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| | number = ML17037C433 | | | number = ML17037C433 |
| | issue date = 08/08/1977 | | | issue date = 08/08/1977 |
| | title = Nine Mile Point, Unit 1 - Letter Regarding Surveillance Requirements and Limiting Conditions for Operations and Requesting an Application for Amendment to the Operating License That Will Change the Technical Specifications | | | title = Letter Regarding Surveillance Requirements and Limiting Conditions for Operations and Requesting an Application for Amendment to the Operating License That Will Change the Technical Specifications |
| | author name = Lear G | | | author name = Lear G |
| | author affiliation = NRC/NRR | | | author affiliation = NRC/NRR |
| | addressee name = Rhode G K | | | addressee name = Rhode G |
| | addressee affiliation = Niagara Mohawk Power Corp | | | addressee affiliation = Niagara Mohawk Power Corp |
| | docket = 05000220 | | | docket = 05000220 |
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| | document type = Letter, Request for Additional Information (RAI), Technical Specification, Amendment | | | document type = Letter, Request for Additional Information (RAI), Technical Specification, Amendment |
| | page count = 20 | | | page count = 20 |
| | | project = |
| | | stage = Other |
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| =Text= | | =Text= |
| {{#Wiki_filter:=AUG81977DocketNo.50-220NiagaramohawkPowerCorporationATTN:Nr.GeraldK.RhodeVicePresident-Engineering300ErieBoulevardWestSyracuse,NewYork13202Gentlemen:RE:NINENILEPOINTNUCLEARSTATIOflUNITNO.1DistributionDocketORB83CParrishGLearSNowickiDVerrelliLocalPDRNRCPDRAttorney,OELDOI&E(3)DEisenhutTBAbernathyNRBuchananACRS(16)Inthepastseveralyears,asignificantnumberofreliefvalvesandsafety-reliefvalveswerefoundtobeinoperableatBWRreactorfacili-ties.ThesevalveswereinstalledintheReactorCoolantSystemand/orAutomaticDepressurizationSystem.Severalprogramshavebeendevelopedtoreducetheincidenceofthesevalvefailures;however,additionalfailurescontinuetooccur.Consequently,wehaveconcludedthatchangestotheSurveillanceRequire-mentsandLimitingConditionsforOperationsforallBHR'sareneededtoprovideadditionalassuranceofreliefvalveandsafety-reliefvalveoperabflityandreliability.Therefore,werequestthatyoumodifyyoursurveillancetestingprogramthroughtheadoptionoftheprogramcontainedinthemodeltechnicalspecificationswehaveprepared.Theelementsofthisprograminclude:Eachremotlyoperatedreliefvalveandsafety-reliefvalvefntheReactorCoolantSystemandAutomaticDepressurizatfonSystemwillbetestedonavariablefrequencyschedulerelatedtodemonstratedreliabilityandoperability.Thetestingintervalisbasedonthenumberofvalvefailuresduringtherequiredtestinterval.Facilitieswithreliablevalveswillprogresstoalongertestintervalwhilethosewithvalvefailureswillprogresstoashortertestinterval.Thisconceptshouldresultinthemaintenanceofamoreuniformlevelofreliabilityforthisequipmentthanpreviouslyobtained.2.TheincreasedsurveillanceprogramwillbecomeeffectiveonHarch1,19T9.Noincreaseinvalve,testingisrequiredbeforethatdate.Theinitialtestinginterval.oftheincreasedsurveillanceprogramwillbebasedonthenumberofremotelyoperatedreliefvalvesandsafety-reliefvalvesfoundinoperableintheprevious18monthspP'September1,1977toMarch1,1979).ThisleadtimewillpermittheresolutionoftheHarkISafety-ReliefValveLoadsandDFPICE~SUANAMEWDATE~NRCFORM'318(9.76)NRCM02404V8OOVEANMENTPAINTIN4OFFICER1024420624 4Fhhr1*NI(I/'('I~fI)'''CItAIhrortSI*.~rl1lfESIih"11('",fhh"Ih(7.'''),rh,firIh,INN,h(It~Nttr r~pNiagaraHohawkPowerCorporationp,UG8197)timeissufficienttopermitthedevelopmentandimplementationofimprovedsafetyandsafety-relfefvalvemaintenanceproceduresand'othercorrectiveactionspriortoimplementingthetestprogram.3.Thereliefand/orsafety-reliefvalvelinerestraintsinthetoruswillbeexaminedpriortoinitiatingthetestprogramandatleastonceeachfuelcycle(f.e.,each18months)toverifycontinuedstructuralintegrity.Werequestthatyousubmitwithin30daysfromyourreceiptofthis'letter,anapplicationforamendmenttoyourlicensethatwillchangeyourtechnicalspecificationstobeinconformancewfththerequirementsoftheenclosedmodeltechnicalspecificationsandassociated'bases.Intheeventyoushoulddesirefurtherdfscussionofthismatter,pleasecontactus.Sincerely,OriginalsignedbV | | {{#Wiki_filter:= |
| | Distribution AUG 8 1977 Docket ORB 83 CParrish Docket No. 50-220 GLear SNowicki DVerrelli Local PDR Niagara mohawk Power Corporation NRC PDR ATTN: Nr. Gerald K. Rhode Attorney, OELD Vice President - Engineering OI&E (3) 300 Erie Boulevard West DEisenhut Syracuse, New York 13202 TBAbernathy NRBuchanan Gentlemen: ACRS (16) |
| | RE: NINE NILE POINT NUCLEAR STATIOfl UNIT NO. 1 In the past several years, a significant number of relief valves and safety-relief valves were found to be inoperable at BWR reactor facili-ties. These valves were installed in the Reactor Coolant System and/or Automatic Depressurization System. Several programs have been developed to reduce the incidence of these valve failures; however, additional failures continue to occur. |
| | Consequently, we have concluded that changes to the Surveillance Require-ments and Limiting Conditions for Operations for all BHR's are needed to provide additional assurance of relief valve and safety-relief valve operabflity and reliability. Therefore, we request that you modify your surveillance testing program through the adoption of the program contained in the model technical specifications we have prepared. The elements of this program include: |
| | Each remot ly operated relief valve and safety-relief valve fn the Reactor Coolant System and Automatic Depressurizatfon System will be tested on a variable frequency schedule related to demonstrated reliability and operability. The testing interval is based on the number of valve failures during the required test interval. Facilities with reliable valves will progress to a longer test interval while those with valve failures will progress to a shorter test interval. This concept should result in the maintenance of a more uniform level of reliability for this equipment than previously obtained. |
| | : 2. The increased surveillance program will become effective on Harch 1, 19T9. No increase in valve, testing is required before that date. |
| | The initial testing interval. of the increased surveillance program will be based on the number of remotely operated relief valves and safety-relief valves found inoperable in the previous 18 months 1, 1977 to March 1, 1979). This lead time will permit pP'September the resolution of the Hark I Safety-Relief Valve Loads and DFPICE~ |
| | SUANAMEW DATE~ |
| | NRC FORM'318 (9.76) NRCM 0240 4V 8 OOVEANMENT PAINTIN4 OFFICER 1024 420 624 |
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| | r~ p 197) p,UG 8 Niagara Hohawk Power Corporation time is sufficient to permit the development and implementation of improved safety and safety-relfef valve maintenance procedures and |
| | 'other corrective actions prior to implementing the test program. |
| | : 3. The relief and/or safety-relief valve line restraints in the torus will be examined prior to initiating the test program and at least once each fuel cycle (f.e., each 18 months) to verify continued structural integrity. |
| | We request that you submit within 30 days from your receipt of this 'letter, an application for amendment to your license that will change your technical specifications to be in conformance wfth the requirements of the enclosed model technical specifications and associated'bases. In the event you should desire further dfscussion of this matter, please contact us. |
| | Sincerely, Original signed bV George Lear, Chief Operating Reactors Branch 83 Division of Operating Reactors |
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| ==Enclosure:== | | ==Enclosure:== |
| ModelTechnicalSpecificationsGeorgeLear,ChiefOperatingReactorsBranch83DivisionofOperatingReactorsCC:ArvinE.Upton,EsquireLeBoeuf,Lamb,Lefby5NacRae1757NStreet,H.W.Washington,D.C.20036AnthonyZ.Rofsman,EsquireRofsman,KesslerandCashdan102515thStreet,N.H.5thFloorWashington,D.C.20005Nr.EugeneG.Saloga,ApplicantCoordinator.,NineNilePointEnergyInformationCenterP.0.Box81Lycoming,Nes(York13093OFFICS~SU/NAME&DATE~ORB3CParrish8/1/77ORBk3SNowickjf8/)/77ORBii3jDV~rlI8//77,ORB83GLearRS(S77'RCFORM318(9-76)NRCM0240AU,'0,OOVSANMSNTFIIINTINOOFFICS<ISTS02~24
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| <<FIF~r'F<I"I1I1<~I~k~rlrF<k<<<l<<'Ih<IIYll<<'tj'l~f'1.,Fkrl<<'.~W REACTORCOOLANTSYSTEM3/4.4.2SAFETYVALVESLIMITINGCONDITIONFOROPERATION3.4.2Atleastthefollowingreactorcoolantsystemcodesafetyvalvesandsafety-reliefvalvesshallbeoperablewithliftsettingswithin+1%oftheindicatedpressures.(2)*Safetyvalves9(1240)psig(3)Safety-reliefvalves9(1100)psig(3)Safety-reliefvalves9(1090)psig(3)Safety-reliefvalves9(1080)psigAPPLICABILITY:WithAverageCoolantTemperature>212'FortheModeSwitchinRun,orStartup/HotStandby.ACTION:Withoneormorereactorcoolantsystemcodesafetyvalve(s)orasafety-reliefvalve(s)inoperableeitherrestorethevalve(s)tooperablestatuswithin15minutesorbe.shutdownwithin12hoursandreduceAverageCoolantTemperatureto<212'Fwithinthenext24hours..SURVEILLANCEREUIREMENTS4.4.2.1a.InadditiontotheapplicableASMEBoilerandPressureVesselCode,SectionXIrequirements,eachsafety-reliefvalveshallbedemonstratedoperable:Atleastonceper24hours,byverifyingbellowsintegritythroughinstrumentindication.b.UntilMarch1,1979,atleastonceper18monthsby:1.Manuallyopeningeachremotelyoperated.safety-reliefvalvewiththereactoratorbelow5%ratedpowerandatnominaloperatingpressure,andverifyingthateither:a.Theturbinebypassvalve(s)indicateacompensatingvalvemovement,orb.Thereactorcoolantsystempressuredecreasesbyanamountequivalenttothevalvepressurerelievingcapacityforthetestconditions.ddf
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| REACTORCOOLANTSYSTEMSURVEILLANCEREUIREMENTS(Continued)2.Conductingavisualinspectionofthesafety-reliefvalvelinerestraintsinthetorustoverifystructuralintegrityforcontinuedoperation.c.AfterMarch1,1979,byperformanceofthefollowingtestprogram:1.Manuallyopeningeachremotelyoperatedsafety-reliefvalveinaccordancewiththetestscheduleofTable4.4-10withthereactoratorbelow5Xratedpowerandasteamatnominaloperatingpressureandverifyingthateither:a.Theturbinebypassvalve(s)indicateacompensatingvalvemovement,orb.Thereactorcoolantsystempressuredecreasesbyanamountequivalenttothevalvepressurerelievingcapacityforthetestconditions.2.TheinitialNextRequiredTestIntervalofTable4.4-10shallbedeterminedbythenumberofremotelyoperatedreliefandsafety-reliefvalvesfoundinoperablefromSeptember1,1977toMarch1,1979.3.TheinitialvalvetestsofTable4.4-10shallbecompletedby,theearlierof:a.ThecompletionofthenextrefuelingoutageoccurringafterMarch1,1979,orb.ThetimeperioddefinedbyMarch1,1979plustheinitialtestinterval,determinedabove.4.Atleastonceper18months,byconductingavisualinspectionofthesafety-reliefvalvelinerestraintsinthetorustoverifystructuralintegrityforcontinuedoperation.4.4.2.2Eachsafetyvalveandthesafetyvalvefunctionofeachsafety-reliefvalveshallbedemonstratedoperablepertherequirementsoftheASMEBoilerandPressureVesselCode()EditionandAddendathrough().V
| | Model Technical Specifications CC: |
| | Arvin E. Upton, Esquire LeBoeuf, Lamb, Lefby 5 NacRae 1757 N Street, H. W. |
| | Washington, D. C. 20036 Anthony Z. Rofsman, Esquire Rofsman, Kessler and Cashdan 1025 15th Street, N. H. |
| | 5th Floor Washington, D. C. 20005 Nr. Eugene G. Saloga, Applicant Coordinator |
| | ., Nine Nile Point Energy Information Center P. 0. Box 81 Lycoming, Nes( York 13093 ORB 3 ORB k3 ORB ii3 j ORB 83 OFFICS~ |
| | SU/NAME& CParrish SNowick jf DV ~rl I GLear DATE~ |
| | 8/1 /77 8/) /77 8/ /77, RS (S 77 |
| | 'RC FORM 318 (9-76) NRCM 0240 A U,'0, OOVSANMSNT FIIINTINO OFFICS< ISTS 02~24 |
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| TABLE4.4-10REHOTELYOPERATEDRELIEFANDSAFETY-RELIEFVALVETESTSCHEDULENUHBEROFREHOTELYOPERATEDRELIEFANDSAFETY-RELIEFVALVESFOUNDINOPERABLEDURINGTESTINGORTESTINTERVAL**NEXTREQUIREDTESTINTERVAL*012318months+25%184days+25K92days+25%31days+25%"Therequiredtestintervalshallnotbelengthenedmorethanonestepatatime.Earlytestsmaybeperformedpriortoenteringthe"nextrequiredtestinterval"(i.e.,inadvanceofthenominaltimelessthenegative25Ktoleranceband).Earlytestsmaybeusedasanewreferencepointfortestsofthesameinterval,however,theyarenotacceptableforlengtheningthetestinterval.**Setpoint-driftisnotconsideredtobeavalvefailureforthepurposesofthistestschedule.
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| I 3/4.4REACTORCOOLANTSYSTEMBASES3/4.4.2SAFETYVALVESThereactorcoolantsystemsafetyvalvesoperatetopreventthereactorcoolantsystemfrombeingpressurizedabovetheSafetyLimitofpsig.Eachsafetyvalveisdesignedtorelieveibsperhooratthevalvesetpoint.ThesystemisdesignedtomeettheASMEBoilerandPressureVesselCoderequirementsthatthenuclearsystemreliefvalvesshal=lfunctiontopreventopeningofthesafetyvalves.Althoughthesafetyvalvefunctionisnotexpectedtoberequiredunderthemostlimitingtransient,aninoperablevalverequiresshutdowninordertocomplywithASMECoderequirements.Thetestingfrequencyapplicabletothereliefvalvefunctionofthesafety-reliefvalvesisprovidedtoensureoperabilityanddemonstratereliabilityofthevalves.Therequiredtestingintervalvarieswithobservedvalvefailures.Thenumberofinoperablevalvesfoundduringbothoperationandtestingofthesevalvesdeterminesthetimeintervalforthenextrequiredtestofthesevalves.Earlytestsmaybeperformedpriortoenteringthenextrequiredtestinterval(i.e.,inadvanceofthenominaltimelessthenegative25%toleranceband).Earlytestsmaybeusedasanewreferencepointfortestsofthesametimeinterval,however,theyarenotacceptableforlengtheningthetestintervalsincetheywerenotperformedwithinthe+25%tolerancebandasrequiredbyTable4.4-10.Oemonstrationofthesafetyvalves'iftsettingswilloccuronlyduringshutdownandwillbeperformedinaccordancewiththeprovisionsofSectionXIoftheASMEBoilerandPressureVesselCode. | | " 1 ~ I |
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| EMERGENCYCORECOOLINGSYSTEMSAUTOMATICDEPRESSURIZATIONSYSTEMLIMITINGCONDITIONFOROPERATION3.5.2TheAutomaticDepressurizationSystem(ADS)shallbeOPERABLEwithatleast(6)*OPERABLEADSvalves.APPLICABILITY:WithAverageCoolantTemperature>212'FortheModeSwitchinRun,orStartup/HotStandby.ACTION:a.WithoneoftheaboverequiredADSvalvesinoperable,operationmaycontinueprovidedtheactuationlogicoftheremainingADSvalvesisoperableandtheCSSandLPCIsystemsareoperable,andtheHPCIsystemisdemonstratedoperablewithin4hours;restoretheinoperableADSvalvetooperablestatuswit'iin14daysorbeshutdownwithin12hoursandreducetheAverageCoolantTemperatureto<212'Fwithinthefollowing24hours.b.WithtwoormoreoftheaboverequiredADSvalvesinoperable,beshutdownwithin12hoursandreducetheAverageCoolantTemperatureto<212'Fwithinthefollowing24hours.SURVEILLANCEREUIREMENTS4.5.2InadditiontotheapplicableASMEBoilerandPressureVesselCode,SectionXIrequirements,theADSshallbedemonstratedoperable:a.Atleastonceper18monthsbyperformanceofasystemfunctionaltestwhichincludessimulatedautomaticactuationthroughtheautomaticdepressurizationsequence,butexcludingvalveactuation.b.UntilMarch1,1979,atleastonceper18monthsby:1.ManuallyopeningeachADSvalvewiththereactoratorbelow5Xratedpowerandatnominaloperatingpressureandverifyingthateither:a.Theturbinebypassvalve(s)indicateacompensatingvalvemovement,orb.Thereactorcoolantsystempressuredecreasesbyanamountequivalenttothevalvepressurerelievingcapacityforthetestconditions.umeroAvavestobeconsistentwithECCSanalysis.
| | REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.4.2 At least the following reactor coolant system code safety valves and safety-relief valves shall indicated pressures. |
| | be operable with lift settings within + 1% of the (2)* Safety valves 9 (1240) psig (3) Safety-relief valves 9 ( 1100) psig (3 ) Safety-relief valves 9 ( 1090) psig (3) Safety-relief valves 9 (1080) psig APPLICABILITY: With Average Coolant Temperature > 212'F or the Mode Switch in Run, or Startup/Hot Standby. |
| | ACTION: |
| | With one or more reactor coolant system code safety valve(s) or a safety-relief valve(s) inoperable either restore the valve(s) to operable status within 15 minutes or be. shutdown within 12 hours and reduce Average Coolant Temperature to < 212'F withinthe next 24 hours.. |
| | SURVEILLANCE RE UIREMENTS 4.4.2.1 In addition to the applicable ASME Boiler and Pressure Vessel Code, Section XI requirements, each safety-relief valve shall be demonstrated operable: |
| | : a. At least once per 24 hours, by verifying bellows integrity through instrument indication. |
| | : b. Until March 1, 1979, at least once per 18 months by: |
| | : 1. Manually opening each remotely operated .safety-relief valve with the reactor at or below 5% rated power and at nominal operating pressure, and verifying that either: |
| | : a. The turbine bypass valve(s) indicate a compensating valve movement, or |
| | : b. The reactor coolant system pressure decreases by an amount equivalent to the valve pressure relieving capacity for the test conditions. |
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| EMERGENCYCORECOOLINGSYSTEMSAUTOMATICDEPRESSURIZATIONSYSTEMSURVEILLANCEREQUIREMENTS(Continued)2.Conductingavisualinspectionofthesafety-reliefandreliefvalvelinerestraintsinthetorustoverifystructuralintegrityforcontinuedoperation.c.AfterMarch1,1979,byperformanceofthefollowingtestprogram:l.ManuallyopeningeachADSvalveinaccordancewiththetestscheduleofTable4.4-10withthereactoratorbelow5%ratedpowerandatnominaloperatingpressureandverifyingthateither:a.Theturbine".ypassvalve(s)indicateacompensatingvalvemovement,orb.Thereactorcoolantsystempressuredecreasesbyanamountequivalenttothevalvepressurerelievingcapacityforthetestconditions.2.TheinitialNextRequiredTestIntervalofTable4.4-10shallbedeterminedbythenumberofremotelyoperatedreliefandsafety-reliefvalvesfoundinoperablefromSeptember1,1977toMarch1,1979.3.TheinitialvalvetestsofTable4.4-10shallbecompletedby,theearlierof:a.ThecompletionofthenextrefuelingoutageoccurringafterMarch1,1979,orb.ThetimeperioddefinedbyMarch1,1979plustheinitialte'stinterval,determinedabove.4.Atleastonceper18monthsbyconductingavisualinspectionofthesafety-reliefandreliefvalvelinerestraintsinthetorustoverifystructuralintegrityforcontinuedoperation.
| | REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS (Continued) |
| | : 2. Conducting a visual inspection of the safety-relief valve line restraints in the torus to verify structural integrity for continued operation. |
| | : c. After March 1, 1979, by performance of the following test program: |
| | : 1. Manually opening each remotely operated safety-relief valve in accordance with the test schedule of Table 4.4-10 with the reactor at or below 5X rated power and a steam at nominal operating pressure and verifying that either: |
| | : a. The turbine bypass valve(s) indicate a compensating valve movement, or |
| | : b. The reactor coolant system pressure decreases by an amount equivalent to the valve pressure relieving capacity for the test conditions. |
| | : 2. The initial Next Required Test Interval of Table 4.4-10 shall be determined by the number of remotely operated relief and safety-relief valves found inoperable from September 1, 1977 to March 1, 1979. |
| | : 3. The initial valve tests of Table 4.4-10 shall be completed by, the earlier of: |
| | : a. The completion of the next refueling outage occurring after March 1, 1979, or |
| | : b. The time period defined by March 1, 1979 plus the initial test interval, determined above. |
| | : 4. At least once per 18 months, by conducting a visual inspection of the safety-relief valve line restraints in the torus to verify structural integrity for continued operation. |
| | 4.4.2.2 Each safety valve and the safety valve function of each safety-relief valve shall be demonstrated operable per the requirements of the ASME Boiler and Pressure Vessel Code ( ) Edition and Addenda through |
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| | TABLE 4.4-10 REHOTELY OPERATED RELIEF AND SAFETY-RELIEF VALVE TEST SCHEDULE NUHBER OF REHOTELY OPERATED RELIEF AND SAFETY-RELIEF VALVES NEXT REQUIRED FOUND INOPERABLE DURING TESTING OR TEST INTERVAL** TEST INTERVAL* |
| | 0 18 months + 25% |
| | 1 184 days + 25K 2 92 days + 25% |
| | 3 31 days + 25% |
| | "The required test interval shall not be lengthened more than one step at a time. |
| | Early tests may be performed prior to entering the "next required test interval" (i.e., in advance of the nominal time less the negative 25K tolerance band). |
| | Early tests may be used as a new reference point for tests of the same interval, however, they are not acceptable for lengthening the test interval. |
| | **Setpoint- drift is not considered to be a valve failure for the purposes of this test schedule. |
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| | I 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.2 SAFETY VALVES The reactor coolant system safety valves operate to prevent the reactor coolant system from being pressurized above the Safety Limit of psig. Each safety valve is designed to relieve ibs per hoor at the valve set point. The system is designed to meet the ASME Boiler and Pressure Vessel Code requirements that the nuclear system relief valves shal=l function to prevent opening of the safety valves. |
| | Although the safety valve function is not expected to be required under the most limiting transient, an inoperable valve requires shutdown in order to comply with ASME Code requirements. |
| | The testing frequency applicable to the relief valve function of the safety-relief valves is provided to ensure operability and demonstrate reliability of the valves. The required testing interval varies with observed valve failures. The number of inoperable valves found during both operation and testing of these valves determines the time interval for the next required test of these valves. Early tests may be performed prior to entering the next required test interval (i.e., in advance of the nominal time less the negative 25% tolerance band). Early tests may be used as a new reference point for tests of the same time interval, however, they are not acceptable for lengthening the test interval since they were not performed within the +25% tolerance band as required by Table 4.4-10. |
| | Oemonstration of the safety valves'ift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code. |
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| | EMERGENCY CORE COOLING SYSTEMS AUTOMATIC DEPRESSURIZATION SYSTEM LIMITING CONDITION FOR OPERATION 3.5.2 The Automatic Depressurization System (ADS) shall be OPERABLE with at least (6 )* OPERABLE ADS valves. |
| | APPLICABILITY: With Average Coolant Temperature > 212'F or the Mode Switch in Run, or Startup/Hot Standby. |
| | ACTION: |
| | : a. With one of the above required ADS valves inoperable, operation may continue provided the actuation logic of the remaining ADS valves is operable and the CSS and LPCI systems are operable, and the HPCI system is demonstrated operable within 4 hours; restore the inoperable ADS valve to operable status wit'iin 14 days or be shutdown within 12 hours and reduce the Average Coolant Temperature to < 212'F within the following 24 hours. |
| | : b. With two or more of the above required ADS valves inoperable, be shutdown within 12 hours and reduce the Average Coolant Temperature to < 212'F within the following 24 hours. |
| | SURVEILLANCE RE UIREMENTS 4.5.2 In addition to the applicable ASME Boiler and Pressure Vessel Code, Section XI requirements, the ADS shall be demonstrated operable: |
| | : a. At least once per 18 months by performance of a system functional test which includes simulated automatic actuation through the automatic depressurization sequence, but excluding valve actuation. |
| | : b. Until March 1, 1979, at least once per 18 months by: |
| | : 1. Manually opening each ADS valve with the reactor at or below 5X rated power and at nominal operating pressure and verifying that either: |
| | : a. The turbine bypass valve(s) indicate a compensating valve movement, or |
| | : b. The reactor coolant system pressure decreases by an amount equivalent to the valve pressur e relieving capacity for the test conditions. |
| | um er o A va ves to be consistent with ECCS analysis. |
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| | EMERGENCY CORE COOLING SYSTEMS AUTOMATIC DEPRESSUR IZATION SYSTEM SURVEILLANCE REQUIREMENTS (Continued) |
| | : 2. Conducting a visual inspection of the safety-relief and relief valve line restraints in the torus to verify structural integrity for continued operation. |
| | : c. After March 1, 1979, by performance of the following test program: |
| | : l. Manually opening each ADS valve in accordance with the test schedule of Table 4.4-10 with the reactor at or below 5% rated power and at nominal operating pressure and verifying that either: |
| | : a. The turbine ".ypass valve(s) indicate a compensating valve movement, or |
| | : b. The reactor coolant system pressure decreases by an amount equivalent to the valve pressure relieving capacity for the test conditions. |
| | : 2. The initial Next Required Test Interval of Table 4.4-10 shall be determined by the number of remotely operated relief and safety-relief valves found inoperable from September 1, 1977 to March 1, 1979. |
| | : 3. The initial valve tests of Table 4.4-10 shall be completed by, the earlier of: |
| | : a. The completion of the next refueling outage occur ring after March 1, 1979, or |
| | : b. The time period defined by March 1, 1979 plus the initial te'st interval, determined above. |
| | : 4. At least once per 18 months by conducting a visual inspection of the safety-relief and relief valve line restraints in the torus to verify structural integrity for continued operation. |
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| | 3/4.5 EMERGENCY CORE COOLING SYSTEM BASES 3/4.5.2 AUTOMATIC DEPRESSURIZATION SYSTEM (ADS) |
| | Upon failure of the HPCIS to function properly after a small break loss-of-coolant accident, the ADS automatically causes the safety-relief valves to open, depressurizing the reactor so that flow from the low pressure cooling systems can enter the core in time to limit fuel cladding temperature to less than 2200'F. ADS is conservatively required to be operable whenever .reactor vessel pressure exceeds ( 150) psig even though low pressure cooling systems provide adequate core cooling up to (350) psig. |
| | ADS automatically controls (7 ) safety-relief valves although the safety analysis only takes credit for (6 ). Therefore it is appropriate to permit (one) valve to be out-of-service without materially reducing system reliability. |
| | The testing frequency applicable to ADS valves is provided to ensure operability and demonstrate reliability of the valves. The required .test',ng interval varies with observed valve failures. The |
| | ,number of inoperable valves found during both operation and testing of these valves determines the time interval for the next required test of these valves. Early tests may be performed prior to entering the next required test interval (i.e., in advance of the nominal time less the negative 25% tolerance band). Early tests may be used as a new reference point for tests of the same time interval, however, they are not acceptable for lengthening the test interval since they were not performed within the +25% tolerance band as required by Table 4.4-10. |
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| 3/4.5EMERGENCYCORECOOLINGSYSTEMBASES3/4.5.2AUTOMATICDEPRESSURIZATIONSYSTEM(ADS)-UponfailureoftheHPCIStofunctionproperlyafterasmallbreakloss-of-coolantaccident,theADSautomaticallycausesthesafety-reliefvalvestoopen,depressurizingthereactorsothatflowfromthelowpressurecoolingsystemscanenterthecoreintimetolimitfuelcladdingtemperaturetolessthan2200'F.ADSisconservativelyrequiredtobeoperablewhenever.reactorvesselpressureexceeds(150)psigeventhoughlowpressurecoolingsystemsprovideadequatecorecoolingupto(350)psig.ADSautomaticallycontrols(7)safety-reliefvalvesalthoughthesafetyanalysisonlytakescreditfor(6).Thereforeitisappropriatetopermit(one)valvetobeout-of-servicewithoutmateriallyreducingsystemreliability.ThetestingfrequencyapplicabletoADSvalvesisprovidedtoensureoperabilityanddemonstratereliabilityofthevalves.Therequired.test',ngintervalvarieswithobservedvalvefailures.The,numberofinoperablevalvesfoundduringbothoperationandtestingofthesevalvesdeterminesthetimeintervalforthenextrequiredtestofthesevalves.Earlytestsmaybeperformedpriortoenteringthenextrequiredtestinterval(i.e.,inadvanceofthenominaltimelessthenegative25%toleranceband).Earlytestsmaybeusedasanewreferencepointfortestsofthesametimeinterval,however,theyarenotacceptableforlengtheningthetestintervalsincetheywerenotperformedwithinthe+25%tolerancebandasrequiredbyTable4.4-10.
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| 0DocketNo.50-220~4l)'g0ig7NiagaraflohawkPowerCorporationATTN:tfr.GeraldK.RhodeVicePresident-Engineering300ErieBoulevardWestSyracuse,NeviYork13202Gentlemen:RE:NINEMILEPOINTNUCLEARSTATIONUNITNO.1Distribution~~DocketORB¹3LocalPDRNRCPDRGLearCParrishJZwetzigSNowickiAttorney,OELDOI8E(3)DEisenhutTBAbernathyJRBuchananACRS(16)Ourreviewofdatareceivedfromreactorvesselmaterialsurveillanceprogramsindicatesthatthematerialsusedinreactorvesselfabricationmayhaveawidervariationfnsensitivitytoradiationdamagethanoriginallyanticipated.Inaddition,somereactorvesselsincorporatemorethanoneheatofmaterials,includingvleldmetalsfntheirbeltlfneregions,butalloftheseheatsmaynotbeincludedinthereactorvesselmaterialsurveillanceprogram.Althoughourreviewofthesedatadoesnotrevealabasisforconcernregardingcontinuedreactorvesselintegrityoverthenextseveralyears,theinformationdoesindicatetheneedforadetailedreviewofthematerialsemployedinreactorvesselconstruction(fnlightofthisrecentdata)andareviewofthespecimensemployedinthesurveillanceprogramtodetermineffthepresentspecimensreasonablyrepresentthelimitingmaterialsfnthe.reactorvesselbeltlfneregion.Inordertoperformthesereviews,wewillneedtheinformationlistedfntheenclosurerelativetoeachofyourreactorvessel(s)andassocfatedsurveillancespecimens.'Accordingly,youarerequestedtosupplyonesignedorfgfnaland39copiesoftheinformationlistedintheenclosurewithin60daysofreceiptofthisletter.This.requestforgenericinformationwasapprovedbyGAOunderablanketclearancenumber8-180225(R0072);thisclearanceexpiresJuly31,1977.Sincerely,OriginalsipnedbyGeorgeLear,ChiefOoeratinReactors8nch¹3officc3a-losureand'cc:nextpageSURNAMCQSeDATC~lormhEC-318(Rev.9-53)AKCM0240DivisonopOpera)ORB¹3~owicki-----....5/..../7.7............ORB%~--JZwetzig'----.....5l..g./.7.7............,...ORB¹3~--CPaZr1sfAmj-5/..5/.7.7.............,QU,4,ODVCANMCNTfNINTIN4OffCCI1474S20eSSORB¹3BLeer~.ah.a.7.............
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| | ~~Docket ORB ¹3 Local PDR |
| }} | | ~4l)'g 0 ig7 Docket No. 50-220 NRC PDR GLear CParrish JZwetzig Niagara flohawk Power Corporation SNowicki ATTN: tfr. Gerald K. Rhode Attorney, OELD Vice President - Engineering OI8E (3) 300 Erie Boulevard West DEisenhut Syracuse, Nevi York 13202 TBAbernathy JRBuchanan Gentlemen: ACRS (16) |
| | RE: NINE MILE POINT NUCLEAR STATION UNIT NO. 1 Our review of data received from reactor vessel material surveillance programs indicates that the materials used in reactor vessel fabrication may have a wider variation fn sensitivity to radiation damage than originally anticipated. In addition, some reactor vessels incorporate more than one heat of materials, including vleld metals fn their beltlfne regions, but all of these heats may not be included in the reactor vessel material surveillance program. |
| | Although our review of these data does not reveal a basis for concern regarding continued reactor vessel integrity over the next several years, the information does indicate the need for a detailed review of the materials employed in reactor vessel construction (fn light of this recent data) and a review of the specimens employed in the surveillance program to determine ff the present specimens reasonably represent the limiting materials fn the .reactor vessel beltlfne region. |
| | In order to perform these reviews, we will need the information listed fn the enclosure relative to each of your reactor vessel(s) and assocfated surveillance specimens. |
| | Accordingly, you are requested to supply one signed orfgfnal and 39 copies of the information listed in the enclosure within60 days of receipt of thi s letter. |
| | This. request for generic information was approved by GAO under a blanket clearance number 8-180225 (R0072); this clearance expires July 31, 1977. |
| | Sincerely, Original sipned by George Lear, Chief Ooeratin Reactors 8 nch ¹3 of ficc3a-Divis on op Opera) ORB ¹3 SURNAMCQ DATC~ |
| | Se losure and next page |
| | 'cc : ORB ¹3~ |
| | --CPaZr1 sfAmj ORB ¹3~ ORB%~ |
| | owi cki----- -- JZwetzi g'---- BLeer ~ |
| | - 5/.. 5/.7.7............., ....5/ ..../7.7............ .....5l..g./.7.7............,... .ah.a.7............. |
| | lorm hEC-318 (Rev. 9-53) AKCM 0240 Q U, 4, ODVCANMCNT fNINTIN4 Off CCI 1474 S20 eSS |
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Category:Letter
MONTHYEARNMP1L3570, Supplemental Information Letter - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation2024-02-0101 February 2024 Supplemental Information Letter - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation IR 05000220/20230042024-02-0101 February 2024 Integrated Inspection Report 05000220/2023004 and 05000410/2023004 NMP1L3569, CFR 50.46 Annual Report2024-01-26026 January 2024 CFR 50.46 Annual Report ML24004A2122024-01-0808 January 2024 Senior Reactor and Reactor Operator Initial License Examinations ML23354A0012024-01-0404 January 2024 Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0059 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) ML23278A1292023-12-14014 December 2023 Units 1 & 2; Limerick, Units 1 & 2; Nine Mile Point, Units 1 & 2; and Peach Bottom, Units 2 & 3 -Revision to Approved Alternatives to Use Boiling Water Reactor Vessel and Internals Project Guidelines NMP1L3566, Radiological Emergency Plan Document Revision. Includes EP-AA-1013, Revision 10, Radiological Emergency Plan Annex for Nine Mile Point Station2023-12-14014 December 2023 Radiological Emergency Plan Document Revision. Includes EP-AA-1013, Revision 10, Radiological Emergency Plan Annex for Nine Mile Point Station IR 05000410/20243012023-12-14014 December 2023 Initial Operator Licensing Examination Report 05000410/2024301 ML23305A1402023-12-13013 December 2023 Units 1 & 2; Nine Mile Point, Unit 2; Peach Bottom, Units 2 & 3; and Quad Cities, Units 1 and 2 - Issuance of Amendments to Adopt Traveler TSTF-580 NMP1L3564, Supplemental Response to Part 73 Exemption Request - Withdrawal of Request for Exemption from 10 CFR 73, Subpart B, Preemption Authority Requirements2023-12-0707 December 2023 Supplemental Response to Part 73 Exemption Request - Withdrawal of Request for Exemption from 10 CFR 73, Subpart B, Preemption Authority Requirements ML23291A4642023-12-0707 December 2023 Issuance of Amendment No. 251 Regarding the Adoption of Title 10 the Code of Federal Regulations Section 50.69, Risk-Informed Categorization and Treatment of SSC for Nuclear Power Plants ML23289A0122023-12-0606 December 2023 Issuance of Amendment No. 250 Regarding the Revision to Technical Specifications to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b NMP1L3563, Submittal of Relief Request I5R-12, Revision 0, Concerning the Installation of a Full Structural Weld Overlay on Reactor Pressure Vessel Recirculation Inlet Nozzle N2E Safe End-to-Nozzle Dissimilar Metal Weld (32-WD-208)2023-12-0404 December 2023 Submittal of Relief Request I5R-12, Revision 0, Concerning the Installation of a Full Structural Weld Overlay on Reactor Pressure Vessel Recirculation Inlet Nozzle N2E Safe End-to-Nozzle Dissimilar Metal Weld (32-WD-208) IR 05000220/20234022023-11-28028 November 2023 Security Baseline Inspection Report 05000220/2023402 and 05000410/2023402 NMP1L3557, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-11-22022 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation ML23317A1192023-11-10010 November 2023 Constellation Energy Generation, LLC - 2023 Annual Report - Guarantees of Payment of Deferred Premiums IR 05000220/20234202023-11-0101 November 2023 Security Baseline Inspection Report 05000220/2023420 and 05000410/2023420 ML23305A0052023-11-0101 November 2023 Operator Licensing Examination Approval IR 05000220/20230032023-10-25025 October 2023 Integrated Inspection Report 05000220/2023003 and 05000410/2023003 IR 05000220/20235012023-10-17017 October 2023 Emergency Preparedness Biennial Exercise Inspection Report 05000220/2023501 and 05000410/2023501 IR 05000220/20230112023-10-16016 October 2023 Comprehensive Engineering Team Inspection Report 05000220/2023011 and 05000410/2023011 RS-23-097, Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans2023-10-12012 October 2023 Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans NMP1L3554, Submittal of Revision 28 to the Final Safety Analysis Report (Updated), Fire Protection Design Criteria Document, 10CFR50.59 Evaluation Summary Report, 10CFR54.37(b) Aging Management Review, and Technical Specifications with Revised Bases2023-10-0606 October 2023 Submittal of Revision 28 to the Final Safety Analysis Report (Updated), Fire Protection Design Criteria Document, 10CFR50.59 Evaluation Summary Report, 10CFR54.37(b) Aging Management Review, and Technical Specifications with Revised Bases C IR 05000220/20233032023-09-20020 September 2023 Retake Operator Licensing Examination Report 05000220/2023303 ML23250A0822023-09-19019 September 2023 Regulatory Audit Summary Regarding LARs to Adopt TSTF-505, Rev. 2, and 10 CFR 50.69 ML23257A1732023-09-14014 September 2023 Requalification Program Inspection IR 05000220/20230052023-08-31031 August 2023 Updated Inspection Plan for Nine Mile Point Nuclear Station, Units 1 and 2 (Report 05000220/2023005 and 05000410/2023005) RS-23-080, Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs2023-08-30030 August 2023 Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs NMP2L2851, Relief Request Associated with Successive Inspections for Generic Letter 88-01 / BWRVIP-75-A Augmented Examinations2023-08-25025 August 2023 Relief Request Associated with Successive Inspections for Generic Letter 88-01 / BWRVIP-75-A Augmented Examinations ML23151A3472023-08-21021 August 2023 Issuance of Amendments to Adopt TSTF-295-A, Modify Note 2 to Actions of PAM Table to Allow Separate Condition Entry for Each Penetration NMP1L3534, License Amendment Request - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation2023-08-18018 August 2023 License Amendment Request - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation ML23220A0262023-08-0808 August 2023 Licensed Operator Positive Fitness-for-Duty Test IR 05000220/20234012023-08-0808 August 2023 Cyber Security Inspection Report 05000220/2023401 and 05000410/2023401 (Cover Letter Only) NMP1L3545, Supplemental Information Letter to Adopt TSTF-505, Provide Risk- Informed Extended Completion Times - RITSTF Initiative 4b, Revision 2 and 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems .2023-08-0404 August 2023 Supplemental Information Letter to Adopt TSTF-505, Provide Risk- Informed Extended Completion Times - RITSTF Initiative 4b, Revision 2 and 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems . RS-23-087, Revision to Approved Alternatives Associated with the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor2023-08-0404 August 2023 Revision to Approved Alternatives Associated with the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor IR 05000220/20230022023-08-0101 August 2023 Integrated Inspection Report 05000220/2023002 and 05000410/2023002 ML23207A0762023-07-14014 July 2023 EN 56557 - Update to Part 21 Report Re Potential Defect with Trane External Auto/Stop Emergency Stop Relay Card Pn: XI2650728-06 NMP1L3544, Fifth Inservice Inspection Interval, First Inservice Inspection Period 2023 Owner'S Activity Report for RFO-27 Inservice Examinations2023-07-14014 July 2023 Fifth Inservice Inspection Interval, First Inservice Inspection Period 2023 Owner'S Activity Report for RFO-27 Inservice Examinations ML23186A1642023-07-0606 July 2023 Operator Licensing Retake Examination Approval NMP2L2846, Nine Mire Point Nuclear Station, Units 1 and 2, General License 30-day Cask Registration Notifications2023-07-0505 July 2023 Nine Mire Point Nuclear Station, Units 1 and 2, General License 30-day Cask Registration Notifications ML23192A0622023-06-30030 June 2023 Engine Systems, Inc., 10CFR21 Reporting of Defects and Non-Compliance, Report No. 10CFR21-0136, Rev. 0 IR 05000220/20230102023-06-29029 June 2023 Biennial Problem Identification and Resolution Inspection Report 05000220/2023010 and 05000410/2023010 ML23131A4242023-06-23023 June 2023 Issuance of Amendment No. 249 Regarding the Revision to Technical Specification 3.3.1 to Adopt Technical Specifications Task Force Traveler TSTF-568 RS-23-077, Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations2023-06-16016 June 2023 Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations NMP1L3539, Day Commitment Response - Relief Request I5R-11 Concerning the Installation of a Weld Overlay on Reactor Pressure Vessel Recirculation Inlet Nozzle N2E Safe End-to-Nozzle Dissimilar Metal Weld (32-WD-208)2023-06-0909 June 2023 Day Commitment Response - Relief Request I5R-11 Concerning the Installation of a Weld Overlay on Reactor Pressure Vessel Recirculation Inlet Nozzle N2E Safe End-to-Nozzle Dissimilar Metal Weld (32-WD-208) ML23159A0052023-06-0505 June 2023 56557-EN 56557 - Paragon - Redlined RS-23-042, Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling2023-05-25025 May 2023 Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling IR 05000410/20233022023-05-15015 May 2023 Initial Operator Licensing Examination Report 05000410/2023302 2024-02-01
[Table view] Category:Request for Additional Information (RAI)
MONTHYEARML23264A7992023-09-21021 September 2023 NRR E-mail Capture - Final RAI - Constellation Energy Generation, LLC Fleet Request License Amendment Request to Adopt TSTF-580, Revision 1 ML23205A2432023-07-19019 July 2023 NRC Staff Follow-up Question on Audit Question 18 TSTF-505 and 50.69 Regulatory Audit (E-mail Dated 7/19/2023) (EPIDs L-2022-LLA-0185 and L-2022-LLA-0186) ML23087A2912023-03-28028 March 2023 Request for Additional Information (3/28/2023 E-mail) - Proposed Emergent I5R-11 Alternative Associated with a Weld Overlay on RPV Recirculation Nozzle N2E DM Weld ML23061A0522023-03-0202 March 2023 Request for Additional Information (3/2/2023 E-mail) - Proposed Alternative Associated with a Weld Overlay Repair to the Torus ML23012A2002023-01-13013 January 2023 Information Request for the Cyber Security Baseline Inspection, Notification to Perform Inspection 05000220/2023401 and 05000410/2023401 ML22207A2162022-07-26026 July 2022 Information Request to Support Triennial Baseline Design-Basis Capability of Power-Operated Valves Inspection; Inspection Report 05000220/2022010 and 05000410/2022010 ML22194A9412022-07-13013 July 2022 Request for Additional Information Relief Request CS-PR-02 (7/13/2022 e-mail) ML22041B5362022-02-10010 February 2022 NRR E-mail Capture - Constellation Energy Generation, LLC - Request for Additional Information Regarding Fleet License Amendment Request to Adopt TSTF-541 ML22020A0642022-01-13013 January 2022 NRR E-mail Capture - Exelon Generation Company, LLC - Request for Additional Information Regarding Proposed Fleet Alternative for Repair of Water Level Instrumentation Partial Penetration Nozzles ML21320A3472021-11-16016 November 2021 Request for Additional Information LAR to Revise TSs to Adopt TSTF-582, Revision 0 ML21306A3312021-11-0202 November 2021 Request for Additional Information Alternative Request GV-RR-10 (11/2/2021 e-mail) ML21256A1902021-09-10010 September 2021 NRR E-mail Capture - Exelon Generation Company, LLC - Request for Additional Information Regarding License Transfer Application ML21144A2132021-05-24024 May 2021 NRR E-mail Capture - Exelon Generation Company, LLC - Request for Additional Information Regarding License Transfer Application ML21117A0342021-05-0505 May 2021 Request for Additional Information Regarding Proposed Alternative to Use ASME Code Case N-893 ML21125A1282021-05-0404 May 2021 OPC Document Request - Feb 2021 ML21110A5112021-04-20020 April 2021 Request for Additional Information Review of License Amendment Request to Revise Technical Specifications to Adopt TSTF-582 ML21088A2682021-03-30030 March 2021 Notification of Conduct of a Fire Protection Team Inspection ML21062A0652021-03-0101 March 2021 NRR E-mail Capture - Exelon Generation Company, LLC - Request for Additional Information Regarding Proposed Fleet Alternative to Documentation Requirements for Pressure Retaining Bolting ML21049A2572021-02-18018 February 2021 Request for Additional Information Byron/Dresden Proposed Changes to Site Emergency Plans to Support Post-Shutdown and Permanently Defueled Conditions (EPID-2020-LLA-0240 & EPID-2020-LLA-0237) ML20365A0092020-12-30030 December 2020 Request for Additional Information Concerning Review of License Amendment Request and Relief Request to Change Excess Flow Check Valve Testing Frequency (EPIDs L-2020-LLA-0188 and L-2020-LLR-0114) ML20358A2602020-12-28028 December 2020 Changes to Draft Request for Additional Information Regarding License Amendment Request and Relief Request to Change Excess Flow Check Valve Testing Frequency ML20272A2802020-09-28028 September 2020 Withdrawal and Replacement of Request for Additional Information to Support Review of License Amendment Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times ML20248H5192020-09-28028 September 2020 Changes to Draft Request for Additional Information Regarding Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times ML20246G6362020-09-0202 September 2020 Request for Additonal Information to Support Review of License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times NMP2L2739, Request for Additional Information for Nine Mile Point Nuclear Station, Unit 2, to Adopt TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, Revision 22020-08-28028 August 2020 Request for Additional Information for Nine Mile Point Nuclear Station, Unit 2, to Adopt TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, Revision 2 ML20239A7982020-08-25025 August 2020 NRR E-mail Capture - Exelon Generation Company, LLC - Fleet License Amendment Request to Adopt TSTF-568, Revision 2 ML20212L8702020-07-30030 July 2020 Request for Additonal Information Review of License Amendment Requests Regarding Riskinformed Categorization and Treatment of Structures, Systems and Components (L-2019-LLA-0290) ML20213A9352020-07-30030 July 2020 Request for Additional Information Review of License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times ML20153A7042020-06-0101 June 2020 NRR E-mail Capture - Preliminary RAI for Fleet Request to Use Alternative OMN-26 ML20135H1972020-05-14014 May 2020 NRR E-mail Capture - Exelon Generation Company, LLC - Request for Additional Information Regarding Request to Extend Safety Relief Valve Test Interval ML20045E3582020-02-14014 February 2020 Draft Request for Additional Information Regarding License Amendment Request to Increase Allowable MSIV Leakage Rates ML19296A1862019-10-23023 October 2019 Request for Additional Information Regarding License Amendment Request to Increase Allowable MSIV Leakage Rates (L-2019-LLA-0115) ML19275H1362019-10-0202 October 2019 NRR E-mail Capture - Exelon Generation Company, LLC - Request for Additional Information Regarding Request to Use ASME Code Case N-879 ML19179A0612019-07-19019 July 2019 Three Mile Point 1 - Supplemental Information Needed to Proposed Alternative to Use ASME Code Case N-879 ML19151A8132019-05-31031 May 2019 Licensed Operator Positive Fitness-For-Duty Test ML19025A1572019-01-25025 January 2019 Request for Additional Information Regarding Primary Containment Oxygen Concentration License Amendment Request ML19025A1202019-01-24024 January 2019 NRR E-mail Capture - Calvert Cliffs, Fitzpatrick, and Nine Mile Point - Request for Additional Information Regarding License Amendment Request to Revise Emergency Response Organization Staffing ML18341A2212018-12-0707 December 2018 Request for Additional Information Regarding Emergency Tech Spec Change Re HPCS Completion Time (EPID -L-2018-LLA-0491) ML18228A6932018-08-15015 August 2018 Request for Additional Information Regarding Reactor Pressure Vessel Water Inventory Control License Amendment Request (L-2017-LLA-0426) ML18205A3922018-07-24024 July 2018 Request for Additional Information Regarding Reactor Pressure Vessel Water Inventory Control License Amendment Request (L-2017-LLA-0426) ML18184A2882018-07-0303 July 2018 Request for Additional Information Regarding Removal of Boraflex Credit from Spent Fuel Pool License Amendment Request(L-2018-LLA-0039) ML18102A2372018-04-12012 April 2018 NRR E-mail Capture - Calvert Cliffs, Ginna, and Nine Mile Point - Request for Additional Information Regarding License Amendment Request to Revise Emergency Action Level Schemes (EPID-L-2017-LLA-0237) ML17331B1342017-12-12012 December 2017 2, and R.E. Ginna Nuclear Power Plant - Request for Additional Information - Regarding ML17285B1962017-10-27027 October 2017 Request for Additional Information Regarding Generic Letter 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools. ML17272A0112017-10-10010 October 2017 Request for Additional Information Regarding License Amendment Concerning Reactor Pressure Vessel Water Inventory Control ML17234A3592017-08-30030 August 2017 Request for Additional Information Regarding Relief Request NMP-RR-001 to Utilize Code Case N-702 ML17241A2752017-08-29029 August 2017 Request for Additional Information Response MSA (MF7946,7) and FE (MG0087,8) E-mail Attachment ML17240A3102017-08-15015 August 2017 NRR E-mail Capture - Nine Mile Point MSA and FE RAI ML17172A0842017-06-27027 June 2017 Request for Additional Information Regarding Relief Request to Utilize ASME Code Case N-702 ML17065A1622017-03-14014 March 2017 Request for Additional Information Regarding Proposed Alternative to Perform Pressure Isolation Valve Leakage Testing at Frequencies Consistent with 10 CFR Part 50, Appendix J 2023-09-21
[Table view] Category:Technical Specification
MONTHYEARRS-23-042, Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling2023-05-25025 May 2023 Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling NMP1L3482, License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2022-12-15015 December 2022 License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b ML22298A0112022-10-25025 October 2022 License Amendment Request to Adopt Technical Specification Task Force TSTF-295-A, Modify Note 2 to Actions of (Post-Accident Monitoring) PAM Table To. ML22292A1312022-10-0505 October 2022 5 to Updated Final Safety Analysis Report, Technical Specification Bases, Rev. 69 RS-22-092, Nine and Quad Cities - Application to Revise Primary Containment Isolation Instrumentation Technical Specifications in Accordance with TSTF-306, Revision 2, Add Action to LCO 3.3.6.1 to Give Option to Isolate the Penetration2022-10-0303 October 2022 Nine and Quad Cities - Application to Revise Primary Containment Isolation Instrumentation Technical Specifications in Accordance with TSTF-306, Revision 2, Add Action to LCO 3.3.6.1 to Give Option to Isolate the Penetration NMP1L3477, License Amendment Request - Application to Partially Adopt Technical Specification Task Force (TSTF) Traveler TSTF-568, Revision 2, Revise Applicability of BWR/4 TS 3.6.2.5 and TS 3.6.3.2, to Revise.2022-08-12012 August 2022 License Amendment Request - Application to Partially Adopt Technical Specification Task Force (TSTF) Traveler TSTF-568, Revision 2, Revise Applicability of BWR/4 TS 3.6.2.5 and TS 3.6.3.2, to Revise. NMP2L2780, Application to Revise Technical Specifications to Adopt TSTF-582, Revision 0, Reactor Pressure Vessel Water Inventory Control (RPV WIC) Enhancements2021-09-30030 September 2021 Application to Revise Technical Specifications to Adopt TSTF-582, Revision 0, Reactor Pressure Vessel Water Inventory Control (RPV WIC) Enhancements NMP2L2762, License Amendment Request Adoption of Technical Specification Task Force (TSTF) Traveler TSTF-501, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control2021-05-26026 May 2021 License Amendment Request Adoption of Technical Specification Task Force (TSTF) Traveler TSTF-501, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control JAFP-20-0049, Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability2020-06-22022 June 2020 Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability NMP2L2718, Supplemental Information No.1 for Nine Mile Point Nuclear Station, Unit 2, to Adopt TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, Revision 22019-12-12012 December 2019 Supplemental Information No.1 for Nine Mile Point Nuclear Station, Unit 2, to Adopt TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, Revision 2 JAFP-19-0067, Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability2019-06-27027 June 2019 Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability RS-19-039, Fleet License Amendment Request - Common Language for Technical Specification High Radiation Area Administrative Controls2019-06-26026 June 2019 Fleet License Amendment Request - Common Language for Technical Specification High Radiation Area Administrative Controls NMP2L2693, Technical Specification Bases, Rev. 572018-10-24024 October 2018 Technical Specification Bases, Rev. 57 NMP2L2664, Supplemental Information No. 2 for Nine Mile Point Nuclear Station, Unit 2, to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 22018-01-12012 January 2018 Supplemental Information No. 2 for Nine Mile Point Nuclear Station, Unit 2, to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 2 NMP2L2652, License Amendment Request - Revise Surveillance Requirement 3.5.1.2 to Remove Note2017-08-22022 August 2017 License Amendment Request - Revise Surveillance Requirement 3.5.1.2 to Remove Note NMP2L2640, License Amendment Request - Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 22017-02-28028 February 2017 License Amendment Request - Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 2 ML16309A4002016-10-24024 October 2016 Revision 47 to Technical Specification Bases, List of Effective Pages ML16309A4022016-10-24024 October 2016 Revision 47 to Technical Specification Bases, Section B2.0, Safety Limits (Sls) ML16309A4032016-10-24024 October 2016 Revision 47 to Technical Specification Bases, Section B3.0, Limiting Condition for Operation (LCO) Applicability ML16309A4012016-10-24024 October 2016 Revision 47 to Technical Specification Bases, Table of Contents ML16309A4082016-10-24024 October 2016 Revision 47 to Technical Specification Bases, Section B3.5, Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) System ML16309A4132016-10-24024 October 2016 Revision 47 to Technical Specification Bases, Section B3.9, Refueling Operations ML16309A4052016-10-24024 October 2016 Revision 47 to Technical Specification Bases, Section B3.2, Power Distribution Limits ML16209A2182016-07-26026 July 2016 Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing. NMP1L3070, License Amendment Request - Reactivity Anomalies Surveillance2016-03-18018 March 2016 License Amendment Request - Reactivity Anomalies Surveillance NMP2L2592, Supplemental Information Regarding TSTF-425 License Amendment Request2015-09-24024 September 2015 Supplemental Information Regarding TSTF-425 License Amendment Request NMP1L3027, Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)2015-05-12012 May 2015 Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3) ML15089A2312015-03-26026 March 2015 Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3) NMP1L3010, Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)2015-03-26026 March 2015 Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3) NMP2L2576, License Amendment Request - Relocation of Secondary Containment Bypass Leakage Paths Table from Technical Specifications to the Technical Requirements Manual2015-03-23023 March 2015 License Amendment Request - Relocation of Secondary Containment Bypass Leakage Paths Table from Technical Specifications to the Technical Requirements Manual NMP2L2563, License Amendment Request - Primary Containment Isolation Instrumentation Technical Specification Allowable Value Change2014-11-17017 November 2014 License Amendment Request - Primary Containment Isolation Instrumentation Technical Specification Allowable Value Change ML14191A2552014-07-10010 July 2014 Nine Mile Point and Ginna, Unit 1 - Application to Revise Technical Specifications to Adopt TSTF-523, Generic Letter 2008-01, Managing Gas Accumulation, Using the Consolidated Line Item Improvement Process ML12156A3632012-06-0101 June 2012 Emergency License Amendment Request Pursuant to 10 CFR 50.90: Revision of the Main Steam Isolation Valve Allowable Leakage Rate Limit ML11306A2982011-10-27027 October 2011 Submittal of Revision 22 to the Final Safety Analysis Report (Updated), 10 CFR 50.59 Evaluation Summary Report, Technical Specifications Bases Changes, and Report Consistent with 10 CFR 54.37(b) ML11214A2142011-07-25025 July 2011 License Amendment Request, Extension of the Completion Time for Inoperable Diesel Generator - Response to NRC Request for Additional Information ML1109602832011-03-30030 March 2011 License Amendment Request Pursuant to 10 CFR 50.90: Request for Adoption of Technical Specification Task Force Traveler TSTF-514, Revision 3, Revise BWR Operability Requirements and Actions for RCS Leakage ... ML0935000902009-12-0909 December 2009 License Amendment Request to Remove Operating Mode Restrictions for Performing Surveillance Testing of the Division 3 Battery - Technical Specification 3.8.4, DC Sources - Operating ML0931602572009-10-26026 October 2009 Submittal of Revision 21 to the Final Safety Analysis Report (Updated), 10 CFR 50.59 Evaluation Summary Report, Technical Specifications Bases Changes, and Report Consistent with 10 CFR 54.37(b) ML0907802502009-03-0909 March 2009 Submittal of Amendment Request for Technical Specification Improvement to Revise Control Rod Scram Time Testing Frequency ML0906403012009-03-0303 March 2009 License Amendment Request Pursuant to 10 CFR 50.90: Relocation of Pressure and Temperature Limit Curves to the Pressure and Temperature Limits Report ML0827502992008-09-29029 September 2008 Technical Specifications ML0802501622008-01-24024 January 2008 Response to NRC Request for Additional Information Regarding Nine Mile Point Nuclear Station, Unit No. 1, Revision of Rod Worth Minimizer Limiting Condition for Operation During Startup ML0726401032007-09-19019 September 2007 License Amendment Request for Adoption of TSTF-484, Revision 0, Use of TS 3.10.1 for Scram Time Testing Activities. ML0721906052007-07-30030 July 2007 Application for Technical Specification Change TSTF-477, Rev. 3, Add Action for Two Inoperable Control Room AC Subsystems to the Technical Specifications Using Consolidated Line Item Improvement Process ML0721205922007-07-23023 July 2007 Application for Technical Specification Improvement to Adopt TSTF-476, Revison 1, Improved BPWS Control Rod Insertion Process (NEDO-33091), Using the Consolidated Line Item Improvement ML0721205872007-07-23023 July 2007 License Amendment Request Pursuant to 10 CFR 50.90: Revision of Rod Worth Minimizer Limiting Condition for Operation During Startup - Technical Specification Section 3.1.1.b ML0720504232007-07-12012 July 2007 Application to Revise Technical Specifications Regarding Control Room Envelope Habitability in Accordance with TSTF-448, Revision 3, Using the Consolidated Line Item Improvement Process ML0720504742007-07-12012 July 2007 Application to Revise Technical Specifications Regarding Control Room Envelope Habitability in Accordance with TSTF-448, Revision 3, Using the Consolidated Line Item Improvement Process ML0707800592007-03-15015 March 2007 Technical Specification 4.1.1 Regarding Control Rod System Surveillance Requirements ML0630406022006-10-19019 October 2006 License Amendment Request Pursuant to 10 CFR 50.90: Revision to Control Rod System Surveillance Requirements - Technical Specification 4.1.1 2023-05-25
[Table view] Category:Amendment
MONTHYEARRS-23-042, Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling2023-05-25025 May 2023 Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling NMP1L3482, License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b2022-12-15015 December 2022 License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b ML22298A0112022-10-25025 October 2022 License Amendment Request to Adopt Technical Specification Task Force TSTF-295-A, Modify Note 2 to Actions of (Post-Accident Monitoring) PAM Table To. RS-22-092, Nine and Quad Cities - Application to Revise Primary Containment Isolation Instrumentation Technical Specifications in Accordance with TSTF-306, Revision 2, Add Action to LCO 3.3.6.1 to Give Option to Isolate the Penetration2022-10-0303 October 2022 Nine and Quad Cities - Application to Revise Primary Containment Isolation Instrumentation Technical Specifications in Accordance with TSTF-306, Revision 2, Add Action to LCO 3.3.6.1 to Give Option to Isolate the Penetration NMP2L2780, Application to Revise Technical Specifications to Adopt TSTF-582, Revision 0, Reactor Pressure Vessel Water Inventory Control (RPV WIC) Enhancements2021-09-30030 September 2021 Application to Revise Technical Specifications to Adopt TSTF-582, Revision 0, Reactor Pressure Vessel Water Inventory Control (RPV WIC) Enhancements JAFP-20-0049, Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability2020-06-22022 June 2020 Application to Adopt TSTF-427, Revision 2, Allowance for Non Technical Specification Barrier Degradation on Supported System Operability NMP2L2718, Supplemental Information No.1 for Nine Mile Point Nuclear Station, Unit 2, to Adopt TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, Revision 22019-12-12012 December 2019 Supplemental Information No.1 for Nine Mile Point Nuclear Station, Unit 2, to Adopt TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, Revision 2 RS-19-039, Fleet License Amendment Request - Common Language for Technical Specification High Radiation Area Administrative Controls2019-06-26026 June 2019 Fleet License Amendment Request - Common Language for Technical Specification High Radiation Area Administrative Controls NMP2L2664, Supplemental Information No. 2 for Nine Mile Point Nuclear Station, Unit 2, to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 22018-01-12012 January 2018 Supplemental Information No. 2 for Nine Mile Point Nuclear Station, Unit 2, to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 2 NMP2L2652, License Amendment Request - Revise Surveillance Requirement 3.5.1.2 to Remove Note2017-08-22022 August 2017 License Amendment Request - Revise Surveillance Requirement 3.5.1.2 to Remove Note NMP2L2640, License Amendment Request - Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 22017-02-28028 February 2017 License Amendment Request - Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control, Revision 2 NMP1L3027, Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)2015-05-12012 May 2015 Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3) NMP2L2576, License Amendment Request - Relocation of Secondary Containment Bypass Leakage Paths Table from Technical Specifications to the Technical Requirements Manual2015-03-23023 March 2015 License Amendment Request - Relocation of Secondary Containment Bypass Leakage Paths Table from Technical Specifications to the Technical Requirements Manual NMP2L2563, License Amendment Request - Primary Containment Isolation Instrumentation Technical Specification Allowable Value Change2014-11-17017 November 2014 License Amendment Request - Primary Containment Isolation Instrumentation Technical Specification Allowable Value Change ML14191A2552014-07-10010 July 2014 Nine Mile Point and Ginna, Unit 1 - Application to Revise Technical Specifications to Adopt TSTF-523, Generic Letter 2008-01, Managing Gas Accumulation, Using the Consolidated Line Item Improvement Process ML1109602832011-03-30030 March 2011 License Amendment Request Pursuant to 10 CFR 50.90: Request for Adoption of Technical Specification Task Force Traveler TSTF-514, Revision 3, Revise BWR Operability Requirements and Actions for RCS Leakage ... ML0935000902009-12-0909 December 2009 License Amendment Request to Remove Operating Mode Restrictions for Performing Surveillance Testing of the Division 3 Battery - Technical Specification 3.8.4, DC Sources - Operating ML0907802502009-03-0909 March 2009 Submittal of Amendment Request for Technical Specification Improvement to Revise Control Rod Scram Time Testing Frequency ML0906403012009-03-0303 March 2009 License Amendment Request Pursuant to 10 CFR 50.90: Relocation of Pressure and Temperature Limit Curves to the Pressure and Temperature Limits Report ML0827502992008-09-29029 September 2008 Technical Specifications ML0720504742007-07-12012 July 2007 Application to Revise Technical Specifications Regarding Control Room Envelope Habitability in Accordance with TSTF-448, Revision 3, Using the Consolidated Line Item Improvement Process ML0707800592007-03-15015 March 2007 Technical Specification 4.1.1 Regarding Control Rod System Surveillance Requirements ML0536202882005-12-16016 December 2005 License Amendment Request: Proposed Changes to Technical Specification 4.1.4 Regarding Core Spray Instrumentation ML0511100732005-04-19019 April 2005 Tech Spec Pages for Amendment 115 and 188 Reporting Requirements for Annual Radiological Environmental Operating Report ML0510204002005-04-0101 April 2005 License Amendment Request Pursuant to 1O CFR 50.90: Request for Amendment to Extend Completion Time for Emergency Uninterruptible Power Supply Inverters ML0503304712005-01-24024 January 2005 Application for Technical Specification Improvement to Eliminate Requirements to Provide Monthly Operating Reports and Occupational Radiation Exposure Reports ML0424003162004-08-17017 August 2004 Unit I, License Amendment Request Pursuant to 10 CFR 50.90: Addition of 24 Hours for Restoration of Primary Containment Oxygen Concentration ML0421103812004-07-16016 July 2004 License Amendment Request Pursuant to 10 CFR 50.90: Revision of Intermediate Range Monitor Surveillance Frequency and Relocation of Selected Instrumentation Requirements to a License-Controlled Document ML0403302112004-01-27027 January 2004 TS Amendment Pages for License No. NPF-069 ML0330305992003-10-27027 October 2003 TS Pages Amendment 183 ML0324608472003-08-15015 August 2003 License Amendment Request Pursuant to 10 CFR 50.90: Revision of Reactor Pressure Vessel Pressure-Temperature Limits ML0233807302002-11-22022 November 2002 Facility Operating License DPR-63, License Amendment Request: Technical Specifications Section 6.0, Administrative Controls - Response to RA Information TAC No. MB2441 ML0229603382002-10-0707 October 2002 Application for Technical Specification Change Regarding Missed Surveillances Using the Consolidated Line Item Improvement Process ML0217101682002-06-0707 June 2002 Application for Technical Specification Improvement to Eliminate the Requirements for the Post Accident Sampling System Using the Consolidated Line Item Improvement Process ML0207305992002-03-13013 March 2002 Technical Specifications, Issuance of Amendment Safety Limit Minimum Critical Power Ratio (Tac No. MB3327) ML0206302682002-03-0101 March 2002 Technical Specifications for Amendment No. 103 ML0204301512002-02-11011 February 2002 Issuance of Amendment Ventilation Requirements During Irradiated Fuel Handling ML18018B0461978-03-22022 March 1978 Letter Transmitting an Application for Amendment to Appendix a of the Operating License ML17037C4331977-08-0808 August 1977 Letter Regarding Surveillance Requirements and Limiting Conditions for Operations and Requesting an Application for Amendment to the Operating License That Will Change the Technical Specifications ML17037C4131977-07-22022 July 1977 Letter Enclosing Application for Amendment to Operating License and Proposed Changes to Technical Specifications Concerning Changes in Management and Organization ML17037C4141977-07-18018 July 1977 Letter Enclosing an Application for Amendment to Operating License and Proposed Changes to the Technical Specifications to Provide for More Flexible Operation and Concerning Load Line Limit Analysis ML17037C4161977-07-14014 July 1977 Application for Amendment to Operating License and Proposed Change to Technical Specification Concerning the Emergency Cooling System ML17037C4181977-05-19019 May 1977 Letter Enclosing an Application to Amend Operating License and a Requested Change in the Technical Specifications Relating to Instrumentation Used to Monitor Drywell Suppression Chamber Differential Pressure and Suppression .. ML17037C4341977-03-0404 March 1977 Letter Regarding a Preliminary Review of the December 7, 1976 Request for Technical Specification Changes to Reactor Operating Limits After Reload for Cycle 5 Operation and Requesting Additional Information to Complete Review ML17037C4391976-09-30030 September 1976 Letter Requesting Submittal of an Application for License Amendment to Incorporate the Requirements of the Model Technical Specifications ML17037C4531976-01-19019 January 1976 Letter Transmitting an Application for Amendment to Operating License and a Proposed Change to Technical Specifications (Appendix a) to Clarify Cloride and Oxygen Levels ML17037C1771975-12-19019 December 1975 Letter Enclosing a Corrected Attachment to License Amendment No. 11 Change No. 27 to the Technical Specifications Provisional Operating License No. DPR-16 ML17037C1831975-11-0303 November 1975 Letter Regarding Application for Amendment to Technical Specifications and an Inadvertent Error Not Made on the Persons Listed in the Certificate of Service ML17037C4681975-09-23023 September 1975 Letter Regarding a Stress Assisted Corrosion Problem and an Enclosed Safety Evaluation by the Office of NRR - Supporting Amendment to License No. DPR-63 and Changes to the Technical Specifications Inoperable Control Rod .. ML17037C4541975-09-0808 September 1975 Letter Transmitting an Application for Amendment to Operating License and a Proposed Change to Technical Specifications Correcting Errors in a Submittal Made August 15, 1975 in Connection with Reload Cycle 5 2023-05-25
[Table view] |
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Distribution AUG 8 1977 Docket ORB 83 CParrish Docket No. 50-220 GLear SNowicki DVerrelli Local PDR Niagara mohawk Power Corporation NRC PDR ATTN: Nr. Gerald K. Rhode Attorney, OELD Vice President - Engineering OI&E (3) 300 Erie Boulevard West DEisenhut Syracuse, New York 13202 TBAbernathy NRBuchanan Gentlemen: ACRS (16)
RE: NINE NILE POINT NUCLEAR STATIOfl UNIT NO. 1 In the past several years, a significant number of relief valves and safety-relief valves were found to be inoperable at BWR reactor facili-ties. These valves were installed in the Reactor Coolant System and/or Automatic Depressurization System. Several programs have been developed to reduce the incidence of these valve failures; however, additional failures continue to occur.
Consequently, we have concluded that changes to the Surveillance Require-ments and Limiting Conditions for Operations for all BHR's are needed to provide additional assurance of relief valve and safety-relief valve operabflity and reliability. Therefore, we request that you modify your surveillance testing program through the adoption of the program contained in the model technical specifications we have prepared. The elements of this program include:
Each remot ly operated relief valve and safety-relief valve fn the Reactor Coolant System and Automatic Depressurizatfon System will be tested on a variable frequency schedule related to demonstrated reliability and operability. The testing interval is based on the number of valve failures during the required test interval. Facilities with reliable valves will progress to a longer test interval while those with valve failures will progress to a shorter test interval. This concept should result in the maintenance of a more uniform level of reliability for this equipment than previously obtained.
- 2. The increased surveillance program will become effective on Harch 1, 19T9. No increase in valve, testing is required before that date.
The initial testing interval. of the increased surveillance program will be based on the number of remotely operated relief valves and safety-relief valves found inoperable in the previous 18 months 1, 1977 to March 1, 1979). This lead time will permit pP'September the resolution of the Hark I Safety-Relief Valve Loads and DFPICE~
SUANAMEW DATE~
NRC FORM'318 (9.76) NRCM 0240 4V 8 OOVEANMENT PAINTIN4 OFFICER 1024 420 624
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r~ p 197) p,UG 8 Niagara Hohawk Power Corporation time is sufficient to permit the development and implementation of improved safety and safety-relfef valve maintenance procedures and
'other corrective actions prior to implementing the test program.
- 3. The relief and/or safety-relief valve line restraints in the torus will be examined prior to initiating the test program and at least once each fuel cycle (f.e., each 18 months) to verify continued structural integrity.
We request that you submit within 30 days from your receipt of this 'letter, an application for amendment to your license that will change your technical specifications to be in conformance wfth the requirements of the enclosed model technical specifications and associated'bases. In the event you should desire further dfscussion of this matter, please contact us.
Sincerely, Original signed bV George Lear, Chief Operating Reactors Branch 83 Division of Operating Reactors
Enclosure:
Model Technical Specifications CC:
Arvin E. Upton, Esquire LeBoeuf, Lamb, Lefby 5 NacRae 1757 N Street, H. W.
Washington, D. C. 20036 Anthony Z. Rofsman, Esquire Rofsman, Kessler and Cashdan 1025 15th Street, N. H.
5th Floor Washington, D. C. 20005 Nr. Eugene G. Saloga, Applicant Coordinator
., Nine Nile Point Energy Information Center P. 0. Box 81 Lycoming, Nes( York 13093 ORB 3 ORB k3 ORB ii3 j ORB 83 OFFICS~
SU/NAME& CParrish SNowick jf DV ~rl I GLear DATE~
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'RC FORM 318 (9-76) NRCM 0240 A U,'0, OOVSANMSNT FIIINTINO OFFICS< ISTS 02~24
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REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.4.2 At least the following reactor coolant system code safety valves and safety-relief valves shall indicated pressures.
be operable with lift settings within + 1% of the (2)* Safety valves 9 (1240) psig (3) Safety-relief valves 9 ( 1100) psig (3 ) Safety-relief valves 9 ( 1090) psig (3) Safety-relief valves 9 (1080) psig APPLICABILITY: With Average Coolant Temperature > 212'F or the Mode Switch in Run, or Startup/Hot Standby.
ACTION:
With one or more reactor coolant system code safety valve(s) or a safety-relief valve(s) inoperable either restore the valve(s) to operable status within 15 minutes or be. shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce Average Coolant Temperature to < 212'F withinthe next 24 hours..
SURVEILLANCE RE UIREMENTS 4.4.2.1 In addition to the applicable ASME Boiler and Pressure Vessel Code,Section XI requirements, each safety-relief valve shall be demonstrated operable:
- a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, by verifying bellows integrity through instrument indication.
- b. Until March 1, 1979, at least once per 18 months by:
- 1. Manually opening each remotely operated .safety-relief valve with the reactor at or below 5% rated power and at nominal operating pressure, and verifying that either:
- a. The turbine bypass valve(s) indicate a compensating valve movement, or
- b. The reactor coolant system pressure decreases by an amount equivalent to the valve pressure relieving capacity for the test conditions.
d d f
REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS (Continued)
- 2. Conducting a visual inspection of the safety-relief valve line restraints in the torus to verify structural integrity for continued operation.
- c. After March 1, 1979, by performance of the following test program:
- 1. Manually opening each remotely operated safety-relief valve in accordance with the test schedule of Table 4.4-10 with the reactor at or below 5X rated power and a steam at nominal operating pressure and verifying that either:
- a. The turbine bypass valve(s) indicate a compensating valve movement, or
- b. The reactor coolant system pressure decreases by an amount equivalent to the valve pressure relieving capacity for the test conditions.
- 2. The initial Next Required Test Interval of Table 4.4-10 shall be determined by the number of remotely operated relief and safety-relief valves found inoperable from September 1, 1977 to March 1, 1979.
- 3. The initial valve tests of Table 4.4-10 shall be completed by, the earlier of:
- a. The completion of the next refueling outage occurring after March 1, 1979, or
- b. The time period defined by March 1, 1979 plus the initial test interval, determined above.
- 4. At least once per 18 months, by conducting a visual inspection of the safety-relief valve line restraints in the torus to verify structural integrity for continued operation.
4.4.2.2 Each safety valve and the safety valve function of each safety-relief valve shall be demonstrated operable per the requirements of the ASME Boiler and Pressure Vessel Code ( ) Edition and Addenda through
( ). V
TABLE 4.4-10 REHOTELY OPERATED RELIEF AND SAFETY-RELIEF VALVE TEST SCHEDULE NUHBER OF REHOTELY OPERATED RELIEF AND SAFETY-RELIEF VALVES NEXT REQUIRED FOUND INOPERABLE DURING TESTING OR TEST INTERVAL** TEST INTERVAL*
0 18 months + 25%
1 184 days + 25K 2 92 days + 25%
3 31 days + 25%
"The required test interval shall not be lengthened more than one step at a time.
Early tests may be performed prior to entering the "next required test interval" (i.e., in advance of the nominal time less the negative 25K tolerance band).
Early tests may be used as a new reference point for tests of the same interval, however, they are not acceptable for lengthening the test interval.
- Setpoint- drift is not considered to be a valve failure for the purposes of this test schedule.
I 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.2 SAFETY VALVES The reactor coolant system safety valves operate to prevent the reactor coolant system from being pressurized above the Safety Limit of psig. Each safety valve is designed to relieve ibs per hoor at the valve set point. The system is designed to meet the ASME Boiler and Pressure Vessel Code requirements that the nuclear system relief valves shal=l function to prevent opening of the safety valves.
Although the safety valve function is not expected to be required under the most limiting transient, an inoperable valve requires shutdown in order to comply with ASME Code requirements.
The testing frequency applicable to the relief valve function of the safety-relief valves is provided to ensure operability and demonstrate reliability of the valves. The required testing interval varies with observed valve failures. The number of inoperable valves found during both operation and testing of these valves determines the time interval for the next required test of these valves. Early tests may be performed prior to entering the next required test interval (i.e., in advance of the nominal time less the negative 25% tolerance band). Early tests may be used as a new reference point for tests of the same time interval, however, they are not acceptable for lengthening the test interval since they were not performed within the +25% tolerance band as required by Table 4.4-10.
Oemonstration of the safety valves'ift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.
EMERGENCY CORE COOLING SYSTEMS AUTOMATIC DEPRESSURIZATION SYSTEM LIMITING CONDITION FOR OPERATION 3.5.2 The Automatic Depressurization System (ADS) shall be OPERABLE with at least (6 )* OPERABLE ADS valves.
APPLICABILITY: With Average Coolant Temperature > 212'F or the Mode Switch in Run, or Startup/Hot Standby.
ACTION:
- a. With one of the above required ADS valves inoperable, operation may continue provided the actuation logic of the remaining ADS valves is operable and the CSS and LPCI systems are operable, and the HPCI system is demonstrated operable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; restore the inoperable ADS valve to operable status wit'iin 14 days or be shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce the Average Coolant Temperature to < 212'F within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- b. With two or more of the above required ADS valves inoperable, be shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce the Average Coolant Temperature to < 212'F within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE RE UIREMENTS 4.5.2 In addition to the applicable ASME Boiler and Pressure Vessel Code,Section XI requirements, the ADS shall be demonstrated operable:
- a. At least once per 18 months by performance of a system functional test which includes simulated automatic actuation through the automatic depressurization sequence, but excluding valve actuation.
- b. Until March 1, 1979, at least once per 18 months by:
- 1. Manually opening each ADS valve with the reactor at or below 5X rated power and at nominal operating pressure and verifying that either:
- a. The turbine bypass valve(s) indicate a compensating valve movement, or
- b. The reactor coolant system pressure decreases by an amount equivalent to the valve pressur e relieving capacity for the test conditions.
um er o A va ves to be consistent with ECCS analysis.
EMERGENCY CORE COOLING SYSTEMS AUTOMATIC DEPRESSUR IZATION SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
- 2. Conducting a visual inspection of the safety-relief and relief valve line restraints in the torus to verify structural integrity for continued operation.
- c. After March 1, 1979, by performance of the following test program:
- l. Manually opening each ADS valve in accordance with the test schedule of Table 4.4-10 with the reactor at or below 5% rated power and at nominal operating pressure and verifying that either:
- a. The turbine ".ypass valve(s) indicate a compensating valve movement, or
- b. The reactor coolant system pressure decreases by an amount equivalent to the valve pressure relieving capacity for the test conditions.
- 2. The initial Next Required Test Interval of Table 4.4-10 shall be determined by the number of remotely operated relief and safety-relief valves found inoperable from September 1, 1977 to March 1, 1979.
- 3. The initial valve tests of Table 4.4-10 shall be completed by, the earlier of:
- a. The completion of the next refueling outage occur ring after March 1, 1979, or
- b. The time period defined by March 1, 1979 plus the initial te'st interval, determined above.
- 4. At least once per 18 months by conducting a visual inspection of the safety-relief and relief valve line restraints in the torus to verify structural integrity for continued operation.
3/4.5 EMERGENCY CORE COOLING SYSTEM BASES 3/4.5.2 AUTOMATIC DEPRESSURIZATION SYSTEM (ADS)
Upon failure of the HPCIS to function properly after a small break loss-of-coolant accident, the ADS automatically causes the safety-relief valves to open, depressurizing the reactor so that flow from the low pressure cooling systems can enter the core in time to limit fuel cladding temperature to less than 2200'F. ADS is conservatively required to be operable whenever .reactor vessel pressure exceeds ( 150) psig even though low pressure cooling systems provide adequate core cooling up to (350) psig.
ADS automatically controls (7 ) safety-relief valves although the safety analysis only takes credit for (6 ). Therefore it is appropriate to permit (one) valve to be out-of-service without materially reducing system reliability.
The testing frequency applicable to ADS valves is provided to ensure operability and demonstrate reliability of the valves. The required .test',ng interval varies with observed valve failures. The
,number of inoperable valves found during both operation and testing of these valves determines the time interval for the next required test of these valves. Early tests may be performed prior to entering the next required test interval (i.e., in advance of the nominal time less the negative 25% tolerance band). Early tests may be used as a new reference point for tests of the same time interval, however, they are not acceptable for lengthening the test interval since they were not performed within the +25% tolerance band as required by Table 4.4-10.
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0 Distribution
~~Docket ORB ¹3 Local PDR
~4l)'g 0 ig7 Docket No. 50-220 NRC PDR GLear CParrish JZwetzig Niagara flohawk Power Corporation SNowicki ATTN: tfr. Gerald K. Rhode Attorney, OELD Vice President - Engineering OI8E (3) 300 Erie Boulevard West DEisenhut Syracuse, Nevi York 13202 TBAbernathy JRBuchanan Gentlemen: ACRS (16)
RE: NINE MILE POINT NUCLEAR STATION UNIT NO. 1 Our review of data received from reactor vessel material surveillance programs indicates that the materials used in reactor vessel fabrication may have a wider variation fn sensitivity to radiation damage than originally anticipated. In addition, some reactor vessels incorporate more than one heat of materials, including vleld metals fn their beltlfne regions, but all of these heats may not be included in the reactor vessel material surveillance program.
Although our review of these data does not reveal a basis for concern regarding continued reactor vessel integrity over the next several years, the information does indicate the need for a detailed review of the materials employed in reactor vessel construction (fn light of this recent data) and a review of the specimens employed in the surveillance program to determine ff the present specimens reasonably represent the limiting materials fn the .reactor vessel beltlfne region.
In order to perform these reviews, we will need the information listed fn the enclosure relative to each of your reactor vessel(s) and assocfated surveillance specimens.
Accordingly, you are requested to supply one signed orfgfnal and 39 copies of the information listed in the enclosure within60 days of receipt of thi s letter.
This. request for generic information was approved by GAO under a blanket clearance number 8-180225 (R0072); this clearance expires July 31, 1977.
Sincerely, Original sipned by George Lear, Chief Ooeratin Reactors 8 nch ¹3 of ficc3a-Divis on op Opera) ORB ¹3 SURNAMCQ DATC~
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