NRC Generic Letter 1985-10

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NRC Generic Letter 1985-010: Technical Specifications for Generic Letter 1983-028, Items 4.3 and 4.4
ML031140409
Person / Time
Issue date: 05/23/1985
From: Thompson H L
Office of Nuclear Reactor Regulation
To:
References
GL-85-010, NUDOCS 8505210131
Download: ML031140409 (16)
v • d • e


UNITED STATESNUCLEAR REGULATORY COMMISSIONWASHINGTON, D. C. 20555May 23, 1985TO ALL BABCOCK AND WILCOX PRESSURIZED WATER REACTOR LICENSEES AND APPLICANTSGentlemen:

SUBJECT: TECHNICAL SPECIFICATIONS FOR GENERIC LETTER 83-28, ITEMS 4.3AND 4.4 (Generic Letter 85-10)Item 4.3 of Generic Letter 83-28, "Required Actions Based on GenericImplications of Salem ATWS Events," established the requirement for theautomatic actuation of the shunt trip attachment for B&W plants. Also,licensees are to submit any needed technical specification change requestsas soon as practical following staff review and approval of the modifieddesign. Item 4.4 of Generic Letter 83-28 requires that the appropriatesurveillance and test sections of the technical specifications be revised toinclude testing of the silicon controlled rectifiers used to interrupt powerto control rods.In the staff's evaluation of the B&W generic design modifications for auto-matic actuation of the shunt trip attachment, as described in individualletters to all B&W operating reactors, dated September 12, 1983, the staffconcluded that technical specification changes should be proposed by licenseesand that they would be reviewed on a plant specific basis. In the staff's* review of plant specific responses to the generic letter, some licensees haveindicated that changes to the technical specifications are not required. Insuch cases, the staff has found this to be unacceptable and has indicated thatproposed technical specification changes should be submitted to reflect inde-pendent testing of the shunt and undervoltage trip attachments consistent withthe design of the test features provided.Therefore, licensees are requested to submit proposed technical specificationchanges which are responsive to the guidance noted in the enclosure. Theenclosed guidance will be used to revise the Standard Technical Specificationsfor B&W plants, and it will be used by the staff as a basis to review changesto technical specifications submitted by licensees and for the review ofproposed technical specifications for operating license applications.For plants which have implemented the shunt trip modifications, a schedule forsubmittal of proposed technical specification changes should be establishedthrough discussions with the individual Project Manager for each facility. 60., X'..~~~VJ'5105ZI0/3/ l C -2-In addition, discussions with the individual Project Managers should establisha schedule for plants which have not implemented the shunt trip modifications.Proposed technical specifications should be submitted as soon as practicalfollowing staff review and approval of modified design. For operating licenseapplicants, proposed technical specifications should include requirements whichare responsive to the enclosed guidance.This request for information was approved by the Office of Management andBudget under clearance number 3150-0011 which expires April 30, 1985. Shouldyou have any questions, the staff contact is R. Karsch. Mr. Karsch can bereached on (301) 492-8563.Tarigbal Signed byHugh L Thompson, JrHugh L. Thompson, Jr., DirectorDivision of Licensing

Enclosure:

Sample Technical SpecificationsList of Generic Letters*PREVIOUS CONCURRENCE SEE DATEORAB*TAlexion:cl03/4/85D:DL -W/ho /8on05/16 /85SL:TSRG*EButcher03/4/85SL:ORAB*JHannon03/6/85C:ICSB*FRosa03/11/85C:ORAB:DL*GHolahan03/7/85AD/SA:DL*DCrutchfield03/18/85% (!

ENCLOSURETECHNICAL SPECIFICATION CHANGESFOR REACTOR TRIP BREAKERS(B&W PLANTS)BackgroundAs a consequence of the Salem ATWS event, Item 4.3 of Generic Letter 83-28established the requirements for the automatic actuation of the shunt trip-attachment for reactor trip breakers. Further, licensees are to submit anyneeded technical specification change requests prior to declaring themodified system operable. A number of the responses from operating reactorshave indicated that no technical specification changes are required for thismodification.The staff has reviewed the guidance provided in the Standard TechnicalSpecifications (STS) for B&W Plants, NUREG-0103, and finds that additionalclarification of both the limiting conditions of operation and surveillancerequirements are appropriate as a result of the design modifications toinclude automatic actuation of the shunt trip attachments. In addition, Item4.4 of the generic letter reauires that technical specification surveillancerequirements be revised to include testing of the silicon controlledrectifiers (SCP). The STS for BW Plants will be revised to include thechanges noted herein. Pending formal revision of the STS, this documentprovides guidance to licensees and operating license applicants onappropriate technical specifications in response to Items 4.3 and 4.4 of theGeneric Lette K>1-2-DiscussionThe operability requirements for the reactor trip breakers are specified inTable 3.3-1 of the STS (see Attachment 2). The specification states thatboth reactor trip breakers shall be operable in Modes 1 and 2, and when thebreakers are in the closed position, the control rod drive system is capableof rod withdrawal, and fuel is in the reactor vessel. The action statementfor an inoperable breaker requires that the breaker be placed in a trippedcondition within one hour.With the addition of the automatic actuation of the shunt trip attachment(STA), diverse features exist to effect a reactor trip for each breaker. Ifone of these diverse trip features is inoperable, a decision would have to bemade with regard to the operability status of the reactor trip breaker. Thedefinition of OPERABLE-OPERABILITY in Section 1.0 of the STS states that acomponent shall be operable or have operability when it is capable ofperforming its safety function. Since either trip feature being operablewould initiate a breaker trip on demand, it would be overly conservative totreat a breaker as inoperable if one of these diverse trip features wereinoperable. However, on the other hand the reliability of the reactor tripsystem would be reduced if each diverse trip feature is not maintained in anoperable statu K)-3 -The reactor trip breaker surveillance test should independently verify theoperability of the shunt and undervoltage trip features of the reactor tripbreaker as part of a single sequential test procedure. Therefore, thesurveillance test which identifies a failure of one diverse trip feature alsoconfirms the operability of the other trip feature. As a consequence, thereis a higher degree of confidence that this trip feature would be capable ofinitiating a reactor trip in the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Accordingly, an additionalaction statement will be included in the STS for the reactor trip breakers asfollows:ACTION -With one of the diverse trip features (undervoltage or shunttrip attachment) inoperable, restore it to OPERABLE status in 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />sor place the breaker in trip in the next hour.In accordance with the requirements of Item 4.4 of the Generic Letter, theSCR relays have also been included in the changes to Table 3.3-1 to definetheir operability requirements. The reactor trip system design for B&Wplants Includes two basic configurations; the Oconee design shown in Figure3.4 and the Davis Besse design shown in Figure 3.5 (see Attachment 1). Inthe Oconee design the SCR relays for the regulating rods duplicate the tripfunction of the DC breakers for the safety rods. Therefore, for this designan inoperable channel should be placed in trip as required by actionstatement 7. However., in the Davis Besse design, the SCR relays provide athird means to insure that power is removed from all rods to Initiate a K>~-4 -reactor trip. Therefore, placing inoperable channels of SCR relays in tripwould only increase the potential for inadvertent reactor trips withoutsignificantly reducing the potential of an ATWS event, when considering theincreased reliability of the reactor trip breaker afforded by the incor-poration of diverse trip features. For plants with the Davis Besse design,a new action statement will be included in Table 3.3-1 of the STS as follows:ACTION -With one or both channels of SCR relays inoperable,restore the channels to OPERABLE status during the next COLDSHUTDOWN exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.The B&W generic design modification includes test features which permitindependent testing to verify the operability of the shunt and undervoltagetrip attachments. As noted above, operability as applied to the diverse tripfeatures of breakers may have different degrees of safety significance. Inorder to be consistent with the intent ot the test features provided, thefollowing notation will be included in the surveillance requirements specifiedin STS Table 4.3-1 for reactor trip breakers (see Attachment 2):"The CHANNEL FUNCTIONAL TEST shall independently verify theOPERABILITY of the undervoltage and shunt trip attachments ofthe Reactor Trip Breakers."

-5-Consistent with the requirements of Item 4.4 of the Generic Letter, thesurveillance requirements for SCR relays will also be included in Table 4.3-1.For plants with the Oconee design, the channel functional test of the SCR-relays will be specified as monthly since the SCR relays used for the regu-lating rods duplicate the trip function of the DC breakers for the safetyrods. For plants with the Davis Besse design, the frequency of the channelfunctional test of the SCR relays will be specified as at least once per 18months. The less frequent testing for the SCR relays in the Davis Bessedesign is due to their less critical, i.e., duplicative, safety function.Attached are marked-up pages of the applicable STS tables with thesechanges. Proposed changes to plant specific technical specifications will beevaluated by the staff based on this guidanc AtL I POWER SOURCE8 FOR AOIVEN ROD GROUP mS EINTERRUPTED IN ORDER FORTHE RODS IN THAT GROUPTO DROP INTO THE COREPLANT PROCMINSTRUMENT CHANNflSISENSORS ANDTANSMIllERS.SISTASLES. ETC.I ANDFIELD

CONTACT

STRIP MODULES IILOGIC CHANNEW31IECONTR ROD DRIVE CONTROL SV3TMlMANUAL TRIP --RECTOR TRIP SysTTRIP MODULEOUTPUT TO THE CROCSIC-)-i-.8(4-SILICON*CONTROLLEDRECTIFIERISCRIRELAYSF URIPAtTAGET~RI ATTCHMENTGROUP 2 GROUP 2GROUP? GROUPSSAFETY RODSREGULATING RODSFigure 3.4 Babcock 8 Wlleox Reactor Trip. System (Oconee. TMI, CR-3, ANO-1, Rancho Seco)

CR00NtRoL ROD OIVE CONiWOt sYSTEIM)PLAN? PROCeSI TUMENT CHANNtEL fISENSORS AN0 fTRANSMIr (S. ABISTASLES. ETC., ANDFIELD

CONTACT

SJ I4mVACMAIN BusMVACSECONOART BUS-I

  • I -lI .*I I, -IAACTRIPOLNG SNEAKERSPOWER TOALL P00GROUPSVit-0 lk -m-I IL ILIL I 111_ILOGIC CHANNELS) I < C U4 ]LCU4]MANUAt TIM IIMAI"IALT --} -----FRACTOR TRIP SSTE fTRIP MODlahOUTPUT TO THE CROCS A CSCR MAINcowmRot POWERCF6IIIIIIIII --I.IIIII1I--___________ ___________ ___________IIIIIIIIICID C-401 CI-'I-ITp I01 CIa t1D c---ipI0-1f---l --fla CS C--Di-1O CGRP U2) GOPGROUPS GROUPS GROUP? GROUPSJ CONTROL POWERI /1 CALL POWER SOURES FOM AGIVN ROD GR"OU MUST StITERRUPTED 11 ORDER FORTHE RODS I" THAT GROUPTO DROP INTO THE COREC'O#4TROLLED' RECTIFIERISCRISAFETY RODSREGULATING RODSa) RELAYSU TIffPAOLTA02 TBTRIP ATTACHMENTFigure 3.5 Babcock EP Wilcox Reactor Trip System (Davis-Besse)

II"-tInTABLE 3.3-1REACTOR PROTECTIONSYSTEM INSTRUMENTATIONSYSTEM INSTRUMENTATIONTOTAL NO.FUNCTIONAL UNIT OF CHANNELS1. Manual Reactor Trip 22. Nuclear Overpower 43. RCS Outlet Temperature--High 44. Nuclear Overpower Based on RCSFlow and AXIAL POWER IMBALANCE 45. RCS Pressure--Low 46. RCS Pressure--High 47. Variable Low RCS Pressure 48. Nuclear Overpower Based onPump Monitor 49. Reactor Containment Pressure--High 410. Intermediate Range, Neutron Fluxand Rate 211. Source Range, Neutron Flux and RateA. Startup *2B. Shutdown 212. Control Rod Drive Trip Breakers 2 per tripsystem13. Reactor Trip Module 2 per tripsystem14. Shutdown Bypass RCS Pressure-High 4CHANNELSTO TRIP .1222(a)(b)2(a)22(a)2(a)(b)20001 per tripsystem1 per tripsystem2MINIMUMCHANNELSOPERABLE2333333332APPLICABLEMODES1, 2, and *1, 21, 21, 21, 21, 21, 21, 21, 21, 2, and *ACTION12#3f22#3#3#3#3#3#4567. 97.8(I'rn.E2 2##, and *1 3, 4 and 52 per 1, 2, and *trip system2 per 1, 2, and *trip system3 1, 2('I15. SCR Relays2221, 2. and *7 (Oconee Desinn).. 10 (Davis BesseDesign)

TABLE 3.3-1 (Continued)ACTION 7ACTION STATEMENTS (Continued)-With the number of OPERABLE channels one less than the TotalNumber of Channels, STARTUP and/or POWER OPERATION may proceedprovided the following conditions are satisfied:a. Within 1 hour:1. Place the inoperable channel in the tripped condition,or2. Remove power supplied to the control rod trip deviceassociated with the inoperable channel.b. One additional channel may be bypassed for up to 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />sfor surveillance testing per Specification 4.3.1.1, andthe inoperable channel above may be bypassed for up to30 minutes in any 24-hour period when necessary to testthe trip breaker associated with the logic of the channelbeing tested per Specification 4.3.1.1. The inoperablechannel above shall not be bypassed to test the logic of achannel of the trip system associated with the inoperablechannel..S ..ACTION 8-ACTION 9 -ACTION 10 -With less than the Minimum Number of Channels OPERABLE, declarethe bypass inoperable and verify that all channels served bythe bypass are OPERABLE, or satisfy the associated ACTIONrequirements.With one of the Reactor Trip Breaker diverse trin features (under-voltane or shunt trip attachment) inonerable, restore it to OPERABLEstatus in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or nlace the breaker in trin in the next hour.With one or both channels of SCR Relays inoperable, restore thechannels to OPERABLE status durinn the next COLD SHUTDOWNexceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.IELW-STS3/4 3-5 l ^03--I4nTABLE 4.3-1INSTRUMENTATION SURVEILLANCE REQUIREMENTSREACTOR PROTECTION SYSTEMFUNCTIONAL UNIT1; -Manual Reactor Trip2. Nuclear Overpower3. RCS Outlet Temprature--High4. Nuclear Overpower Based on RCSFlow and AXIAL POWER IMBALANCE5. RCS Pressure--LowCHANNELCHECKN. A.SSS(4)SCHANNELCALIBRATIONN.A.D(2) and Q(6)RCHANNELFUNCTIONALTESTS/U(l)MHMMODES FOR WHICHSURVEILANCE ISREQUIRED1, 2, and *1, 21, 2M(3)Rand Q(6,7)MH1, 21, 2.S6. RCS Pressure--High S7. Variable Low RCS Pressure S8. Nuclear Overpower Basedon Pump Monitor S9. Reactor Containment Pressure--High S10. Intermediate Range, NeutronFlux and Rate S11. Source Range, Neutron Fluxand Rate S12. Control Rod Drive Trip Breaker N.A.13. Reactor Trip Module N.A.14. Shutdown Bypass RCSPressure-High SRRRRR(rMMM1, 21, 2I, 2I, 2MMftCfo)* f1.A.N. A.S/U(I)(5)M and S/U(1)(5)*M and S/U(l)(10)S* ':S/U(8)I, 7, alld *7. :, 14, 'J, amVlI, 7, ,anl *I, 7, anm *1, 2Ift('))15. SCR RelaysH. A.Ni. A.M (Oconee Desinn) 1, 2, and *R (Davis Besse Design)I TABLE 4.3-1 (Continued)NOTATION* -With any control rod drive trip breaker closed.(1) -If not performed in previous 7 days.(2) -Heat balance only, above 15% of RATED THERMAL POWER. Adjust channelif absolute difference is greater than or equal to (2) percent.(3) -Compare incore to out-of-core measured AXIAL POWER IMBALANCE above15% of RATED THERMAL POWER. Recalibrate if absolute difference isgreater than or equal to (2) percent.(4) -AXIAL POWER IMBALANCE and loop flow indications only.(5) -Verify at least one decade overlap if not verified in previous7 days.(6) -Neutron detectors may be excluded from CHANNEL CALIBRATION.(7) -Flow rate measurement sensors may be excluded from CHANNELCALIBRATION. However, each flow measurement sensor shall becalibrated at least once per 18 months.(8) -Logic only, if not performed in previous 92 days.(9) -The total bypass function shall be demonstrated OPERABLE duringCHANNEL CALIBRATION testing of each channel affected by bypassoperation.(10) -The CHANNEL FUNCTIONAL TEST shall independentlv verify the OPERABILITYof the undervoltane and shunt trip attachments of the Reactor Trip Breakers.B&W- STS3/4 3-8 I4LIST OF RECENTLY ISSUED GENERIC LETTERSGENERICLETTER NO.84-2084-2184-2284-2384-2485-0185-0285-0385-0485-0585-0685-0785-0885-0985-10SUBJECTScheduling Guidance for Licensee Submittalsof Reloads that Involve Unreviewed SafetyQuestionsLong Term Low Power Operation in PWR'sNot usedReactor Vessel Water Level Instrumentationin BWRsClarification of Compliance to 10 CFR 50.49Environmental Qualification of ElectricalEquipment Important to Safety for NuclearPower PlantsFire Protection Policy Steering CommitteeReportStaff Recommended Actions Stemming From NRCIntegrated Program for the Resolution ofUnresolved Safety Issues Regarding SteamGenerator Tube IntegrityClarification of Equivalent Control CapacityFor Standby Liquid Control SystemsOperator Licensing ExaminationsInadvertent Boron Dilution EventsQuality Assurance Guidance for ATWSEquipment that is not Safety-RelatedImplementation of Integrated Schedulesfor Plant Modifications10 CFR 20.408 Termination Reports -FormatTechnical Specifications for GenericLetter 83-28, Item 4.3Technical Specifications for GenericLetter 83-28, Items 4.3 and 4.4DATE8/20/8410/16/8410/26/8412/27/841/9/854/15/851/28/851/29/851/31/854/16/855/02/855/23/855/23/855/23/85 i'-2-For plants which have not implemented the shunt trip modifications, proposedtechnical specifications should be submitted as soon as practical followingstaff review and approval of modified design. For operating licenseapplicants, proposed technical specifications should include requirementswhich are responsive to the enclosed guidance.This request for information was approved by the Office of Management andBudget under clearance number 3150-0011 which expires April 30, 1985.Should you have any questions, sL AJ- A 6e/0Ca 1 eu7e u4 rHugh L. Thompson, DirectorDivision of Licensing

Enclosure:

Sample Technical SpecificationsOSI s S : AB C:ICSB C Ski AD/SA:DLTAl l E her J annon FRosa Alahan DCrutchfield03/.4/85 03/L /85 03/( /85 03//l/85 03/7 /85 03/ /85D:DLHThompson03/ /85

-2 -For plants which have not implemented the shunt trip modifications, proposedtechnical specifications should be submitted as soon as practical followingstaff review and approval of modified design. For operating licenseapplicants, proposed technical specifications should include requirementswhich are responsive to the enclosed guidance.This request for information was approved by the Office of Management andBudget under clearance number 3150-0011 which expires April 30, 1985.Should you have any questions, they should be directed to the NRC ProjectManager for your facility.Hugh L. Thompson, Jr., DirectorDivision of Licensing

Enclosure:

Sample Technical Specifications*PREVIOUS CONCURRENCE SEE DATEORAB*TAlexion:cl03/4/85SL:TSRG*EButcher03/4/85SL:ORAB*JHannon03/6/85C:ICSB*FRosa03/11/85C:ORAB:DL*GHolahan03/7/85DCrt chfield03/\ /85.hD:DLHThompson03/ /85

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