report this event under the criteria of 10 CFR 50.72 (b)(2)(iii)(D), based on input from NS&L that a+/-1%tolerance for safety valve actuation is an assumption for several design basis events.The NRC was notified at approximately 2011 EST on February 3.Subsequent evaluations have not been able to conclusively determine the time that the setpoint shift occurred, nor even if the shift occurred during cycle 23.Review of design basis events has confirmed that this condition does not meet the reporting criteria of 10 CFR 50.72.INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT: None D.OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED: None METHOD OF DISCOVERY:
RG&E QA Surveillance of the test data identified that the test results were unacceptable.
This information
'was forwarded to NES, and NCR 95-005 was initiated.
During a review of this NCR between NES and System Engineering, this condition was evaluated as potentially reportable.
HRC FORH 366A (5-92)
NRC FORM 366A (5-92>U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME (1)DOCKET NUMBER (2)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PPROVED BY OHB NO.3150.0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REOUEST: 50'NRS.FORWARD COHHENTS REGARDIHG BURDEN ESTIHATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (HNBB 7714), U.S.NUCLEAR REGULATORY COMHISSION, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140 0104>, OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.PAGE (3)LER NUHBER (6)R.E.Ginna Nuclear Power Plant 05000244 YEAR 95 001 00 SEOUENTIAL REVISION 5 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)F.OPERATOR ACTION: The System Engineer notified the Shift Supervisor of the test results that affected both PRZR safety valves that were previously installed and considered operable during cycle 23, and that these results did not affect currently installed equipment.
A decision was made to notify the NRC per 10 CFR 50.72 (b)(2)(iii)(D).This notification was made at approximately 2011 EST on February 3, 1995.Since this event did not affect installed plant equipment, no other operator, actions were necessary.
SAFETY SYSTEM RESPONSES:
None III.CAUSE OF EVENT: IMMEDIATE CAUSE: The immediate cause for both PRZR safety valves being considered inoperable was that the"as-found" lift settings for these valves were above the setpoint tolerance of Technical Specification 3.1.1.3.c.
B.INTERMED1ATE CAUSE: The intermediate cause for the"as-found" lift settings above the'etpoint tolerance of Technical Specifications was a shift in the setpoints from the"as-left" conditions of March, 1993, to the"as-found" conditions of January, 1995.NRC FORH 366A (5-92>
NRC FORM 366A.(5-92).S.NUCLEAR REGULATORY COHHISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PROVED BY OHB NO.3150.0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORHATION COLLECTION REQUEST: 50.0 NRS.FORWARD COHHENTS REGARDING BURDEH ESTIHATE TO THE INFORHATION AND RECORDS HANAGEHENT BRANCH (HNBB 7714), U.S.NUCLEAR REGULATORY COHHISSIONg WASHINGTON, DC 20555-0001, AHD TO THE PAPERWORK REDUCTIOH PROJECT (3150-0104), OFFICE OF HANAGEHENT AND BUDGET WASHINGTON DC 20503.FACILITY NAHE (1)R.E.Ginna Nuclear Power Plant DOCKET NUHBER (2)05000244 YEAR 95 001--00 LER NUHBER (6)SEQUENTIAL REVISION PAGE (3)6 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)C.ROOT CAUSE: The underlying cause of the shift in the setpoints is attributed to a combination of factors, including attendant variables affecting the long-term operation of the valves, and the subsequent removal, decontamination, handling, and shipping of the valves to an off-site facility for testing..The Ginna Technical Specification requirements of+/-1-'.may not be appropriate with respect to the allowances for normal setpoint shifts during operation, removal, and shipping.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)(vii)(D), which requires a report of,"any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to...mitigate the consequences of an accident." Both independent trains of pressure relief for the PRZR were considered inoperable due to the"as-found" lift settings above the tolerance of the Technical Specifications.
NRC FORH 366A (5-92)
NRC FORM 366A (5.92).S.NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PROVED BY OHB NO.3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORHATIOH COLLECTION REOUEST: 50.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORHATION AHD RECORDS MANAGEMENT BRANCH (HNBB 7714), U.S.NUCLEAR REGULATORY COMHISSION, WASHINGTON, DC 20555-0001, AHD 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 NUHBER (2)05000244 LER NUMBER (6)SEOUENT I AL 95--001--REVISION 00 PAGE (3)7 OF 9 TEXT (If more space is required, use additional copies of HRC Form 366A)(17)The two limiting design basis events that challenge reactor coolant system (RCS)integrity and rely on the PRZR safety valves to mitigate their consequences are the Locked Rotor transient and the Loss of Load transients These transients were reanalyzed.by Westinghouse on behalf of RG&E.k An assessment was performed considering both the safety consequences and implications of this event with the following results and conclusions:
~The setpoints for the PRZR safety valves shifted at some unknown time between March, 1993, and January, 1995.This condition did not create a significant safety hazard for the following reasons: 1.While the valves would have been declared inoperable (had the condition been known)based on the Technical Specification tolerance, the reanalyses of the two limiting transients shows that if the valves had lifted at the"as-found" pressure during a design basis event, they would have performed their design function with acceptable results.Thus, the acceptance criteria of all UFSAR Chapter 15 design basis events would still be satisfied.
2.The design basis conditions bound the actual conditions that existed during cycle 23.Factors that would have made the limiting events less severe are: (a)Steady-state reactor power was approximately 98%during cycle 23, versus the design condition of 100'%b)The"as-found" setpoints were bounded by the Westinghouse reanalysis assumptions for the setpoint tolerance.
~No event that would have required PRZR safety valve actuation occurred during cycle 23.There were no operational or safety consequences or implications attributed to the shift in lift setpoints, because all the required RCS pressure limitations were met, using the"as-found" setpoints.
Based on the above, it can be concluded that the public's health and safety was assured at all times.HRC FORH 366A (5-92)
NRC FORH 366A (5-92).S.NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PROVEO BY OHB NO.3150~0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORHATION COLLECTION REQUEST: 50.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE IHFORHATIOH AND RECORDS HANAGEHENT BRANCH (MNBB 7714), U.S.NUCLEAR REGULATORY COHMISSION, WASHINGTON, DC 20555-0001, AHD 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 NUHBER (2)05000244 YEAR 95 LER NUMBER (6)SEQUENTIAL
--001--REVISION M 00 PAGE (3)8 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)V~A.CORRECTIVE ACTION: ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:~The valve seats were lapped and the valves adjusted as necessary to bring the lift settings into conformance with the Technical Specifications tolerance.
~The accident analyses that are affected by a PRZR safety valve with a larger tolerance than required by Technical Specification 3.1.1.3.c have been reanalyzed by Westinghouse.
The results are that the functions of the PRZR safety valves were never unacceptable with the"as-found" lift settings.B.ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:
Based on the results of the accident reanalysis, a revision to the Technical Specifications to provide more realistic and achievable lift setting tolerances and acceptance criteria for operability will be pursued on a priority basis, as part of the RG&EiNRC effort to implement the Improved Technical Specifications (ITS).To minimize the chance of a shift in the setpoint due to activities associated with on-site removal, handling, decontamination, packaging, shipping, storing, and reinstallation, the administrative controls for removal, shipping, testing, and reinstallation of these valves will be evaluated and enhanced, as appropriate, to ensure that proper controls are in place for key activities that could inadvertently affect the lift settings.To increase the repeatability of test results, NES will evaluate the adequacy of the test requirements of MET-049, and revise the test specification, as appropriate.
NRC FORH 366A (5-92)
NRC FORM 366A (5-92).S.NUCLEAR REGULATORY COHHISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION PROVED BY OHB NO.3150~0104 EXP I RES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORHATION 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 MANAGEHENT AND BUDGET WASHINGTON DC 20503.FACILITY NAME (1)R.E.Ginna Nuclear Power Plant DOCKET NUMBER (2)05000244 95 001 00 LER NUHBER (6)YEAR SEQUENTIAL REVISION PAGE (3)9 OF 9 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)VI.ADDITIONAL INFORMATION:
A.FAILED COMPONENTS:
The failed components are Crosby Valve and Gage Co.safety valves, Model HB-BP-86E, serial numbers N69877-00-0006 and N69877-00-0007.
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 identified.
C.SPECIAL COMMENTS: None NRC FORH 366A (5-92)