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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.50-220NiagaramohawkPowerCorporation ATTN:Nr.GeraldK.RhodeVicePresident | | {{#Wiki_filter:= |
| -Engineering 300ErieBoulevard WestSyracuse, NewYork13202Gentlemen: | | 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:NINENILEPOINTNUCLEARSTATIOflUNITNO.1Distribution DocketORB83CParrishGLearSNowickiDVerrelli LocalPDRNRCPDRAttorney, OELDOI&E(3)DEisenhut TBAbernathy NRBuchanan ACRS(16)Inthepastseveralyears,asignificant numberofreliefvalvesandsafety-relief valveswerefoundtobeinoperable atBWRreactorfacili-ties.Thesevalveswereinstalled intheReactorCoolantSystemand/orAutomatic Depressurization System.Severalprogramshavebeendeveloped toreducetheincidence ofthesevalvefailures; however,additional failurescontinuetooccur.Consequently, wehaveconcluded thatchangestotheSurveillance Require-mentsandLimitingConditions forOperations forallBHR'sareneededtoprovideadditional assurance ofreliefvalveandsafety-relief valveoperabflity andreliability.
| | 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. |
| Therefore, werequestthatyoumodifyyoursurveillance testingprogramthroughtheadoptionoftheprogramcontained inthemodeltechnical specifications wehaveprepared. | | 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: |
| Theelementsofthisprograminclude:Eachremotlyoperatedreliefvalveandsafety-relief valvefntheReactorCoolantSystemandAutomatic Depressurizatfon Systemwillbetestedonavariablefrequency schedulerelatedtodemonstrated reliability andoperability.
| | 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. |
| Thetestingintervalisbasedonthenumberofvalvefailuresduringtherequiredtestinterval.
| | : 2. The increased surveillance program will become effective on Harch 1, 19T9. No increase in valve, testing is required before that date. |
| Facilities withreliablevalveswillprogresstoalongertestintervalwhilethosewithvalvefailureswillprogresstoashortertestinterval. | | 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~ |
| Thisconceptshouldresultinthemaintenance ofamoreuniformlevelofreliability forthisequipment thanpreviously obtained.
| | SUANAMEW DATE~ |
| 2.Theincreased surveillance programwillbecomeeffective onHarch1,19T9.Noincreaseinvalve,testingisrequiredbeforethatdate.Theinitialtestinginterval. | | NRC FORM'318 (9.76) NRCM 0240 4V 8 OOVEANMENT PAINTIN4 OFFICER 1024 420 624 |
| oftheincreased surveillance programwillbebasedonthenumberofremotelyoperatedreliefvalvesandsafety-relief valvesfoundinoperable intheprevious18monthspP'September 1,1977toMarch1,1979).Thisleadtimewillpermittheresolution oftheHarkISafety-Relief ValveLoadsandDFPICE~SUANAMEWDATE~NRCFORM'318(9.76)NRCM02404V8OOVEANMENT PAINTIN4OFFICER1024420624 4Fhhr1*NI(I/'('I~fI)'''CItAIhrortSI*.~rl1lfESIih"11('",fhh"Ih(7.'''),rh,firIh,INN,h(It~Nttr r~pNiagaraHohawkPowerCorporation p,UG8197)timeissufficient topermitthedevelopment andimplementation ofimprovedsafetyandsafety-relfef valvemaintenance procedures and'othercorrective actionspriortoimplementing thetestprogram.3.Thereliefand/orsafety-relief valvelinerestraints inthetoruswillbeexaminedpriortoinitiating thetestprogramandatleastonceeachfuelcycle(f.e.,each18months)toverifycontinued structural integrity.
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| Werequestthatyousubmitwithin30daysfromyourreceiptofthis'letter,anapplication foramendment toyourlicensethatwillchangeyourtechnical specifications tobeinconformance wfththerequirements oftheenclosedmodeltechnical specifications andassociated'bases.
| | 4 F |
| Intheeventyoushoulddesirefurtherdfscussion ofthismatter,pleasecontactus.Sincerely, OriginalsignedbV
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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:== |
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| ModelTechnical Specifications GeorgeLear,ChiefOperating ReactorsBranch83DivisionofOperating ReactorsCC:ArvinE.Upton,EsquireLeBoeuf,Lamb,Lefby5NacRae1757NStreet,H.W.Washington, D.C.20036AnthonyZ.Rofsman,EsquireRofsman,KesslerandCashdan102515thStreet,N.H.5thFloorWashington, D.C.20005Nr.EugeneG.Saloga,Applicant Coordinator
| | Model Technical Specifications CC: |
| .,NineNilePointEnergyInformation CenterP.0.Box81Lycoming, Nes(York13093OFFICS~SU/NAME&DATE~ORB3CParrish8/1/77ORBk3SNowickjf8/)/77ORBii3jDV~rlI8//77,ORB83GLearRS(S77'RCFORM318(9-76)NRCM0240AU,'0,OOVSANMSNT FIIINTINO OFFICS<ISTS02~24 | | Arvin E. Upton, Esquire LeBoeuf, Lamb, Lefby 5 NacRae 1757 N Street, H. W. |
| <<FIF~r'F<I"I1I1<~I~k~rlrF<k<<<l<<'Ih<IIYll<<'tj'l~f'1.,Fkrl<<'.~W REACTORCOOLANTSYSTEM3/4.4.2SAFETYVALVESLIMITINGCONDITION FOROPERATION 3.4.2Atleastthefollowing reactorcoolantsystemcodesafetyvalvesandsafety-relief valvesshallbeoperablewithliftsettingswithin+1%oftheindicated pressures.
| | Washington, D. C. 20036 Anthony Z. Rofsman, Esquire Rofsman, Kessler and Cashdan 1025 15th Street, N. H. |
| (2)*Safetyvalves9(1240)psig(3)Safety-relief valves9(1100)psig(3)Safety-relief valves9(1090)psig(3)Safety-relief valves9(1080)psigAPPLICABILITY:
| | 5th Floor Washington, D. C. 20005 Nr. Eugene G. Saloga, Applicant Coordinator |
| WithAverageCoolantTemperature
| | ., Nine Nile Point Energy Information Center P. 0. Box 81 Lycoming, Nes( York 13093 ORB 3 ORB k3 ORB ii3 j ORB 83 OFFICS~ |
| >212'FortheModeSwitchinRun,orStartup/Hot Standby.ACTION:Withoneormorereactorcoolantsystemcodesafetyvalve(s)orasafety-relief valve(s)inoperable eitherrestorethevalve(s)tooperablestatuswithin15minutesorbe.shutdownwithin12hoursandreduceAverageCoolantTemperature to<212'Fwithinthe next24hours..SURVEILLANCE REUIREMENTS 4.4.2.1a.Inadditiontotheapplicable ASMEBoilerandPressureVesselCode,SectionXIrequirements, eachsafety-relief valveshallbedemonstrated operable:
| | SU/NAME& CParrish SNowick jf DV ~rl I GLear DATE~ |
| Atleastonceper24hours,byverifying bellowsintegrity throughinstrument indication.
| | 8/1 /77 8/) /77 8/ /77, RS (S 77 |
| b.UntilMarch1,1979,atleastonceper18monthsby:1.Manuallyopeningeachremotelyoperated.safety-relief valvewiththereactoratorbelow5%ratedpowerandatnominaloperating
| | 'RC FORM 318 (9-76) NRCM 0240 A U,'0, OOVSANMSNT FIIINTINO OFFICS< ISTS 02~24 |
| : pressure, andverifying thateither:a.Theturbinebypassvalve(s)indicateacompensating valvemovement, orb.Thereactorcoolantsystempressuredecreases byanamountequivalent tothevalvepressurerelieving capacityforthetestconditions.
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| REACTORCOOLANTSYSTEMSURVEILLANCE REUIREMENTS (Continued) 2.Conducting avisualinspection ofthesafety-relief valvelinerestraints inthetorustoverifystructural integrity forcontinued operation.
| | 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. |
| c.AfterMarch1,1979,byperformance ofthefollowing testprogram:1.Manuallyopeningeachremotelyoperatedsafety-relief valveinaccordance withthetestscheduleofTable4.4-10withthereactoratorbelow5Xratedpowerandasteamatnominaloperating pressureandverifying thateither:a.Theturbinebypassvalve(s)indicateacompensating valvemovement, orb.Thereactorcoolantsystempressuredecreases byanamountequivalent tothevalvepressurerelieving capacityforthetestconditions.
| | 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. |
| 2.TheinitialNextRequiredTestIntervalofTable4.4-10shallbedetermined bythenumberofremotelyoperatedreliefandsafety-relief valvesfoundinoperable fromSeptember 1,1977toMarch1,1979.3.TheinitialvalvetestsofTable4.4-10shallbecompleted by,theearlierof:a.Thecompletion ofthenextrefueling outageoccurring afterMarch1,1979,orb.ThetimeperioddefinedbyMarch1,1979plustheinitialtestinterval, determined above.4.Atleastonceper18months,byconducting avisualinspection ofthesafety-relief valvelinerestraints inthetorustoverifystructural integrity forcontinued operation. | | ACTION: |
| 4.4.2.2Eachsafetyvalveandthesafetyvalvefunctionofeachsafety-reliefvalveshallbedemonstrated operablepertherequirements oftheASMEBoilerandPressureVesselCode()EditionandAddendathrough().V
| | 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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| TABLE4.4-10REHOTELYOPERATEDRELIEFANDSAFETY-RELIEF VALVETESTSCHEDULENUHBEROFREHOTELYOPERATEDRELIEFANDSAFETY-RELIEF VALVESFOUNDINOPERABLE DURINGTESTINGORTESTINTERVAL**
| | REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS (Continued) |
| NEXTREQUIREDTESTINTERVAL*
| | : 2. Conducting a visual inspection of the safety-relief valve line restraints in the torus to verify structural integrity for continued operation. |
| 012318months+25%184days+25K92days+25%31days+25%"Therequiredtestintervalshallnotbelengthened morethanonestepatatime.Earlytestsmaybeperformed priortoenteringthe"nextrequiredtestinterval" (i.e.,inadvanceofthenominaltimelessthenegative25Ktolerance band).Earlytestsmaybeusedasanewreference pointfortestsofthesameinterval, however,theyarenotacceptable forlengthening thetestinterval.
| | : c. After March 1, 1979, by performance of the following test program: |
| **Setpoint-driftisnotconsidered tobeavalvefailureforthepurposesofthistestschedule.
| | : 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: |
| I 3/4.4REACTORCOOLANTSYSTEMBASES3/4.4.2SAFETYVALVESThereactorcoolantsystemsafetyvalvesoperatetopreventthereactorcoolantsystemfrombeingpressurized abovetheSafetyLimitofpsig.Eachsafetyvalveisdesignedtorelieveibsperhooratthevalvesetpoint.ThesystemisdesignedtomeettheASMEBoilerandPressureVesselCoderequirements thatthenuclearsystemreliefvalvesshal=lfunctiontopreventopeningofthesafetyvalves.Althoughthesafetyvalvefunctionisnotexpectedtoberequiredunderthemostlimitingtransient, aninoperable valverequiresshutdowninordertocomplywithASMECoderequirements.
| | : a. The turbine bypass valve(s) indicate a compensating valve movement, or |
| Thetestingfrequency applicable tothereliefvalvefunctionofthesafety-relief valvesisprovidedtoensureoperability anddemonstrate reliability ofthevalves.Therequiredtestingintervalvarieswithobservedvalvefailures.
| | : b. The reactor coolant system pressure decreases by an amount equivalent to the valve pressure relieving capacity for the test conditions. |
| Thenumberofinoperable valvesfoundduringbothoperation andtestingofthesevalvesdetermines thetimeintervalforthenextrequiredtestofthesevalves.Earlytestsmaybeperformed priortoenteringthenextrequiredtestinterval(i.e.,inadvanceofthenominaltimelessthenegative25%tolerance band).Earlytestsmaybeusedasanewreference pointfortestsofthesametimeinterval, however,theyarenotacceptable forlengthening thetestintervalsincetheywerenotperformed withinthe+25%tolerance bandasrequiredbyTable4.4-10.Oemonstration ofthesafetyvalves'ift settingswilloccuronlyduringshutdownandwillbeperformed inaccordance withtheprovisions ofSectionXIoftheASMEBoilerandPressureVesselCode.
| | : 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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| EMERGENCY CORECOOLINGSYSTEMSAUTOMATIC DEPRESSURIZATION SYSTEMLIMITINGCONDITION FOROPERATION 3.5.2TheAutomatic Depressurization System(ADS)shallbeOPERABLEwithatleast(6)*OPERABLEADSvalves.APPLICABILITY:
| | 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* |
| WithAverageCoolantTemperature
| | 0 18 months + 25% |
| >212'FortheModeSwitchinRun,orStartup/Hot Standby.ACTION:a.WithoneoftheaboverequiredADSvalvesinoperable, operation maycontinueprovidedtheactuation logicoftheremaining ADSvalvesisoperableandtheCSSandLPCIsystemsareoperable, andtheHPCIsystemisdemonstrated operablewithin4hours;restoretheinoperable ADSvalvetooperablestatuswit'iin14daysorbeshutdownwithin12hoursandreducetheAverageCoolantTemperature to<212'Fwithinthefollowing 24hours.b.WithtwoormoreoftheaboverequiredADSvalvesinoperable, beshutdownwithin12hoursandreducetheAverageCoolantTemperature to<212'Fwithinthefollowing 24hours.SURVEILLANCE REUIREMENTS 4.5.2Inadditiontotheapplicable ASMEBoilerandPressureVesselCode,SectionXIrequirements, theADSshallbedemonstrated operable:
| | 1 184 days + 25K 2 92 days + 25% |
| a.Atleastonceper18monthsbyperformance ofasystemfunctional testwhichincludessimulated automatic actuation throughtheautomatic depressurization | | 3 31 days + 25% |
| : sequence, butexcluding valveactuation.
| | "The required test interval shall not be lengthened more than one step at a time. |
| b.UntilMarch1,1979,atleastonceper18monthsby:1.ManuallyopeningeachADSvalvewiththereactoratorbelow5Xratedpowerandatnominaloperating pressureandverifying thateither:a.Theturbinebypassvalve(s)indicateacompensating valvemovement, orb.Thereactorcoolantsystempressuredecreases byanamountequivalent tothevalvepressurerelieving capacityforthetestconditions.
| | 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). |
| umeroAvavestobeconsistent withECCSanalysis.
| | 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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| EMERGENCY CORECOOLINGSYSTEMSAUTOMATIC DEPRESSUR IZATIONSYSTEMSURVEILLANCE REQUIREMENTS (Continued) 2.Conducting avisualinspection ofthesafety-relief andreliefvalvelinerestraints inthetorustoverifystructural integrity forcontinued operation. | | 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. |
| c.AfterMarch1,1979,byperformance ofthefollowing testprogram:l.ManuallyopeningeachADSvalveinaccordance withthetestscheduleofTable4.4-10withthereactoratorbelow5%ratedpowerandatnominaloperating pressureandverifying thateither:a.Theturbine".ypassvalve(s)indicateacompensating valvemovement, orb.Thereactorcoolantsystempressuredecreases byanamountequivalent tothevalvepressurerelieving capacityforthetestconditions. | | 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. |
| 2.TheinitialNextRequiredTestIntervalofTable4.4-10shallbedetermined bythenumberofremotelyoperatedreliefandsafety-relief valvesfoundinoperable fromSeptember 1,1977toMarch1,1979.3.TheinitialvalvetestsofTable4.4-10shallbecompleted by,theearlierof:a.Thecompletion ofthenextrefueling outageoccurringafterMarch1,1979,orb.ThetimeperioddefinedbyMarch1,1979plustheinitialte'stinterval, determined above.4.Atleastonceper18monthsbyconducting avisualinspection ofthesafety-relief andreliefvalvelinerestraints inthetorustoverifystructural integrity forcontinued operation. | | 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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| 3/4.5EMERGENCY CORECOOLINGSYSTEMBASES3/4.5.2AUTOMATIC DEPRESSURIZATION SYSTEM(ADS)-UponfailureoftheHPCIStofunctionproperlyafterasmallbreakloss-of-coolant
| |
| : accident, theADSautomatically causesthesafety-relief valvestoopen,depressurizing thereactorsothatflowfromthelowpressurecoolingsystemscanenterthecoreintimetolimitfuelcladdingtemperature tolessthan2200'F.ADSisconservatively requiredtobeoperablewhenever.reactorvesselpressureexceeds(150)psigeventhoughlowpressurecoolingsystemsprovideadequatecorecoolingupto(350)psig.ADSautomatically controls(7)safety-relief valvesalthoughthesafetyanalysisonlytakescreditfor(6).Therefore itisappropriate topermit(one)valvetobeout-of-service withoutmaterially reducingsystemreliability.
| |
| Thetestingfrequency applicable toADSvalvesisprovidedtoensureoperability anddemonstrate reliability ofthevalves.Therequired.test',ng intervalvarieswithobservedvalvefailures.
| |
| The,numberofinoperable valvesfoundduringbothoperation andtestingofthesevalvesdetermines thetimeintervalforthenextrequiredtestofthesevalves.Earlytestsmaybeperformed priortoenteringthenextrequiredtestinterval(i.e.,inadvanceofthenominaltimelessthenegative25%tolerance band).Earlytestsmaybeusedasanewreference pointfortestsofthesametimeinterval, however,theyarenotacceptable forlengthening thetestintervalsincetheywerenotperformed withinthe+25%tolerance bandasrequiredbyTable4.4-10.
| |
| ~- | | ~- |
| 0DocketNo.50-220~4l)'g0ig7NiagaraflohawkPowerCorporation ATTN:tfr.GeraldK.RhodeVicePresident
| | 0 Distribution |
| -Engineering 300ErieBoulevard WestSyracuse, NeviYork13202Gentlemen: | | ~~Docket ORB ¹3 Local PDR |
| RE:NINEMILEPOINTNUCLEARSTATIONUNITNO.1Distribution
| | ~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) |
| ~~DocketORB¹3LocalPDRNRCPDRGLearCParrishJZwetzigSNowickiAttorney, OELDOI8E(3)DEisenhut TBAbernathy JRBuchanan ACRS(16)Ourreviewofdatareceivedfromreactorvesselmaterialsurveillance programsindicates thatthematerials usedinreactorvesselfabrication mayhaveawidervariation fnsensitivity toradiation damagethanoriginally anticipated.
| | 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. |
| Inaddition, somereactorvesselsincorporate morethanoneheatofmaterials, including vleldmetalsfntheirbeltlfneregions,butalloftheseheatsmaynotbeincludedinthereactorvesselmaterialsurveillance program.Althoughourreviewofthesedatadoesnotrevealabasisforconcernregarding continued reactorvesselintegrity overthenextseveralyears,theinformation doesindicatetheneedforadetailedreviewofthematerials employedinreactorvesselconstruction (fnlightofthisrecentdata)andareviewofthespecimens employedinthesurveillance programtodetermine ffthepresentspecimens reasonably represent thelimitingmaterials fnthe.reactorvesselbeltlfneregion.Inordertoperformthesereviews,wewillneedtheinformation listedfntheenclosure relativetoeachofyourreactorvessel(s) andassocfated surveillance specimens.
| | 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. |
| 'Accordingly, youarerequested tosupplyonesignedorfgfnaland39copiesoftheinformation listedintheenclosure within60 daysofreceiptofthisletter.This.requestforgenericinformation wasapprovedbyGAOunderablanketclearance number8-180225(R0072);thisclearance expiresJuly31,1977.Sincerely, OriginalsipnedbyGeorgeLear,ChiefOoeratinReactors8nch¹3officc3a-losureand'cc:nextpageSURNAMCQSeDATC~lormhEC-318(Rev.9-53)AKCM0240DivisonopOpera)ORB¹3~owicki-----....5/..../7.7............
| | 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. |
| ORB%~--JZwetzig'----.....5l..g./.7.7............,...
| | 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. |
| ORB¹3~--CPaZr1sfAmj-5/..5/.7.7.............,
| | This. request for generic information was approved by GAO under a blanket clearance number 8-180225 (R0072); this clearance expires July 31, 1977. |
| QU,4,ODVCANMCNT fNINTIN4OffCCI1474S20eSSORB¹3BLeer~.ah.a.7.............
| | Sincerely, Original sipned by George Lear, Chief Ooeratin Reactors 8 nch ¹3 of ficc3a-Divis on op Opera) ORB ¹3 SURNAMCQ DATC~ |
| NI1IkPI}}
| | 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 |
| | |
| | N I |
| | 1 I k P I}} |
|
|---|
Category:Letter
MONTHYEARML24268A3382024-10-16016 October 2024
Issuance of Amendment No. 253 Regarding the Modification of TS Surveillance Requirement 4.3.6.a Related to Adoption of TSTF-425, Revision 3
RS-24-093, Response to Request for Additional Information - Alternative Request to Utilize Code Case OMN-32, Alternative Requirements for Range and Accuracy of Pressure, Flow, and Differential Pressure Instruments Used in Pump Tests2024-10-10010 October 2024
Response to Request for Additional Information - Alternative Request to Utilize Code Case OMN-32, Alternative Requirements for Range and Accuracy of Pressure, Flow, and Differential Pressure Instruments Used in Pump Tests
NMP2L2890, Submittal of Revision 26 to the USAR and Reference Figures, 10 CFR 50.59 Evaluation Summary Report, TS Bases, TRM Requirements Manual Changes, and 10 CFR 54.37(b) Aging Management Review (Excludes Attachment 6)2024-10-0404 October 2024
Submittal of Revision 26 to the USAR and Reference Figures, 10 CFR 50.59 Evaluation Summary Report, TS Bases, TRM Requirements Manual Changes, and 10 CFR 54.37(b) Aging Management Review (Excludes Attachment 6)
IR 05000220/20243022024-10-0303 October 2024
Initial Operator Licensing Examination Report 05000220/2024302
ML24275A2442024-10-0303 October 2024
Reassignment of the U.S. Nuclear Regulatory Commission Branch Chief, Division of Operating Reactor Licensing
ML24190A0012024-09-26026 September 2024
Issuance of Amendment Nos. 252 and 197 Regarding the Revision to Technical Specification Design Features Section to Remove Nine Mile Point Unit 3 Project Designation
NMP1L3608, Supplemental Information Letter No. 3 - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation2024-09-20020 September 2024
Supplemental Information Letter No. 3 - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation
RS-24-090, Response to Request for Additional Information - Relief Request Concerning Extension of Permanent Relief from Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds2024-09-12012 September 2024
Response to Request for Additional Information - Relief Request Concerning Extension of Permanent Relief from Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds
ML24249A1362024-09-0404 September 2024
EN 57304 - Westinghouse Electric Company, LLC, Final Report - No Embedded Files. Notification of the Potential Existence of Defects Pursuant to 10 CFR Part 21
IR 05000220/20240052024-08-29029 August 2024
Updated Inspection Plan for Nine Mile Point Nuclear Station, Units 1 and 2 (Report 05000220/2024005 and 05000410/2024005)
IR 05000220/20240102024-08-22022 August 2024
Age-Related Degradation Inspection Report 05000220/2024010 and 05000410/2024010
NMP1L3603, Submittal of Preliminary Decommissioning Cost Estimate and Irradiated Fuel Management Plan2024-08-20020 August 2024
Submittal of Preliminary Decommissioning Cost Estimate and Irradiated Fuel Management Plan
ML24222A6772024-08-0909 August 2024
Response to Request for Additional Information for Application to Revise Technical Specifications to Adopt TSTF-591-A, Revise Risk Informed Completion Time (RICT) Program Revision 0 and Revise 10 CFR 50.69 License Condition
IR 05000220/20240022024-08-0505 August 2024
Integrated Inspection Report 05000220/2024002 and 05000410/2024002
ML24215A3002024-08-0202 August 2024
Operator Licensing Examination Approval
ML24213A1412024-07-31031 July 2024
Requalification Program Inspection
NMP1L3601, Supplemental Information Letter No. 2 - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation2024-07-31031 July 2024
Supplemental Information Letter No. 2 - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation
NMP2L2883, Fourth Inservice Inspection Interval, Second Inservice Inspection Period 2024 Owner’S Activity Report for RFO-19 Inservice Examinations2024-07-24024 July 2024
Fourth Inservice Inspection Interval, Second Inservice Inspection Period 2024 Owner’S Activity Report for RFO-19 Inservice Examinations
ML24198A0852024-07-16016 July 2024
Senior Reactor and Reactor Operator Initial License Examinations
RS-24-070, Independent Spent Fuel Storage Installation, Nine Mile Point, Units 1 and 2, Quad Cities, Units 1 and 2, R. E. Ginna - Nuclear Radiological Emergency Plan Document Revisions2024-07-12012 July 2024
Independent Spent Fuel Storage Installation, Nine Mile Point, Units 1 and 2, Quad Cities, Units 1 and 2, R. E. Ginna - Nuclear Radiological Emergency Plan Document Revisions
RS-24-061, Constellation Energy Generation, LLC, Response to NRC Regulatory Issue Summary 2024-01, Preparation and Scheduling of Operator Licensing Examinations2024-06-14014 June 2024
Constellation Energy Generation, LLC, Response to NRC Regulatory Issue Summary 2024-01, Preparation and Scheduling of Operator Licensing Examinations
NMP1L3584, License Amendment Request to Revise Technical Specifications to Adopt TSTF-230, Revision 1, Add New Condition B to LCO 3.6.2.3, RHR Suppression Pool Cooling2024-06-13013 June 2024
License Amendment Request to Revise Technical Specifications to Adopt TSTF-230, Revision 1, Add New Condition B to LCO 3.6.2.3, RHR Suppression Pool Cooling
IR 05000220/20244012024-05-30030 May 2024
Security Baseline Inspection Report 05000220/2024401 and 05000410/2024401(Cover Letter Only)
ML24079A0762024-05-23023 May 2024
Issuance of Amendments to Adopt TSTF 264
NMP1L3591, Response to Ny State Pollutant Discharge Elimination System (SPDES) Permit Request for Information & Modification Request2024-05-18018 May 2024
Response to Ny State Pollutant Discharge Elimination System (SPDES) Permit Request for Information & Modification Request
NMP1L3589, Special Report: Containment High Range Radiation Monitor Instrumentation Channel 12 Inoperable2024-05-16016 May 2024
Special Report: Containment High Range Radiation Monitor Instrumentation Channel 12 Inoperable
NMP1L3582, 2023 Annual Radioactive Environmental Operating Report for Nine Mile Point Units 1 and 22024-05-15015 May 2024
2023 Annual Radioactive Environmental Operating Report for Nine Mile Point Units 1 and 2
ML24158A2052024-05-15015 May 2024
Annual Radioactive Environmental Operating Report
IR 05000220/20240012024-05-10010 May 2024
Integrated Inspection Report 05000220/2024001 and 05000410/2024001
RS-24-049, Updated Notice of Intent to Pursue Subsequent License Renewal Applications2024-05-0909 May 2024
Updated Notice of Intent to Pursue Subsequent License Renewal Applications
RS-24-038, Relief Request Concerning Extension of Permanent Relief from Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds2024-05-0202 May 2024
Relief Request Concerning Extension of Permanent Relief from Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds
05000410/LER-2024-001, Automatic Reactor Scram on Turbine Trip Due to Low Condenser Vacuum2024-05-0101 May 2024
Automatic Reactor Scram on Turbine Trip Due to Low Condenser Vacuum
NMP1L3581, Independent Spent Fuel Storage Installation (ISFSI) - 2023 Radioactive Effluent Release Report2024-04-30030 April 2024
Independent Spent Fuel Storage Installation (ISFSI) - 2023 Radioactive Effluent Release Report
RS-24-041, Alternative Request to Utilize Code Case OMN-32, Alternative Requirements for Range and Accuracy of Pressure, Flow, and Differential Pressure Instruments Used in Pump Tests2024-04-30030 April 2024
Alternative Request to Utilize Code Case OMN-32, Alternative Requirements for Range and Accuracy of Pressure, Flow, and Differential Pressure Instruments Used in Pump Tests
NMP2L2877, 2023 Annual Environmental Operating Report2024-04-19019 April 2024
2023 Annual Environmental Operating Report
NMP2L2878, Core Operating Limits Report2024-04-16016 April 2024
Core Operating Limits Report
ML24103A2042024-04-12012 April 2024
Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-591-A, Revise Risk Informed Completion Time (RICT) Program Revision 0 and Revise 10 CFR 50.69 License Condition
ML24092A3352024-04-0101 April 2024
NRC Office of Investigations Case No. 1-2023-002
RS-24-002, Constellation Energy Generation, LLC - Annual Property Insurance Status Report2024-04-0101 April 2024
Constellation Energy Generation, LLC - Annual Property Insurance Status Report
ML24074A2812024-03-14014 March 2024
Request for Information and Notification of Conduct of IP 71111.21.N.04, Age-Related Degradation, Reference Inspection Report 05000220/2024010 and 05000410/2024010
NMP1L3577, Special Report: Containment High Range Radiation Monitor Instrumentation Channel 12 Inoperable2024-03-13013 March 2024
Special Report: Containment High Range Radiation Monitor Instrumentation Channel 12 Inoperable
IR 05000220/20230062024-02-28028 February 2024
Annual Assessment Letter for Nine Mile Point Nuclear Station, Units 1 and 2, (Reports 05000220/2023006 and 05000410/2023006)
NMP1L3570, 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
05000410/LER-2023-001, Supplement to LER 2023-001-00, Automatic Reactor Scram on Low Level Due to Partial Loss of Feedwater2024-01-30030 January 2024
Supplement to LER 2023-001-00, Automatic Reactor Scram on Low Level Due to Partial Loss of Feedwater
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))
05000220/LER-2023-002, Average Power Range Monitors Declared Inoperable Due to Trip of Reactor Recirculation Pump 122023-12-15015 December 2023
Average Power Range Monitors Declared Inoperable Due to Trip of Reactor Recirculation Pump 12
IR 05000410/20243012023-12-14014 December 2023
Initial Operator Licensing Examination Report 05000410/2024301
2024-09-04
[Table view]
Category:Request for Additional Information (RAI)
MONTHYEARML24312A1662024-11-0707 November 2024
Request for Additional Information (11/7/2024 E-mail) - LAR to Revise TSs to Adopt TSTF-230, Revision 1, Add New Condition B to LCO 3.6.2.3
ML24197A0162024-07-12012 July 2024
NRR E-mail Capture - Final RAI - Constellation Energy Generation, LLC - Fleet Request - License Amendment Request to Adopt TSTF-591
ML24074A2812024-03-14014 March 2024
Request for Information and Notification of Conduct of IP 71111.21.N.04, Age-Related Degradation, Reference Inspection Report 05000220/2024010 and 05000410/2024010
ML23264A7992023-09-21021 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
ML20213A9352020-07-30030 July 2020
Request for Additional Information Review of License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times
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)
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)
ML18067A1482018-03-0808 March 2018
Enclosurequest for Additional Information (Letter to P. R. Simpson Request for Additional Information Regarding Exelon Generating Company, Llc'S Decommissioning Funding Plan Update for Independent Spent Fuel Storage Installation)
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
2024-07-12
[Table view]
Category:Technical Specification
MONTHYEARNMP2L2890, Submittal of Revision 26 to the USAR and Reference Figures, 10 CFR 50.59 Evaluation Summary Report, TS Bases, TRM Requirements Manual Changes, and 10 CFR 54.37(b) Aging Management Review (Excludes Attachment 6)2024-10-0404 October 2024
Submittal of Revision 26 to the USAR and Reference Figures, 10 CFR 50.59 Evaluation Summary Report, TS Bases, TRM Requirements Manual Changes, and 10 CFR 54.37(b) Aging Management Review (Excludes Attachment 6)
NMP1L3601, Supplemental Information Letter No. 2 - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation2024-07-31031 July 2024
Supplemental Information Letter No. 2 - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation
ML24103A2042024-04-12012 April 2024
Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-591-A, Revise Risk Informed Completion Time (RICT) Program Revision 0 and Revise 10 CFR 50.69 License Condition
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
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
ML16309A4052016-10-24024 October 2016
Revision 47 to Technical Specification Bases, Section B3.2, Power Distribution Limits
ML16309A4012016-10-24024 October 2016
Revision 47 to Technical Specification Bases, Table of Contents
ML16309A4132016-10-24024 October 2016
Revision 47 to Technical Specification Bases, Section B3.9, Refueling Operations
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
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
2024-07-31
[Table view]
Category:Amendment
MONTHYEARML24103A2042024-04-12012 April 2024
Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-591-A, Revise Risk Informed Completion Time (RICT) Program Revision 0 and Revise 10 CFR 50.69 License Condition
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
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 ..
2024-04-12
[Table view] |
Text
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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~
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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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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