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{{#Wiki_filter:BEFORETHEUNXTEDSTATESNUCLEARREGULATORY COMMISSSXON IntheMatterofROCHESTER GASANDELECTRICCORPORATION (R.E.GinnaNuclearPowerStation,UnitNo.1)CERTIFICATE OFSERVICEDocketNo.52-244IherebycertifythatIhaveservedadocumententitled"Application forAmendment toOperating License"withthree(3)documents, Attachments A,B,andC,attachedthereto,bymailingcopiesthereoffirstclass,postagepre-paid,toeach,ofthefollowing personsthis3rddayofAugust,1979:Mr.MichaelL.Slade12Trailwood CircleRochester, NewYork14618WarrenB.Rosenbaum, Esq.OneMainStreet707WilderBuildingRochester, NewYork14614EdwardG.Ketchen,Esp.OfficeoftheExecutive LegalDirectorU.S.NuclearRegulatory Commission Washington, D.C.20555Mr.RobertN.PickneySupervisor TownofOntario107RidgeRoadWestOntario,NewYork14519V908080Q/(p'-
{{#Wiki_filter:BEFORE THE UNXTED STATES NUCLEAR REGULATORY COMMISSSXON In the Matter of ROCHESTER GAS AND ELECTRIC CORPORATION (R.E.Ginna Nuclear Power Station, Unit No.1)CERTIFICATE OF SERVICE Docket No.52-244 I hereby certify that I have served a document entitled"Application for Amendment to Operating License" with three (3)documents, Attachments A, B, and C, attached thereto, by mailing copies thereof first class, postage pre-paid, to each, of the following persons this 3rd day of August, 1979: Mr.Michael L.Slade 12 Trailwood Circle Rochester, New York 14618 Warren B.Rosenbaum, Esq.One Main Street 707 Wilder Building Rochester, New York 14614 Edward G.Ketchen, Esp.Office of the Executive Legal Director U.S.Nuclear Regulatory Commission Washington, D.C.20555 Mr.Robert N.Pickney Supervisor Town of Ontario 107 Ridge Road West Ontario, New York 14519 V 908080 Q/(p'-
JeffreyL.Cohen,Esq.NewYorkStateEnergyOfficeSwanStreetBuildingCore1,SecondFloorEmpireStatePlazaAlbany,NewYorkl2223EdwardLuton,Esq.AtomicSafetyandLicensing BoardU.S.NuclearRegulatory CommissionWashington, D.C.20555Dr.EmmethA.LuebkeAtomicSafetyandLicensing BoardU.S.NuclearRegulatory CommissionWashington, D.C.20555Dr.DixonCallihanUnionCarbideCorporation P.O.BoxYOakRidge,Tennessee 37830LxK.LarsonLeBoeuf,Lamb,Leiby6MacRaeAttorneys forRochester GasandElectricCorporation II'if UNITEDSTATESOFAMERICANUCLEARREGULATORY COMMISSION IntheMatterofRochester GasandElectricCorporation (R.E.GinnaNuclearPowerPlant,UnitNo.l),)))Docket.No.50-244))APPLICATION FORAMENDMENT TOOPERATING LICENSE<iPursuanttoSection50.90'oftheregulations oftheU.S.NuclearRegulatory Commission (the"Commission"
Jeffrey L.Cohen, Esq.New York State Energy Office Swan Street Building Core 1, Second Floor Empire State Plaza Albany, New York l2223 Edward Luton, Esq.Atomic Safety and Licensing Board U.S.Nuclear Regulatory Commis sion Washington, D.C.20555 Dr.Emmeth A.Luebke Atomic Safety and Licensing Board U.S.Nuclear Regulatory Commis sion Washington, D.C.20555 Dr.Dixon Callihan Union Carbide Corporation P.O.Box Y Oak Ridge, Tennessee 37830 L x K.Larson LeBoeuf, Lamb, Leiby 6 MacRae Attorneys for Rochester Gas and Electric Corporation II'i f UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of Rochester Gas and Electric Corporation (R.E.Ginna Nuclear Power Plant, Unit No.l),)))Docket.No.50-244))APPLICATION FOR AMENDMENT TO OPERATING LICENSE<i Pursuant to Section 50.90'of the regulations of the U.S.Nuclear Regulatory Commission (the"Commission"), Rochester Gas and Electric Corporation
),Rochester GasandElectricCorporation
("RGB"), holder of Provisional Operating License No.DPR-18, hereby requests that the Technical Specifica-tions set forth in Appendix A to that license be amended to add requirements for undervoltage protection.
("RGB"),holderofProvisional Operating LicenseNo.DPR-18,herebyrequeststhattheTechnical Specifica-tionssetforthinAppendixAtothatlicensebeamendedtoaddrequirements forundervoltage protection.
This request for a change in the Technical Specifications revises and supersedes our request of December 22, 1977, and is submitted in response to a letter from A.Schwencer, Chief, Operating Reactors Branch 51, dated June 3, 1977.The proposed technical specification change is set forth in Attachment A to this Application.
ThisrequestforachangeintheTechnical Specifications revisesandsupersedes ourrequestofDecember22,1977,andissubmitted inresponsetoaletterfromA.Schwencer, Chief,Operating ReactorsBranch51,datedJune3,1977.Theproposedtechnical specification changeissetforthinAttachment AtothisApplication.
A safety evaluation is set forth in Attachment B.This evaluation also demonstrates that the proposed change does not involve a significant change in the types or a significant increase in the amounts of effluents or any change in the authorized power level of the facility.Attachment C describes why no fee under 10 CFR 170.22 is required.  
Asafetyevaluation issetforthinAttachment B.Thisevaluation alsodemonstrates thattheproposedchangedoesnotinvolveasignificant changeinthetypesorasignificant increaseintheamountsofeffluents oranychangeintheauthorized powerlevelofthefacility.
Attachment Cdescribes whynofeeunder10CFR170.22isrequired.  


WHEREFORE, Applicant respectfully requeststhatAppendixAtoProvisional Operating LicenseNo.DPR-18beamendedintheformattachedheretoasAttachment A.Rochester GasandElectricCorporation ByL.D.White,Jr.VicePresident, ElectricandSteamProduction Subscribed andsworntobeforemeonthis3IdayofJuly1979.ROSEMARIEPERROHEIIQTftIIY PUBLIC,StateofII.Y.,MonroeCountyMyCommission ExpiresMarch30,Qg~
WHEREFORE, Applicant respectfully requests that Appendix A to Provisional Operating License No.DPR-18 be amended in the form attached hereto as Attachment A.Rochester Gas and Electric Corporation By L.D.White, Jr.Vice President, Electric and Steam Production Subscribed and sworn to before me on this 3I day of July 1979.ROSE MARIE PERROHE IIQTftIIY PUBLIC, State of II.Y., Monroe County My Commission Expires March 30, Qg~
Attachment A1.Removepages2.3-4,2.3-8,3.5-4,3.5-4a,and4.1-7.2.Inserttheenclosedrevisedpages2.3-4,2.3-8,2.3-9,3.5-4aand4.1-7.
Attachment A 1.Remove pages 2.3-4, 2.3-8, 3.5-4, 3.5-4a, and 4.1-7.2.Insert the enclosed revised pages 2.3-4, 2.3-8, 2.3-9, 3.5-4a and 4.1-7.
f.Lowreactorcoolantflow->90%ofnormalindicated flow~g.Lowreactorcoolantpumpfrequency
f.Low reactor coolant flow->90%of normal indicated flow~g.Low reactor coolant pump frequency->57.5 Hz.2.3.1.3 Other reactor tri s 2.3.2 a.High pressurizer water level-<92%of span b.Low-low steam generator water level->5%of narrow range instrument span Protective instrumentation settings for reactor trip inter-locks shall be as follows: 2.3.2.1 Remove bypass of"at power" reactor trips at high power (low pressurizer pressure and low reactor coolant flow)for both loops: Power range nuclear flux-<8.'5%of rated power (1)(Note: During cold rod drop tests, the pressurizer high level trip may be bypassed.)
->57.5Hz.2.3.1.3Otherreactortris2.3.2a.Highpressurizer waterlevel-<92%ofspanb.Low-lowsteamgenerator waterlevel->5%ofnarrowrangeinstrument spanProtective instrumentation settingsforreactortripinter-locksshallbeasfollows:2.3.2.1Removebypassof"atpower"reactortripsathighpower(lowpressurizer pressureandlowreactorcoolantflow)forbothloops:Powerrangenuclearflux-<8.'5%ofratedpower(1)(Note:Duringcoldroddroptests,thepressurizer highleveltripmaybebypassed.)
2.3.2.2 Remove bypass of single loss of flow trip at, high power: 2.3.3 Power range nuclear flux-<50%of rated power Relay settings for 480 volt safeguards bus protection shall be as follows: 2.3.3.1 Loss of voltage relay operating time<8.5 seconds for 480 volt safeguards bus voltages<368 volts 2.3.3.2 Acceptable degraded voltage relay operating times and setpoints, for 480 volt safeguards bus voltages<414 volts'and>368 volts are defined by the safeguard equip-ment thermal capability curve shown in Figure 2.3-1.Basis: TTTe Eigh flux reactor trip (low set point)provides redundant pro-tection in the power range for a power excursion beginning fop low power.This trip value was used in the safety analysis.2.3-4 PROPOSED e 0 the minimum DNB ratio increases at lower flow because the maximum enthalpy rise does not increase.For this reason the single pump loss of flow trip can be bypassed below 50%power.The loss of voltage and degraded voltage trips ensure opera-bility of safeguards equipment during a postulated design basis (9)(10)(11) event concurrent with a degraded bus voltage condition.
2.3.2.2Removebypassofsinglelossofflowtripat,highpower:2.3.3Powerrangenuclearflux-<50%ofratedpowerRelaysettingsfor480voltsafeguards busprotection shallbeasfollows:2.3.3.1Lossofvoltagerelayoperating time<8.5secondsfor480voltsafeguards busvoltages<368volts2.3.3.2Acceptable degradedvoltagerelayoperating timesandsetpoints, for480voltsafeguards busvoltages<414volts'and
>368voltsaredefinedbythesafeguard equip-mentthermalcapability curveshowninFigure2.3-1.Basis:TTTeEighfluxreactortrip(lowsetpoint)providesredundant pro-tectioninthepowerrangeforapowerexcursion beginning foplowpower.Thistripvaluewasusedinthesafetyanalysis.
2.3-4PROPOSED e0theminimumDNBratioincreases atlowerflowbecausethemaximumenthalpyrisedoesnotincrease.
Forthisreasonthesinglepumplossofflowtripcanbebypassedbelow50%power.Thelossofvoltageanddegradedvoltagetripsensureopera-bilityofsafeguards equipment duringapostulated designbasis(9)(10)(11) eventconcurrent withadegradedbusvoltagecondition.


==References:==
==References:==


(1)FSAR14.1.1(2)FSAR,Page14-3(3)FSAR14.3.1(4)FSAR14.1.2(5)FSAR72I 7~3(6)FSAR3.2.1(7)FSAR14.1.6(8)FSAR14.1.9(9)LetterfromL.D.White,Jr.toA.Schwencer, NRC,datedSeptember 30,1977(10)LetterfromL.D.White,Jr.toA.Schwencer, NRC,datedSeptember 301977(11)LetterfromL.D.White,Jr.toD.Ziemann,NRC,datedJuly24,19782.3-8PROPOSED o.016001400IISecondLevelI~I.'UnderVoltageProtection LIFigure2.3-11200~ltdI88d001000800600-+-'I'I~I~I~~I"I~1400I+.AllowaI"blIlaOeReyperI'IlatingI200~iReion0Secondary Volts(120Volts).PrimaryVolts(480Volts)t240PercentVolts(460VoltBase)5287080909232036870%80%103.5'1490%120=480104%ISafeguards BusVoltage'.3-9 PROPOSEDs~I Page3of312356NO.ofMIN.MIN.PERMISSIBLE OPERATORACTIONNO.ofCHANNELSOPERABLEDEGREEOFBYPASSIFCONDITIONS OFCHANNELSTOTRIPCHANNELSREDUNDANCY CONDITIONS COLUMN3or5CANNOTBEMET17.Circulating WaterFloodProtection a.Screenhouse 2b.Condenser 18.LossofVoltage/DegradedVoltage480VoltSafe-guardsBus4/bus2/bus2/busNOTE1:Whenblockcondition exists,maintainnormaloperation.
(1)FSAR 14.1.1 (2)FSAR, Page 14-3 (3)FSAR 14.3.1 (4)FSAR 14.1.2 (5)FSAR72I 7~3 (6)FSAR 3.2.1 (7)FSAR 14.1.6 (8)FSAR 14.1.9 (9)Letter from L.D.White, Jr.to A.Schwencer, NRC, dated September 30, 1977 (10)Letter from L.D.White, Jr.to A.Schwencer, NRC, dated September 30 1977 (11)Letter from L.D.White, Jr.to D.Ziemann, NRC, dated July 24, 1978 2.3-8 PROPOSED o.0 1600 1400 I I Second Level I~I.'Under Voltage Protection L I Figure 2.3-1 1200~l td I 8 8 d 0 0 1000 800 600-+-'I'I~I~I~~I"I~1 400 I+.All owa I" bl I la O e Re y per I'I l ating I 200~iRe ion 0 Secondary Volts (120 Volts).Primary Volts (480 Volts)t 240 Percent Volts (460 Volt Base)528 70 80 90 92 320 368 70%80%103.5'14 90%120=480 104%ISafeguards Bus Voltage'.3-9 PROP OSED s~I Page 3 of 3 1 2 3 5 6 NO.of MIN.MIN.PERMISSIBLE OPERATOR ACTION NO.of CHANNELS OPERABLE DEGREE OF BYPASS IF CONDITIONS OF CHANNELS TO TRIP CHANNELS REDUNDANCY CONDITIONS COLUMN 3 or 5 CANNOT BE MET 17.Circulating Water Flood Protection a.Screenhouse 2 b.Condenser 18.Loss of Voltage/Degraded Voltage 480 Volt Safe-guards Bus 4/bus 2/bus 2/bus NOTE 1: When block condition exists, maintain normal operation.
Poweroperation maybecontinued foraperiodofupto7dayswith1channelinoperable orforaperiodof24h~withtwochannelsin~operable.
Power operation may be continued for a period of up to 7 days with 1 channel inoperable or for a period of 24 h~with two channels in~operable.Power operation may be continued for a period of up to 7 days with 1 channel inoperable or for a period of 24 hrs.with two channels inoperable.
Poweroperation maybecontinued foraperiodofupto7dayswith1channelinoperable orforaperiodof24hrs.withtwochannelsinoperable.
Maintain hot shut-down or place bus on diesel generator.
Maintainhotshut-downorplacebusondieselgenerator.
F.P.***Full Power Not Applicable If both rod misalignment monitors (a and b)inoperable for 2 hours or more, the nuclear overpower trip shall be reset to 93%of rated power in addition to the in-creased surveillance noted.If a functional unit is operating with the minimum operable channels, the number of channels to trip the reactor will be column 3 less column 4.A channel is considered operable with 1 out of 2 logic or 2 out of 3 logic.
F.P.***FullPowerNotApplicable Ifbothrodmisalignment monitors(aandb)inoperable for2hoursormore,thenuclearoverpower tripshallberesetto93%ofratedpowerinadditiontothein-creasedsurveillance noted.Ifafunctional unitisoperating withtheminimumoperablechannels, thenumberofchannelstotripthereactorwillbecolumn3lesscolumn4.Achannelisconsidered operablewith1outof2logicor2outof3logic.
TABLE 4.1-1 (CONTINUED)
TABLE4.1-1(CONTINUED)
Channel Check Calibrate Test Remarks 25.Containment Pressure 26.Steam Generator Pressure Narrow range containment pressure (-3.0,+3 psig excluded)27.Turbine First Stage Pressure S 28.Emergency Plan Radiation Instruments 29.Environmental Monitors N.A.N.A.30.Loss of Voltage/Degraded Voltage 480 Volt Safe-guards Bus.N.A.S-Each Shift D-Daily B/N-Biweekly Q-Quarterly M-Monthly P-Prior to each startup if not done previous week R.-Each Refueling Shutdown N.A.-Not applicable Attachment B By letter dated June 3;1977, the NRC requested that RG&E assess the susceptibility of safety related electrical equipment with regard to (1)sustained degraded voltage conditions at the offsite power sources and (2)interaction between the offsite and onsite emergency power sources.An analysis of undervoltage protection at Ginna Station was performed and submitted to the NRC in a letter dated July 21, 1977.This analysis reviewed the current undervoltage protection and described the basis for a modification which would reduce the Station's susceptibility to a sustained degraded voltage.However, due to lack of specific information regarding equipment, setpoints for the additional undervoltage protection were not established at that time.Following receipt of additional information, RG&E submitted setpoints in a letter to the NRC dated September 30, 1977.The proposed specifications are identical to those presented in the September 30 letter.These Specifications will provide protection for a complete loss of 480 volt.bus voltage as well as for degraded voltage conditions.
ChannelCheckCalibrate TestRemarks25.Containment Pressure26.SteamGenerator PressureNarrowrangecontainment pressure(-3.0,+3psigexcluded) 27.TurbineFirstStagePressureS28.Emergency PlanRadiation Instruments 29.Environmental MonitorsN.A.N.A.30.LossofVoltage/Degraded Voltage480VoltSafe-guardsBus.N.A.S-EachShiftD-DailyB/N-BiweeklyQ-Quarterly M-MonthlyP-PriortoeachstartupifnotdonepreviousweekR.-EachRefueling ShutdownN.A.-Notapplicable Attachment BByletterdatedJune3;1977,theNRCrequested thatRG&Eassessthesusceptibility ofsafetyrelatedelectrical equipment withregardto(1)sustained degradedvoltageconditions attheoffsitepowersourcesand(2)interaction betweentheoffsiteandonsiteemergency powersources.Ananalysisofundervoltage protection atGinnaStationwasperformed andsubmitted totheNRCinaletterdatedJuly21,1977.Thisanalysisreviewedthecurrentundervoltage protection anddescribed thebasisforamodification whichwouldreducetheStation's susceptibility toasustained degradedvoltage.However,duetolackofspecificinformation regarding equipment, setpoints fortheadditional undervoltage protection werenotestablished atthattime.Following receiptofadditional information, RG&Esubmitted setpoints inalettertotheNRCdatedSeptember 30,1977.Theproposedspecifications areidentical tothosepresented intheSeptember 30letter.TheseSpecifications willprovideprotection foracompletelossof480volt.busvoltageaswellasfordegradedvoltageconditions.
Both relaying systems will assure that assumptions of all safety analysis are met.Specifically, equipment will be loaded onto the diesel generators within the time assumed in the safety analyses.The undervoltage relay protection, which was identified as"second-level protection" in the June 3, 1977 NRC letter and in our two subsequent letters, will protect equipment against a bus voltage which is greater than the loss of voltage relay setpoint but.less than the voltage guaranteed by equipment manufacturers for continuous duty for Ginna safeguards equipment.
Bothrelayingsystemswillassurethatassumptions ofallsafetyanalysisaremet.Specifically, equipment willbeloadedontothedieselgenerators withinthetimeassumedinthesafetyanalyses.
The proposed undervoltage setpoints will provide the required protection while also assuring that, spurious trips will not occur while equipment is being sequenced onto the diesel generators.
Theundervoltage relayprotection, whichwasidentified as"second-level protection" intheJune3,1977NRCletterandinourtwosubsequent letters,willprotectequipment againstabusvoltagewhichisgreaterthanthelossofvoltagerelaysetpointbut.lessthanthevoltageguaranteed byequipment manufacturers forcontinuous dutyforGinnasafeguards equipment.
The NRC Staff requested that the proposed system be modified to provide"coincident logic" as described in position 1, part b of their June 3, 1977 letter.Provision for"coincident, logic" does not.effect the setpoint information contained in the December 16, 1977 Application for Change to Operating License and re-submitted herewith.However, it has changed the required number of relays necessary to trip.The coincident logic scheme is described in our July 24, 1978 letter (ref.11 to the Technical Specification Basis).Loss of voltage and degraded voltage conditions simulated during the refueling shutdown are performed to verify both system performance and relay calibration.
Theproposedundervoltage setpoints willprovidetherequiredprotection whilealsoassuringthat,spurioustripswillnotoccurwhileequipment isbeingsequenced ontothedieselgenerators.
The safeguards 480 volt loss of voltage and degraded voltage protection systems are not part of the reactor protection system.It is not practical to simulate these conditions during plant operation to satisfy a monthly test requirement.
TheNRCStaffrequested thattheproposedsystembemodifiedtoprovide"coincident logic"asdescribed inposition1,partboftheirJune3,1977letter.Provision for"coincident, logic"doesnot.effectthesetpointinformation contained intheDecember16,1977Application forChangetoOperating Licenseandre-submitted herewith.
Based on the analyses provided in our letters of July 21, 1977, September 30, 1977, and July 24, 1978 the proposed Technical Specification will provide protection against 480 volt bus under-voltage.  
However,ithaschangedtherequirednumberofrelaysnecessary totrip.Thecoincident logicschemeisdescribed inourJuly24,1978letter(ref.11totheTechnical Specification Basis).Lossofvoltageanddegradedvoltageconditions simulated duringtherefueling shutdownareperformed toverifybothsystemperformance andrelaycalibration.
Thesafeguards 480voltlossofvoltageanddegradedvoltageprotection systemsarenotpartofthereactorprotection system.Itisnotpractical tosimulatetheseconditions duringplantoperation tosatisfyamonthlytestrequirement.
BasedontheanalysesprovidedinourlettersofJuly21,1977,September 30,1977,andJuly24,1978theproposedTechnical Specification willprovideprotection against480voltbusunder-voltage.  


AsstatedinourletterofJuly21,1977,weexpectthat,following receiptofapprovalofthemodification designbytheNRC,detailedengineering andprocurement ofequipment willrequireapproximately 9months.Installation mustbeperformed at.acoldorrefueling shutdownandwouldrequireapproximately twoweeks..Theeffective dataofaTechnical Specification shouldbesetconsistent withthisschedule.
As stated in our letter of July 21, 1977, we expect that, following receipt of approval of the modification design by the NRC, detailed engineering and procurement of equipment will require approximately 9 months.Installation must be performed at.a cold or refueling shutdown and would require approximately two weeks..The effective data of a Technical Specification should be set consistent with this schedule.
Attachment CTheTechnical Specification changeproposalrevisesaproposaloriginally submitted December22,1977whichwasfiledpriortotheeffective dateof10CPR170.22.Therevisionisnecessary toincorporate achangeindesignwhichwasrequested bytheNRCStaff.Thedesignchangewasdescribed inoursubmittal ofJuly24,1978.BecauseitisarevisionwhichtheStaffrequested andisnowreviewing, andisnotanewrequest,nofeeisrequired.
Attachment C The Technical Specification change proposal revises a proposal originally submitted December 22, 1977 which was filed prior to the effective date of 10 CPR 170.22.The revision is necessary to incorporate a change in design which was requested by the NRC Staff.The design change was described in our submittal of July 24, 1978.Because it is a revision which the Staff requested and is now reviewing, and is not a new request, no fee is required.
I}}
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Revision as of 16:09, 7 July 2018

Application for Amend to License DPR-18 Re Undervoltage Protection
ML17244A742
Person / Time
Site: Ginna Constellation icon.png
Issue date: 07/31/1979
From: WHITE L D
ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17244A741 List:
References
NUDOCS 7908080446
Download: ML17244A742 (18)


Text

BEFORE THE UNXTED STATES NUCLEAR REGULATORY COMMISSSXON In the Matter of ROCHESTER GAS AND ELECTRIC CORPORATION (R.E.Ginna Nuclear Power Station, Unit No.1)CERTIFICATE OF SERVICE Docket No.52-244 I hereby certify that I have served a document entitled"Application for Amendment to Operating License" with three (3)documents, Attachments A, B, and C, attached thereto, by mailing copies thereof first class, postage pre-paid, to each, of the following persons this 3rd day of August, 1979: Mr.Michael L.Slade 12 Trailwood Circle Rochester, New York 14618 Warren B.Rosenbaum, Esq.One Main Street 707 Wilder Building Rochester, New York 14614 Edward G.Ketchen, Esp.Office of the Executive Legal Director U.S.Nuclear Regulatory Commission Washington, D.C.20555 Mr.Robert N.Pickney Supervisor Town of Ontario 107 Ridge Road West Ontario, New York 14519 V 908080 Q/(p'-

Jeffrey L.Cohen, Esq.New York State Energy Office Swan Street Building Core 1, Second Floor Empire State Plaza Albany, New York l2223 Edward Luton, Esq.Atomic Safety and Licensing Board U.S.Nuclear Regulatory Commis sion Washington, D.C.20555 Dr.Emmeth A.Luebke Atomic Safety and Licensing Board U.S.Nuclear Regulatory Commis sion Washington, D.C.20555 Dr.Dixon Callihan Union Carbide Corporation P.O.Box Y Oak Ridge, Tennessee 37830 L x K.Larson LeBoeuf, Lamb, Leiby 6 MacRae Attorneys for Rochester Gas and Electric Corporation II'i f UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of Rochester Gas and Electric Corporation (R.E.Ginna Nuclear Power Plant, Unit No.l),)))Docket.No.50-244))APPLICATION FOR AMENDMENT TO OPERATING LICENSE90%of normal indicated flow~g.Low reactor coolant pump frequency->57.5 Hz.2.3.1.3 Other reactor tri s 2.3.2 a.High pressurizer water level-<92%of span b.Low-low steam generator water level->5%of narrow range instrument span Protective instrumentation settings for reactor trip inter-locks shall be as follows: 2.3.2.1 Remove bypass of"at power" reactor trips at high power (low pressurizer pressure and low reactor coolant flow)for both loops: Power range nuclear flux-<8.'5%of rated power (1)(Note: During cold rod drop tests, the pressurizer high level trip may be bypassed.)

2.3.2.2 Remove bypass of single loss of flow trip at, high power: 2.3.3 Power range nuclear flux-<50%of rated power Relay settings for 480 volt safeguards bus protection shall be as follows: 2.3.3.1 Loss of voltage relay operating time<8.5 seconds for 480 volt safeguards bus voltages<368 volts 2.3.3.2 Acceptable degraded voltage relay operating times and setpoints, for 480 volt safeguards bus voltages<414 volts'and>368 volts are defined by the safeguard equip-ment thermal capability curve shown in Figure 2.3-1.Basis: TTTe Eigh flux reactor trip (low set point)provides redundant pro-tection in the power range for a power excursion beginning fop low power.This trip value was used in the safety analysis.2.3-4 PROPOSED e 0 the minimum DNB ratio increases at lower flow because the maximum enthalpy rise does not increase.For this reason the single pump loss of flow trip can be bypassed below 50%power.The loss of voltage and degraded voltage trips ensure opera-bility of safeguards equipment during a postulated design basis (9)(10)(11) event concurrent with a degraded bus voltage condition.

References:

(1)FSAR 14.1.1 (2)FSAR, Page 14-3 (3)FSAR 14.3.1 (4)FSAR 14.1.2 (5)FSAR72I 7~3 (6)FSAR 3.2.1 (7)FSAR 14.1.6 (8)FSAR 14.1.9 (9)Letter from L.D.White, Jr.to A.Schwencer, NRC, dated September 30, 1977 (10)Letter from L.D.White, Jr.to A.Schwencer, NRC, dated September 30 1977 (11)Letter from L.D.White, Jr.to D.Ziemann, NRC, dated July 24, 1978 2.3-8 PROPOSED o.0 1600 1400 I I Second Level I~I.'Under Voltage Protection L I Figure 2.3-1 1200~l td I 8 8 d 0 0 1000 800 600-+-'I'I~I~I~~I"I~1 400 I+.All owa I" bl I la O e Re y per I'I l ating I 200~iRe ion 0 Secondary Volts (120 Volts).Primary Volts (480 Volts)t 240 Percent Volts (460 Volt Base)528 70 80 90 92 320 368 70%80%103.5'14 90%120=480 104%ISafeguards Bus Voltage'.3-9 PROP OSED s~I Page 3 of 3 1 2 3 5 6 NO.of MIN.MIN.PERMISSIBLE OPERATOR ACTION NO.of CHANNELS OPERABLE DEGREE OF BYPASS IF CONDITIONS OF CHANNELS TO TRIP CHANNELS REDUNDANCY CONDITIONS COLUMN 3 or 5 CANNOT BE MET 17.Circulating Water Flood Protection a.Screenhouse 2 b.Condenser 18.Loss of Voltage/Degraded Voltage 480 Volt Safe-guards Bus 4/bus 2/bus 2/bus NOTE 1: When block condition exists, maintain normal operation.

Power operation may be continued for a period of up to 7 days with 1 channel inoperable or for a period of 24 h~with two channels in~operable.Power operation may be continued for a period of up to 7 days with 1 channel inoperable or for a period of 24 hrs.with two channels inoperable.

Maintain hot shut-down or place bus on diesel generator.

F.P.***Full Power Not Applicable If both rod misalignment monitors (a and b)inoperable for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or more, the nuclear overpower trip shall be reset to 93%of rated power in addition to the in-creased surveillance noted.If a functional unit is operating with the minimum operable channels, the number of channels to trip the reactor will be column 3 less column 4.A channel is considered operable with 1 out of 2 logic or 2 out of 3 logic.

TABLE 4.1-1 (CONTINUED)

Channel Check Calibrate Test Remarks 25.Containment Pressure 26.Steam Generator Pressure Narrow range containment pressure (-3.0,+3 psig excluded)27.Turbine First Stage Pressure S 28.Emergency Plan Radiation Instruments 29.Environmental Monitors N.A.N.A.30.Loss of Voltage/Degraded Voltage 480 Volt Safe-guards Bus.N.A.S-Each Shift D-Daily B/N-Biweekly Q-Quarterly M-Monthly P-Prior to each startup if not done previous week R.-Each Refueling Shutdown N.A.-Not applicable Attachment B By letter dated June 3;1977, the NRC requested that RG&E assess the susceptibility of safety related electrical equipment with regard to (1)sustained degraded voltage conditions at the offsite power sources and (2)interaction between the offsite and onsite emergency power sources.An analysis of undervoltage protection at Ginna Station was performed and submitted to the NRC in a letter dated July 21, 1977.This analysis reviewed the current undervoltage protection and described the basis for a modification which would reduce the Station's susceptibility to a sustained degraded voltage.However, due to lack of specific information regarding equipment, setpoints for the additional undervoltage protection were not established at that time.Following receipt of additional information, RG&E submitted setpoints in a letter to the NRC dated September 30, 1977.The proposed specifications are identical to those presented in the September 30 letter.These Specifications will provide protection for a complete loss of 480 volt.bus voltage as well as for degraded voltage conditions.

Both relaying systems will assure that assumptions of all safety analysis are met.Specifically, equipment will be loaded onto the diesel generators within the time assumed in the safety analyses.The undervoltage relay protection, which was identified as"second-level protection" in the June 3, 1977 NRC letter and in our two subsequent letters, will protect equipment against a bus voltage which is greater than the loss of voltage relay setpoint but.less than the voltage guaranteed by equipment manufacturers for continuous duty for Ginna safeguards equipment.

The proposed undervoltage setpoints will provide the required protection while also assuring that, spurious trips will not occur while equipment is being sequenced onto the diesel generators.

The NRC Staff requested that the proposed system be modified to provide"coincident logic" as described in position 1, part b of their June 3, 1977 letter.Provision for"coincident, logic" does not.effect the setpoint information contained in the December 16, 1977 Application for Change to Operating License and re-submitted herewith.However, it has changed the required number of relays necessary to trip.The coincident logic scheme is described in our July 24, 1978 letter (ref.11 to the Technical Specification Basis).Loss of voltage and degraded voltage conditions simulated during the refueling shutdown are performed to verify both system performance and relay calibration.

The safeguards 480 volt loss of voltage and degraded voltage protection systems are not part of the reactor protection system.It is not practical to simulate these conditions during plant operation to satisfy a monthly test requirement.

Based on the analyses provided in our letters of July 21, 1977, September 30, 1977, and July 24, 1978 the proposed Technical Specification will provide protection against 480 volt bus under-voltage.

As stated in our letter of July 21, 1977, we expect that, following receipt of approval of the modification design by the NRC, detailed engineering and procurement of equipment will require approximately 9 months.Installation must be performed at.a cold or refueling shutdown and would require approximately two weeks..The effective data of a Technical Specification should be set consistent with this schedule.

Attachment C The Technical Specification change proposal revises a proposal originally submitted December 22, 1977 which was filed prior to the effective date of 10 CPR 170.22.The revision is necessary to incorporate a change in design which was requested by the NRC Staff.The design change was described in our submittal of July 24, 1978.Because it is a revision which the Staff requested and is now reviewing, and is not a new request, no fee is required.

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