ML20137T269

1996 Annual Environ Operating Rept

ML20137T269
Person / Time
Site:	Grand Gulf
Issue date:	12/31/1996
From:	Hughey W ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GNRO-97-00028, GNRO-97-28, NUDOCS 9704160048
Download: ML20137T269 (48)

Text

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l O Entsgy Oper1 tion 3,Inc.

. h[g P.O. Box 756 i

Port Gbson, MS 39150 Tel 601437-6470 j W.K.Hughey -

Director I Nuclear Safety & RecyAtory l April 9, 1997 m l l

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1 U.S. Nuclear Regulatory Commission 1 Mail Station P1-37 Washington, D.C. 20555  !

Attention: Document Control Desk

Subject:

Grand Gulf Nuclear Station Docket No. 50-416 License No. NPF-29 1996 Grand Gulf Nuclear Station (GGNS) Annual Environmental Operating Report (AEORI GNRO-97/00028 Gentlemen:

Attached is the Grand Gulf Nuclear Station (GGNS) Annual Environmental Operating Report (AEOR) for the period I January 1, 1996 through December 31, 1996. This report is submitted in accordance with the Environmental Protection Plan, Appendix B to the GGNS Operating License (NPF-29),

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Section 5.4, " Station Reporting Requirements .

If you have any questions or require additional information concerning this report,.please contact Gary C. Coker at (601) j 437-2242, or this office. '

Yours truly, '

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, / COO l e 1 l F#

l WKH/MJL

( attachment: 1996 Annual Environmental Operating Leport cc: (See Next Page) 9704160048 961231 PDR ADOCK 05000416 R PDR

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April 9, 1997 GNRO-97/00028 Page 2 of 2 cc: GGNS NRC Senior Resident (w/a)

Mr. R. B. McGehee (w/a)

Mr. N. S. Reynolds (w/a)

Mr. H. L. Thomas (w/o)

Mr. J. W. Yelverton (w/o)

Mr. E. W. Merschoff (w/a)

Regional Administrator  !

U.S. Nuclear Regulatory Commission i Region IV I 611 Ryan Plaza Drive, Suite 400 1

, Arlington, TX 76011 Mr. J. N. Donohew, Project Manager (w/2)- I Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission

! Mail Stop 13H3 Washington, D.C. 20555 4

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k AEOR96. DOC

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l GRAND GULF NUCLEAR STATION 1996 ANNUAL ENVIRONMENTAL OPERATING REPORT I

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1 PREFACE i

l The Annual Environmental Operating Report (AEOR) provides information 4

I and data obtained from implementation of Grand Gulf Nuclear Station's

- (GGNS) Environmental Protection Plan (EPP), Appendix B to the GGNS Operating License (NPF-29), which only requires terrestrial issues to be l

addressed, for the period January 1 through December 31,1996.

{. The GGNS Final Environment Statement did not identify any aquatic issues.

!- Consequently, the EPP does not ad.iress any. The GGNS National Pollutant Discharge Elimination System (NPDES) Permit issued by the Mississippi Department of Environmental Quality (MDEQ) contains effluent limitations

] and monitoring requirements for aquatic matters. The MDEQ regulates matters involving water quality and aquatic biota.

I This report addresses only those issues required by the EPP. In the past, the AEOR included activities associated with the GGNS Construction Permit, j and an Updated Final Safety Analysis Report (UFSAR) requirement which involved reporting regional and perched groundwater levels and precipitation data in the AEOR. However, the Nuclear Regulatory Commission approved

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cancellation of Construction Permit CPPR-119 for Unit 2 on August 21,1991 1 (GNRI-91/00176), and GGNS deleted the UFSAR AEOR reporting requirement j in 1993 (GNRI-93/00025); therefore, GGNS terminated reporting activities j associated with these items. i

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l TABLE OF CONTENTS i

l PAGE PREFACE.. . . . . . . . . . ii '

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SECTION TOPIC i

1.0 INTRODUCTION

.. . . . . . . . . . 1 1.1 Impact Assessment and Summary.. . .. .. . 1 2.0 ENVIRONMENTAL SURVEILLANCE ACTIVITIES... . .. 1 -

l 2.1 Transmission Line Surveys.. . . 1 l

2.2 Cooling Tower Dria Program.. . . ..... . . . . . . . I 2.3 Environmental Evaluations.. .. .. 1

, 3.0 -OBSERVATIONS AND DISCUSSIONS.. . 2 3.1 Environmental Evaluations.. . 2 1

4.0 ADMINISTRATIVE REQUIREMENTS.. . 2 4.1 EPP Changes.. . . . .. 2 4.2 EPP Noncompliances... . . . 2 4.3 Nonroutine Reports... .. .. 2 4.4 Potentially Significant Unreviewed Environmental Issues.. . . . . .. 2 TABLE TOPIC 4-1 1996 Emironmental Evaluation Summary.. . . 3 i

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1.0 INTRODUCTION

1.1 Impact Assessment and Summary GGNS personnel monitored the environmental impact of plant operational activities between January I and December 31,1996. The monitoring results contained in the following sections indicate no adverse impact on the environment due to operation of GGNS. In addition, GGNS personnel have not observed harmful effects or evidence of trends toward irreversible damage to the surrounding environment at GGNS.

i 2.0 ENVIRONMENTAL SURVEILLANCE ACTIVITIES 2.1 Transmission Line Surveys GGNS discontinued this program in 1988.

2.2 Cooling Tower Drift Program GGNS discontinued this program in 1992.

2.3 Environmental Evaluations The EPP permits changes in GGNS design or operation and performance of tests or experiments that affect the environment, provided they do not

! mvolve a change in the EPP or an unreviewed environmental question.

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However, EPP requirements do not apply to changes, tests or experiments l which do not affect the environment. Also, EPP requirements do not relieve GGNS of 10 CFR 50.59 requirements, " Changes, Tests and l Experiments," which address the question of safety associated with proposed changes, tests and experiments.

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The EPP excludes changes, tests or experiments from the evaluatian:

- If all measurable environmental effects confined to onsite areas -

previously disturbed during site preparation and plant construction, or

- If required to achieve compliance with other federal, state or local requirements.

3.0 OBSERVATIONS AND DISCUSSIONS i 3.1 Environmental Evaluations l

GGNS activities did not include any unreviewed environmental questions during 1996. Review of environmental evaluations indicated routine matters within the scope of expected activities. GGNS did not observe any environmental consequences as a result of conduct of the activities 1 evaluated. Table 4-1, which has the evaluations attached, summarizes environmental evaluations performed in 1996.

4.0 ADMINISTRATIVE REQUIREMENTS l

4.1' EPP Changes l GGNS made no changes to the EPP in 1996.

4.2 EPP Noncompliances l GGNS activities contained no EPP noncompliances during 1996. ,

I 4.3 . Nonroutine Repons GGNS submitted no nonroutine reports in 1996.

4.4 Potentially Significant Unreviewed Environmental Issues GGNS encountered no potentially significant unreviewed environmental issues in 1996. GGNS personnel made changes in station design and operation, tests and experiments, of which none resulted in an unreviewed environmental question, in accordance with the EPP, paragraph 3.1, Plant Design and Operation.

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l Table 4-1 1996 Environmental Evaluation Summary

• Safety and Environmental Evaluation Number Gescription 96-0026-R00 Revises SAR wording which requires the Radiation Control Manager to review each plant design change or modification, after the change has been through the development review and approval process, during the implementation authorization review process for ALARA concerns as part of the ALARA program. The change is to require the review of change documents as directed by the GGNS General Manager.

96-0069-R00 Incorporates the use of electronic alarming dosimeters i into the UFSAR l 96-0069-R00! Evaluates the use of electronic alarming dosimeters and modifies Table 12.5-1 of the UFSAR. I 96-0071-R00 Storage of radioactive materials in areas outside the controlled access area, protected area and plant structures.

96-0085-R00 Reduce the surveillance test frequency ofinterlock functions.

96-0087-R00 Documents troubleshooting efforts for recirculation flow control valve B (IB33F0608).

96-0098-R00 Revises Technical Requirements Manual surveillance require nent 7.6.3.3.g.2 to allow the Standby Liquid Control pump relief valves to be tested at least once per 18 months during plant or system shutdowns to verify that they open within the specified 3% tolerance.

I See attached for completed evaluations.

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we cue ' s RELATE 0 000Mlg GRAND GULF NUCLEAR STATION UNrr1 CHANGES, TESTS OR ExrERaWENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM I. Safety Evaluation Overview Page _L,of.g A. Reference Data ORIGINATOR:

N Dece8 FRANMUN DEFT /88CT: " P&SE EVAL f: " T(p co,g/, . ( .co (9

DOCUserrEVAWATED: UFSAR CR #  %- oic .

{ SYrmdNO:(8) m N/A ( INSERT N/A 1F NOTWUCASLE)

REFERENCES:

"Reo. Guce 8.8, NPEAP 311 FSAR CHANOE REQUIRED 7 m E Yes a No Ca # 9(, .o a 3-i FSAR SECTIONS TO BE REVISED: 12.1.1.2 TRM CHANGE REQUDLED7 a O Yes E No j TECH. SPEC. CHANGE REQUEED7 a a Yes E No CA# CR or(n/s) h Is THE VAuDrrY OF THIS SAFETY EVAWADON DEPENDENT ON ANY CHANGES ODER THAN O Yes THE CHANGE BEINO EVALUATED (E.O. PROCEDURAL, OPERA 110NAL CONDIDONS)? "'

E No 1

EXPLAIN: NO OTHER CHANGES ARE REQUIRED TO SUPPORT THis SAR CHANGE REQUEST. .

1 i IF YES TO THE IAST QUESTION, HAVE DE ORGANIZATIONS RESPONSIBLE FOR DIOSE CHANOES O Yes BEEN NOTTFIED7 UU

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1 (THE REsPONSIBIE ORGANIZATIONS MUFF BE NOTIFED PRIOR 10 IMPLEMEN11NO THIS CHANOE.)

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} Signature 8 and Approval 8 of Attached Safety and Environmental Evaluation 1

gM Evaluated: on l n, , , , , ,,,

j (PRINT NAME) ORIGINATOR i (SiONATURE) DATE i

j Reviewed: 08 ggg , ,_  % / 4./.96 (PRDrr NAMI) INDEPENDENT REVIEWER / IONATURE) DATE Reviewed: "* EickV bD / / I~

(PRIRT NAME) OTHER REVIEWER (IF REQUIRED) (810 NATURE) DATE Plant Safety Review Committee Review on A -

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CHAll5 TAN SRC DATE O

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SE no. 46 ' 00Ho - A -00 Page 1 of e j B. Executive Summary (Au0 saavns As owr TO waC stma4AaY anroar) j BRN DEscarn0N OF CHANoE, TEST OR EXPERIMENT O '

l i M Pn0Poeno CNmet nEwEES SAR wonome nHCH neouneES THE RAcurKw CoNTnot MAnnoEn T i

nEVEwEACNMANTDeanN CNmes on MoonCArts, APTER Tw CNmetNAS aEEN THnoveN THE

' DDELOPMENT REVnW AnD APPROVAL PROCESS, DUNNG H nePLEMENTATKw AUTNoneATKw nEmEw

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PROCESS PoM ALARA CONCenNS. AS A PAR? OP TNE ALARA Pn00nAnt. M CHANGEIS TO nEQ ntwtw of chm 0EDOCin8ENTS AS ONRECTED BY THE GGNS GENERAL MANAGER.

REASoW POR CHAHoE, TEST OR EXPERDdENTM l

1 M PnoPo8ED CNm0E n NTEFDE170 MLP STREAACBE H Pn0 CESS FOR AUTNOMQATxw op CHANGES 1 PORnePLEMENTATKW, ONCEISSMDBYENGanEnNGASANAPPROVEDDESGNPACKAaq. THERActATKw PROTECTKw MANAGER REwEWn MEDWOANT To orHER snanAM nEoungMENTS EPPECTlwr DUnNG DESMN j

OEv5LoPMENT AND nonK nonsneENTA TKw PROCESSES AND SS NOT AEEDEO.

i1 SAFETY EVAWAnoN

SUMMARY

AND CoNCWSONS ""

I Tkt PMoP0sEc CNeet To TNar SAR n TO ELnaNATE A neomamT nEmEw nY THE RActATxw PnoTECTxW MANAGEM OF APPROVED (ENGnEEnWG Ano PLANT STAPP) DEsMN CNMGES, DunNG TM 1

CNm0E nFLEnENTATIoN AUTNonGATKw PMOCESS. ThE PMoPoJ83 CNANGE To TM SAR n To AN ADnenSTMATIVE PnoCESS TNAT NAS NO ORECT OM #CenECT nFACT CN H oPERATKw, CoNTMoL oM FUNCTNw of ANY GGNS PLANT EoJFenNT, Ano THetEPonE neLL NOT APPECT THE SAFETY OF opatATMMS. DESSN CNmGES UNDEnGo ALARA DEmeN nEwEwOUnNG THE CDELoPAGENTPnoCESSN ACConaNr2 wrTN DEmeN ENonamnwe AcnenSTMATME PMOCEDUMES, Ano EACN wonn onnEn THAT InoLEneGNTS A MANT Cnmes on neoonCATnw ss nEntwEn av APPnoPntATE MAou TKw CONTnotAMALTH 3

a PHYmCSPersonnEL Fon ALARA CONCEnNS. M PnoPosEO CNmGE CAAWOTNCnEASE THE PRoSAGEJTY i or oCCUnnENCE on CoustoMNCES of AN ACCcENTPREwovsLY EVALUATED N THE SAR, 909 EASE THE l

t PMo8A8KITY of OCCUnnENCE oR CONSE%lKMNCES of A MAU%MCTloN of ECWnENT nFoMTANT TO SAFETY PnewCUsLY EVALUATED N TM SAR, CnEATE TM PosanUTY of AN ACCDENT OR MALPWCTKw of EQUFMENT nePORTANT TO SAPETY OF A DPPmRENT TYPE THAN ANY PREwCUSLY EVALUATED N THE SAR, j.. ANO wiLL NOT REDUCE THE neAMON of SAPETY AS DEPNED N THE BASIS Pon ANY TECMwCAL SPECMCATxw.

i j II. Safety Evaluation on Not apphcable per Safety Evaluation /P=- ng/Applicatulity

Review 4

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{ A. Tachnical SW&=4rs c"8 l

1. Impla===tariaa or perfonnance of the action described in the evaluated Aae==ie will O Yes require a change to the GGNS Unit 1 Technical trardea'ia= E5 No

BASIS: ThE PnoPoaa> CNmet To THE SAR IS TO EuenNATE A nEntacANT nEwEw of APPn0VEo OEmeN CNmeCS DUnne THE nFLEnnNTATnw AuTNoneATxw Pn0 CESS aY INE RAcnATnw PROTECinow MAnneER. M TECmnCAL SPECnCA1Kws 00 NOT ADDnESS Tw ALARA PM00MAnt

• i on ITS notaMENTATKw PMOCESsES AND nEQUREnGNTS. M TECm0 CAL SPECMCATKwt 00 neount GGNS GanRAL MANAGER AUTNoneATxw PoM nePLEnnNTATxw 09 NUCLEAR SAFETY nELATao creeN CNeets, sur 00 NOT orSCUSS OTNEn MEwEw on APPMOVAL nEQUMEMENTS PoM l.

4 CHmeCS. TumaEPonEt A asANos To rw TECHnnCAL SPECMCATNwsIS NOTMEQUREO e e 0

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IMPLEMENTATION OR FEAPORMANCE OF THE ACHON Dur'aram IN THE EVALUA11D DOCUnElfr:

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1. May increase the probability of occumace d an accident previously evatusend in tiz SAR.

O Yes

! E NO l BASIS: M PROPOSEO CNANGE TO H SAR n TO nneuTE A MEntnEuNT MEveW SY THE i RAcun0N PMOTECn0N MANAGEM of APPROVE 0 (ENQNEERNQ AND PLANT STAFF) DES l

CNMGES DuMNQ THE CHANGEnePLBGEWTADON AUTHONGADON PROCESS. M PM0 POSED CNMGE TO THE SAR IS TO AN ADeenSTMnVE PROCESS TNAT HAS NO CMECT OM NDNIECT nePACT ON ACCDENT ANALYSES N THE SAR, AND THEREPOM DOEE NOT APPECT N ANY WAY THE PROBASE.lTY l OP OCCUMMENCE W ANY ACCDENT PREwCU8LY EVALUAT80 N TM SAR.

i j 2. May increase the consequences d an aawlaat primously evalueesd in the SAR. O Yes

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E NO y BASIS: M PM0 POSED CHANGE TO M SAR IS 70 nn8NATE A MED(MDANT MEVEW SY THE

RADtAn0N PROTECDON MANAGEM OF APPROVED (ENGNEEMNQ Ano PLANT STAFF) DECnN CHANGESDUMNQ THE CHMGEnePLEMENTADON AUTHONGADONPROCESS. MPM0 POSED CNMGE

To THE SAR IS TO AN ADenBSTMnVE PM0 CESS. M AbhRA PROGRAM MEVEWIS NTENDED 70

' MANTAN MotAn0N ENPOSURES DURNG NOMseAL PLANT OPEMn0N AS LOW AS NEASONAnY ACHevAnt, ANO NAS NO I:oACT ON DOSE CONSE0wNCES DunNQ ACCDENr8 EVALUATED N THE l

i SAR.

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May increase the probability c(occurrence af a malfunction af oquipaust important to O Yes

' safety previously evaluated in the SAR.

E NO

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BASIS: M PROPOSea CHANGE To The SAR n 70 EneNATE A neolmmWT MEvnw sY THE b RActAn0N PROTECDON MANAGEM OF APPROVE 0 (ENGNEERNQ AND PLANT STAFF) D 4

CHANGES, OUMNQ nE CHANGE lhePLEneENTAn0N AUTNORGAn0N PROCESS. M PROPOSED i

CHANGE TO THE SAR IS TO AN AcaenSTMnvE PMOCESS THATNAS No ontECT OM NoutECT nePACT i

ON THE OPEMn0N, CONTMOL OR FUNCDON OF E00FanNT nePORTANT TO SAFETY. TkEMEFOM, THE PROPOSED CHANGE CAhNOT NCMEASE H PROSABEJTY OF OCCURRENCE OF A MALF;MCn0N OF l EQUIPMENTIMPORTANT TO SAFETYPREHOUnY EVALUATED N H SAR.

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May increase the consequences of a =3+ o(equipment important to safety 0 Yes '

I previousiv evaluatedin the SAR.

E No l BASIS: M PMoPOSto CNMGE TO THE SAR IS TO nneNATE A ME00muMT MeMEW sY THE 1

RActADON PMOTECn0N MANAGEM OF APPROVED (ENGNEEMNQ AND PLANT STAFF) DESIGN CHANGES, DUMNG H CHANGE lhePLEheENTADON AUTHOMGADON PROCESS. M PROPOSED

\ CHANGE TO THE SAR IS TO AN ADMNSTMnVE PROCESS THAT NAS NO DeMECT OM NotMCT nePACT ON THE OPEMDON, CONTMOL OR FLNCDON OF E0l9n8ENT ne90MTANT TO SAFETY. MMEFORE,

• THE PROPOSGO CHANGE CANNOT NCMEASE H CONSEOMNCES OF A MALFLNCDON OF EOUtPMENT IMPORTANT TO SAFETYPMHOUnY EVALUATED N THE SAR.

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May create the possibility for an accident of a ddferent type then any previously 0 Yes

} evaluated in the SAR.

i E No BASIS: M Pnoposso CNmeE To nm SAR a To ausaNATE A MEoWonNT ngMEw sY THE 1

RActanoN PnorEcnoN MANAoEn of APPnoken (ENonnEnne AND Pt. ANT STAW) cesmN cNmets, OURNe THE CHANet apLEnnENTAnoN AUTNoneAnoN PnocEss. M PnoPosto

CNmet To THE SAR m To m AceannsTnAnvf PnoCEss TNAT Nas No onocT on moutEcT nnPAc7 ,

ON THE OPEnAnoN, coNTnot on FWCnoN of ANY GGNS PLANT EourneENT. MaEFont, THE l

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PMoPosto CMmeE CANNor cnEATE THE PosaaUTY Pon m ACCDENT of A 0FFERENT TYPE THAN mvPnEwousLYEVALUATEDN THESAR. ,

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, 6. May create the possibility for a malfunction af equipsment important to safety of a O Yes different type than any pnmously evaluated in the SAR. .

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E No BASIS: M PnoPosEO CNnNet To THE SAR a To SundNnFE A nSDWo4NT nEVIEW sY THE i RActAnoN PMoTECnoN MANAGER of APPnov50 (ENaNEEnwe ANO PILANr STAPP) DEa cNANets, OUnNe THE CNmet n89URGENTAnoN AUmoneAnoN PnoCEss. M PnoPo380 i cNmet To THE SAR m To AN AcaansTMnvE PnocEss rNArnas No annecT on monEcr anPAcT ON THE OPEMnoN, coNTnot on FUNCnoN of ANY GGNS PLANT soWanENT. ThEnEPont, THE j PnoPosto CNANot CAhNOT cnEA TE 1>d PossnUTY Pon A hlALPWCnoN of toWanENT nWonTANT To SAFETY OF A crPEnENT TYPE THM ANY PnEWousLY EVALUATED W THE SAR.

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!' 7. Will reduce the margia of safety u amanad in the BASIS: for any Technical 1

! Speci&=*i-O Yes I i( E No l BASIS: THE Pnoposto cNmot To THE SAR a To tuneN4TE A nSoWomT ndMEW sy THE  !

l RAotAnow PnoTucnoN MANAetn of APPnovEn (ENonEneNe Ano PLANT STAW) cesmN CHANGES, OURNG TFM CHANoE nePLEnNNTAn0N AUTHontAnoN PnoCEss.

l M PnoPosto i 1 CHANGE To THE SAR Is To AN AonenwsTunvt enoCEss THATNAs No onRECT on NotatcT anPACT l

ON THE OPEMnoN, CoNTnot on FWCnoN of ANY GGNS PLANT EoWRENT. THEntFont, THE PnoPosEc CHANGE CAhNOT REoUCE THE hiAMGN of SAFETY As cERNeo w THE BASIS Fon ANY l TECHMCAL SPECmCAnoN.

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DeFLEMENTATION oR FERFoRMANCE OF THE ACnoN DESCRIBED IN THE EVALUATED DoCUh0NT

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i l 1. Will reqmre a chmage in the Enviran===*=1 Protection Plan. U Yes

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BASIS: M PMoPOSED CHANGE To THE SAR MEs MtTH THE MEVWW 0F DESGN CMNGEs ANO uoanCAnous av TM RAonAnoN PMoTECnoN MANAGER FoM ALARA CONuoEMTKws, ANO Is UNMELATED To THE ENWRONuGNTAL PMoTECnoN PLAN oM THE FNAL ENHMONRGENTM STATEh8ENT.

MPMoPosEO CHANeuNNo WAYAMECTS oMNAs THEPoTENnAL To A99ECT THEENMRoNMENT.

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l B. Unreviewed Environmental 0-ion **

1.. Concerns a maner which may result in a significant increase in any adverns

! O Yes

' ennranmenemi impact previously evaluated in the Final Enviremmental Statement (PES) as modified by the NRC staffs ==a=y to the Aeoaiic Safety and Liceamag Bonni E No (ASLB), supplements to the PES, canrnama=ml impact appresal, or la any decimons of l the ASIJB.

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BASIS: M PMoPoSED CHANGE To THE SAR CErs NTH THE MeMEW OP 0830N CHANGES ANO MoonCAnoNs BY TM MAomnoN PMoTECnoN MANAGEM FoM ALARA CONuoEMnoNs, ANO Is UNMELATEo To THE ENwMoNRGENTA PMoTECn0N PLAN oM THE FNAL ENwMoNanENTAL STATEMENT.

TM PMoPOSEO CHANGE N No WAY AFFECTS oM mas H PoTENnAL To AFFECT TM EMAMoNheENT.

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2. Concerns a significant change in efRucats or power level O Yes E No 1 BASIS: M PnoPoSEC CHAN3E To Tw SAR Mrs nsTH TM nEviEW OF DEsGN CHANGES ANO MorwnCAnoNs av THE RAoum PMoTECnow MANAGEM Fon ALARA CONuonunous, ANO Is UNMELATED To THE REACToM oPEMTKws oM CONTMols AND CANNOT NAVE ANY AWECT ON EFFLUENTLEYELsFMoM THEPLANT.

3. Concerns a matter not previously reviewed and evaluated in the documents specified in O Yes III.B.1 above, which may have a significant environmental impact. E No ,

BASIS: M PMorosEc Cno:et To THE SAR DErs NTH THE MenEW or DEsMN CHANGES AhD MoonCAnoNs sY THE RADMnoN PMoTECnoN MANAGER FoM ALARA CONsn0EMnoNs, ANO is UNMELATED To THE ENwRoNheENTM PMorECnoN PLAN oM TM FNAL ENwRoNheENTAL STATEh8ENT.

Tw PMorosEO CHANos N No WAY AFFECTs oM NAs THE PoTENnAL To AWECT TM ENwMoMufNT.

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%-065- HP GRANU GULF NUCLEAR STATION b csr:c:.; i ADMINISTRATZVE PROCEDURE

• F *- MIG 3 3 01-S-06-24 Revision: 102 fwip:s S!H q4 caz 1o/1/4 /. Attachment V Page 1 of 4 INN"lUn GRAND GULF NUCLEAR STATION UNIT 1 CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM I. Safety Evaluation Overview Page __ of__,

A. Reference Data OlUOINATOR: BRIAN D. PATRICK DEFT / SECT: HP / DOSIMETRY EVA!> #: 96-0069-R00 l 1

DOCUMENT EVALUARD: UFSAR SECT.12 (SEE ATTACHED)  !

1 SYSTEMNO:(S) N/A i REFERENCES FSAR CHANGE REQURED7 X Yes a No CR # 96 091 FSAR SECTIONS TO BE REVISED: 12.1.3.1.t.,12.5.2.1.1,12.5.2.2.3, TABLE 12.5-1

• O Yes TRM CHANGE REQURED7 X No l i

TECH. SPEC.CHANGEREQUEED7

• O Yes X No CR # N/A IS THE VAUDITY OF DGS SAFETY EVALUADON DEPENDENT ON ANY CHANGESOOTHER Yes MAN TIE CHANGE BEING EVALUATED (E.O. PROCEDURAL, OPERATIONAL CONDfTIONS)?

X No l

EXPLAIN:

i IF YES TO THE LAST QUESDON, HAVE THE ORGANIZADONS RESPONSIBLE FOR THOSE CHANGES O Yes l BEEN NOTIFIED 7 l

(THE RESPONSIBLE ORGANIZATIONS MUST BE NOTIFIED PIUOR TOIMPLEMENTING THIS CHANG Signatures and Approvals of Attached Safety and Environmental Evaluation

// ~

~ Evaluated:"23 g) pg (PRINT NAME)

ORIGINATOR N

)

'(3DNATURE)

, g,,pp ,

DATE Reviewed: "*) g yc f g,,, , ,

/ f[27./'96 (PIUNT NAME) INDEPENDENT REVIEWER (SIONATURE) DATE Reviewed: "* y,, f (PIUNT NAME) OTHER REVIEWER (IF REQURED) (SIGNATURE) DATE Plant Safety Review Committee Review 05)

$}Qa O ,, 9

{ C SRC DATE J:\ADM,SRVS\ TECH _ PUB \ REVISION \1\1S0624.A5 1

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• ADMINISTRATIVE PROCEDURE  !

1 01-S-06-24 Revision: 102 Attachment V Page 2 of 4 l

SE Ho. f$~$0$f- 80 0 Page X of 4 I

B. Executive SummMY (ALSO SERVES AS INPUT TO NRC

SUMMARY

REPORT)

BRIEF DESCRIPTION OF CHANGE, TEST OR EXPERIMENT INCORPORATES T1lE USE OF ELECTRONIC ALARMING DOSIMETERS (EADS) INTO THE UFSAR.

REASON FOR CHANGE, TEST OR EXPERIMENT ALLOWS THE USE OF EADS AS MONITORING DEVICES FOR PERSONNEL SAFETY EVALUATION

SUMMARY

AND CONCLUSIONS SEE ATTACHED ACCEPTANCE TESTING REPOR ELECTRONIC DOSIMETER / n.D COMPARISON REPORTS l

II. SSfety EvSluStion M Safety Evaluation not required per Completed Safety and Environmental Pre-Screening or Applicability Review. Proceed to Section III for Environmental Evaluation.

A. Technical Soecifications

• 1.

Implementation or performance of the action described in the evaluated document O Ye8 will require a change to the GGNS Unit 1 Technical Specifications. X NO BASIS: TECH SPECIFICATMW NUMBER 6.7.1 (SEE ATTACHED) REQUIRES THAT ANY INOMOUAL OR GROUP PERRAITTED TO ENTER A HIGH RADIATION AREA SHALL BE PROVIDED WITH ONE OR MORE OF THE FOLLOWING:

• A RADIATION IBONITORWG DEVICE THAT CONTWUOUSLY WDICATES THE RADIATION DOSE RATE THE AREA

• A RADIATIoM 000NITORWG DEMCE THAT CONTWUOUSLY INTEGRATES 1NE RADIATION DOSE RA THE ARLa AND ALAR 198 WHEN A PRESET INTEGRATED DOSE IS RECEIVED...

ISSUING ALL WORKERS THAT ENTER RAD 60LOGl?. ALLY POSTED AREAS EADS WILL SATISFY THIS REQUIREMENT BECAUSE IT NTEGRATES DOSE AND ALARMS WHEN A PRESET UtslT IS RECEIVED.

B. Unreviewed Safety Ouestion "

!MPLEMENTATION OR PERFORMANCE OF THE ACTION DESCRIBED IN THE EVALUATED DOCUMENT.

1.

May increase the probability of occurrence of an accident previously evaluated in the O Ye8 SAR. . X No

> BASIS:THE USE OF EADS TO R$0NITOR PERSONNEL WORKING INSIDE RADIOLOGICALLY POSTED AREAS WILL NOT EFFECT ANY SAFETY RELATED EQUIPMENT OR OTHERWISE WCREASE THE

(

i PROsAaluTY OF AN ACCIDENT THAT HAS BEEN EVALUATED W THE san. THE EAD SYSTEtt WORKS INDEPENDENTLY OF ANY OTHER EXISTING SYSTEnd.17 WILL NOT AFFECT THE OPERATION OR l FUNCTION OF ANY OTHER SYSTElf, SAFETY OR OTHERWISE.

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Attachment V Page 3 of 4

2. May increase the consequences of an accident previously evaluated in the SAR O Yes.  ;

BASIS: TuE USE Or mads wtLL NOT INCREAaE ANY CONSEQUENCES. THEY w!LL X No SIMPLY BE AN ENRANCEMENT,380ULD AN ACCIDENT OCCUR, OVER THE USE OF i POCKET ION CHAMBERS BECAUSE OF THEIR ABILITY TO GIVE DOSE AND RATES ALTERNATELY AND ALARM IF EITHER REACHES A PREBET LIMIT.

3. May increase the probability of occurrence of a =alam ofequipment unportant to O Yes s safety previously evaluated in the SAR.

X No BASIS: THE USE OF EADS WILL NOT EFFECT ANY SAFETY REIATED EQUIPMENT. THE EAD  !

SYSTEM WILL OPERATE INDEPENDENTLY OF ANY EXIFTING SYFTEM, SAFETY OR OTHERWISE.  ;

1

4. May increase the consequences of a malfunction of equipment unportant to safety 0 Yes previously evaluatedin the SAR.

X No BASIS : THE USE OF EADS WILL HAVE NO EFFECT ON THE CONSEQUENCES OF A MALFUNCTION i OF EQUIPMENT IMPORTANT TO SAFE 1Y. IT WILL BE AN ENHANCEMENT OVER THE USE OF l POCKET ION CHAMBERS DUE TO THEIR ABILTTY TO DISPLAY DOSE AND RATE AND AIARM WHEN A PREDETERMINED LIMTT IS EXCEEDED.

5.

May crease the possibility for an accident of a differait type than any prevuxr,1y 0 Yes evaluatedin the SAR.

X No BASIS: THE USE OF EADS AS PERSONNEL MONITORING DEVICES INSTEAD OF POCKET ION CHAMBERS WILL NOT CREATE ANY POSSIBILTIY FOR AN ACCIDENT OF A DIFFERENT TYPE T11AN ANY PREVIOUSLY EVALUATED IN THE SAR DUE TO THE FACT THAT TT OPERATES INDEPENDENTLY OF ALL OTHER SYKTEMS.

. i

6. May create the possibility for a malfunction of equipment unportant to safety of a O Yes different type than any previously evaluated in the SAR. X No BASIS: EADS WILL NOT EFFECT EQUIPMF.NT IMPORTANT TO SAFETY OF A DIFFERENT 1YPE THAN ANY PREVIOUSLY EVALUATED IN THE SAR DUE TTS INDEPENDENT OPERATION.

7. Will reduce the naargin of safety as defined in the BASIS: for any Technical O Yes Specification.

X No BASIS: THE UsE OF EADs WILL NOT REDUCE THE MARGIN OF SAFETY DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATIONS. IT WILL BE AN ENHANCEMENT TOR TECH SPEC 5.7.1 (SE ATTACHED) DUE TO ITS ASEJTY TO DISPLAY DOSE AND RATE WITH A SNGLE INSTRUMENT AND THEREFORE WIU. WCREASE THE MAROM OF SAFETY.

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GRAND GULF NUCLEAR STATION ADMINISTRAT2VE PROCEDURE 01-S-06-24 Revision: 102 Attachment V Page 4 of 4 III. Environamental EValuatioR*) O Environmentai tvaivation not required per Completed Safety and Environmental Pre-Screening or Applicability Review.

IMPLEMENTA110N OR PERFORMANCE OF THE ACTION DESCRIBED IN THE EVALUATED DOCUMENT A. Environmental Protection Plan *)

i 1. Will require a change in the Environmental Protection Pir.n. U Yes BASIS: THE USE OF EADS 10 MONITOR PERSONNEL RADIATION DOSE WILL o HAVE NO 5FFECT ON THE ENYlRONMENTAL PROTECTION PLAN.

B. Unreviewed Environmental Owion *)

1.. Concerns a matter which may result in a significant increase in any adverse environmental i

impact previously evaluated in the Final Environmental Statement (FES) as modified by O Yes the NRC staffs tesnmony to the Atomic Safety and Licensing Board (ASLB), X No

+

supplementr to the FES, environmental unpact appraisal, or in any decisions of the ASLB.

BASIS: THE USE OF EADS TO MONITOR PERSONNEL FOR RADIATION DOSE WILL NOT CONCERN A MATTER WIDCH WILL HAVE AN ADVERSE ENVIRONMENTAL IMFACT PREVIOUSLY EVALUATED. EADS ARE USED EXCLUSIVELY FOR MONITORING PERSONNEL FOR EXTERNAL RADIATION.

2. Concerns a significant change in effluents or power level. O Yes BASIS: THE USE OF EADS FOR PERSONNEL MONITORING WILL HAVE NO EFFECT X No ON EFFLUENT DISCHARGES OR FOWER LEVEI. IT IS AN LNDEFENDENT SYSTEM.

3. Concerns a matter not previously reviewed and evaluated in the documents specified in O Yes III.B.I above, which may have a significant environmental impact.

X No BASIS: THE USE OF EADS DOES NOT CONCERN ANY MATTERS OF ENVIRONMLNTAL SIGNIFICANCE WIUCH WERE NOT ADDRESSED ABOVE.

J \ADM_SRVS\ TECH _ PUB \ REVISION \1\1S0624.A5

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e.t caMcca3 u.s rue. 01-S-06-24 Revision: 102 t.se-Ei.:s c si a n 3 Attachment V Page 1 of 4 MJUBEL d/A 4

l l GRAND GULF NUCLEAR STATION UNIT 1

] CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM I. Safety Evaluation Overview Page _ of _

i A. Reference Data et  ;

ORIGINATOR: BRIAN D. PATRICK DEFT / SECT: HP / DOSIMETRY EVAL #:

96-0069-R99-[

DOCUMENT EVAWAED: UFSAR SECT.12 (SEE ATTACHED) l SYSTEM NO:(S) N/A REFERENCES FSAR CHANGE REQUIRED 7 X Yes O No CR # 96-091 FSAR SECDONS M BE REVISED: TABLE 12.5-1 8)

TRM CHANGE REQUIRED 7 O Yes X No TECH. SPEC. CHANGE REQUIRED 7

• O Yes X No CR # N/A IS mE VAUDTTY OF MIS SAFETY EVAWADON DEPENDENT ON ANY CHANGES OMER MAN O Yes THE CHANGE BEING EVALUATED (E.O. PROCEDURA1, OPERADONAL CONDTTIONS)?

X No EXPLAIN:

IF YES TO THE LAST QUESDON, HAVE THE ORGANIZADONS RESPONSIBLE FOR THOSE CHANGES O Yes BEEN NODFIED7 (mE RESPONSIBLE ORGANIZATIONS MUST BE NODFIED PRIOR TO IMPLEMENTING THIS CHANGE.)

Signatures and Approvals of Attached Safety and Environmental Evaluation Evaluated: "2) gg g gMg ,[h fp.22-f6

- (PRNT NAME) ORIGIAKTOlt "T/ (SIGNATURE) DATE Reviewed:"*) g, gf,y ,j f 9-/a g (PRNT NAME) INDEPENDENT REVIEWER (SIGNATURE) DATE Reviewed:"* y,, f (PRINT NAME) OTHER REVIEWER (lF REQUIRED) (SIGNATURE) DATE

! Plant Safety Review Committee Review 05)

Cm C , PSRC 4-t *Z.~9 Q DATE l

l J:\ADM_SRVS\ TECH, PUB \ REVISION \1\150624.A5 ,

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Page fd = COG 9=){

of g

.g B. Executive Summyy (ALSO SERVES AS INPUT TO NRC

SUMMARY

REPORT)

BRIEF DESCRIPTION OF CHANGE, TEST OR EXPERIMENT I

EVALUATES THE USE OF ELECIRONIC ALARMING DOSIMETERS (EADS). MODIFYS TASLE 12.5-1 SEE ATTACHED REASON FOR CHANGE, TEST OR EXPERIMENT ALIAWS 11tE USE OF EADS AS MONITORING DEVICES FOR PERSONST.L SAFETY EVALUATION

SUMMARY

AND CONCLUSIONS SEE ATTACHED ACCEPTANCE TESTING REPORT AND ELECTRONIC DOSD4ETER / TLD COMPARISON REPORTS H. Safety Evaluation p Safety Evaluanon not required per Completed Safety and Environmental Pre-Screening or Applicability Review. Proceed to Section 111 for Envronmental Evaluation.

A. Technical SDecifications *

1. Implementation or performance of the action described in the evaluated document O Yet I will require a change to the GGNS Unit 1 Technical Specifications. X No BASIS: TECH SPECFICATION NUMBER 6.7.1 (SEE ATTACHED) REQUNtES THAT ANY WDIVIOUAL OR OROUP PERMITTED TO ENTER A HIGH RADIATION AREA SHALL BE PROVIDED WITH ONE OR MORE OF THE FOLLOWWG:

• A RADIATION MONITOfUNG DEVICE THAT CONTWUOUSLY INDICATES THE RADIATION DOSE RATE IN THE AREA e A RADIATION MONITORWG DEVICE THAT CONTINUOUSLY WTEGRATES THE RADIATION DOSE RATE IN THE AREA AND ALARMS WHEN A PRESETINTEGRATED DOSE IS RECEIVED .

lSSUWG ALL WORKERS 7 HAT ENTER RADIOLOGICALLY POSTED AREAS EADS WILL SATISFY THIS REQUIREMENT SECAUSE IT WTEGRATES DOSE AND ALARMS WHEN A PRESET UMIT IS RECEIVED. I B. Unreviewed SafetV Ouestion "  !

IMPLEMENTA110N OR PERFORMANCE OF THE ACnON DESCRIBED IN THE EVALUATED DOCUMENT:

1. May increase the probability of occurrence of an accident previously evaluated in the O Yes SAR. X NO BASIS: THE USE OF EADS TO a00NITOR PERSONNEL WORKING WSIDE RADIOLOGICALLY POSTED AREAS WILL NOT EFFECT ANY SAFETY RELATED EQUIPMENT OR OTHERWISE WCREASE THE PROSAB8UTY OF AN ACCIDEidT THAT HAS BEEN EVALUATED IN THE SAR. THE EAD SYSTEM WORKS INDEPENDENTLY OF ANY OTHER EXISTING SYSTEM. IT WILL NOT AFFECT THE OPERATION OR FUNCTION OF ANY OTHER SYSTEM, SAFETY OR OTHERWISE.

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Attachment V Page 3 of 4 2.

May increase the consequences of an accident previously evaluated in the SAR. O Yes BASIS: THE UsE CF EADS WILL NOT INCREASE ANY CONSEQUENCES. THEY X NoWILL

' SIMPLY BE AN ENHANCEMENT, SHOULD AN ACCIDENT OCCUR, OVER 11fE USE OF POCKET JON CHAMBERS BECAUSE OF THEIR ABILITY TO GIVE DOSE AND RATES ALTERNATELY AND ALARM IF EI11tER REACHES A PRESET LIMIT.

4 -

3.

May increase the probability of occurrence of a malfunction of equipment important O Yes

. to safety previously evaluated in the SAR.

X No BASIS: THE USE OF EADS WILL NOT EFFECT ANY SAFETY RELATED EQUIPMENT. THE EA SYS11M WILL OPERATE INDEFENDEN11.Y OF ANY EXISTING SYSTEM, SAFETY OR OTHERWISE.

4.

May increase the consequences of a malfunction of equipment important to safety 0 Yes l previously evaluated in the SAR.

X No BASIS

THE USE OF EADS WILL HAVE NO EFFECT ON Tite CONSEQUENCES OF A MALFUNCT OF EQUIPMENT IMPORTANT TO SAFETY. IT WILL BE AN ENHANCEMENT OVER THE USE OF

] POCKETION CHAMBERS DUE TO THEIR ABILITY TO DISPLAY DOSE AND RATE AND ALARM WHEN A PREDETERMINED LIMITIS EXCEEDED.

5.

May create the possibility for an accident of a different type than any previously 0 Yes evaluated in the SAR.

X No BASIS: THE USE OF EADS AS PERSONNEL MONTTORING DEVICES INSTEAD OF i POCKET ION CHAMBERS WILL NOT CREATE ANY POSSIBILITY FOR AN ACCIDENT OF A DIFTERENT TYPE TRAN ANY PREYlOUSLY EVALUATED IN 11tE SAR DUE TO THE FACT THAT IT OPERATES INDEPENDENTLY OF ALL OTHER SYSTEMS.

A i 6.

j May create the possibility for a malfunction of equipment impor: ant to safety of a O Yes different type than any previously evaluated in the SAR.

l X No BASIS: EADS WILL NOT EFFECT EQUIPMENT IMPORTANT TO SAFETY OF A DIFFERENT TYPE THAN ANY PREVIOUSLY EVALUATED IN THE SAR DUE ITS IhDEFENDENT OPERATION.

i

7. Will reduce the margin of safety as defined in the BASIS: for any Technical Specification.

O Ye5 X No BASIS: THE USE OF EADS WILL NOT RE3UCE T1tE MARGIN OF SAFETY DEFINED IN THE BASIS j FOR ANY TECHNICAL SPECIFICAT10NS. fT WILL BE AN ENHANCEMENT FOR TECH SPEC 6J.1 ATTACHED) DUE TO ITS ABILITY TO DISPLAY DOSE AND RATE WITH A SINGLE INSTRUMENT AND THEREFORE WILL INCREASE THE MARGIN OF SAFETY.

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GRAND GULF NUCLEAR STAT 3ON

, ADMINISTRATIVE PROCEDURE 01-S-06-24 Revision: 102 Attachment V Page 4 of 4 III. EnvironamentalEvaluation") Environmental Evaluation not required per Completed f

Safety and Environmental Pre-Screening or Applicability '

Review.

IMPLEMENTAllON OR PERFORMANCE OF THE ACrlON DESCRIBED IN THE EVALUATED DOCUMENT. .

A. Environmental Protection Plan *

1. Will require a change in the Environmental Protection Plan.

U Yes o

BASIS: THE USE OF EADS TO MONTTOR PERSONNEL RADIATION DOSE WILL HAVE NO EFFECT ON THE ENVIRONMENTAL PROTECIlON PLAN.

I l

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B. Unreviewed Environmental Oteion "

l l 1.. Concerns a matter which may result in a significant increase in any adverse environmental j impact previously evaluated in the Final Environmental Statement (FES) as modified by O Yes the NRC staffs testunony to the Atomic Safety and Licensing Board (ASLB), x No supplements to the FES, environmental impact appraisal, or in any decisions of the ASLB.

BASIS: THE USE OF EADS TO MONITOR PERSONNEL FOR RADIATION DOSE WILL NOT CONCERN A MATTER WHICH WILL HAVE AN ADVEltSE ENVIRONMENTAL IMPACT PRE %10USLY f

I EVALUA1ED. EADS ARE USED EXCLUSIVELY FOR MONITORING PERSONNEL FOR EXTERNAL RADIATION.

l l 2. Concerns a significant change in effluents or power level.

! O Yes BASIS: THE USE OF EADS FOR PERSONNEL MONITORING %1LL HAVE NO EFFECr X No ON EFFLUENT DISCHARGES OR POWER LEVEL. IT IS AN INDEPENDENT SYSTEM.

3. Concems a nutter not previously reviewed and evaluated in the documents specified in O Yes Ill.B.1 above, which may have a significant environmental impact.

X No BASIS: THE USE OF EADS DOES NOT CONCERN ANY MATTERS OF ENVIRONMENTAL l SIGNIF1CANCE WHICH WERE NOT ADDRESSED ABOVE.

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/A/7/96 '

uu=gr GRAND GULF NUCLEAR STATION UNFF 1 Mr , _55i OR ExFERIMENTS SAFETV AND ENVIRONMENTAL EVALUATION FORM i

L Safety Evaluation Overview Page 1 of jlL, A. Reference Data I onMDIATOR: MutCS MKl.Et DErr/sscr: Hzu.m.*nmcr EVAL.s:96-0071-R00 DOCUMENT EVAWARD: QDR 0144-96 SYsHMNO:(S) N/A ( INGERT N/A IP Not Appucasta) 1 nrERENCEs: N/A l FSAR CHANGE REQUUtED7 m Yes a No cA # 96 100  !

FSAR sECTHus To at revised: 12.2.1.6 I 1

TRM CHANOE REQUDLED7 a Yes E No  !

j t

TECH. SPEC. CHANGE uQUIRED7 0 Yes a No cA# Me Is m VAuDrry Or nas SAFE 1Y EVAWADON DEFENDENr ON ANY CHANOES OMER THAN O Yes m CHANGE BEDIO EVAWATED (E.G. PROCEDURA1, OPERADONAL CONDm0NS)?

gg ExPLAm: TNatr AAf M Pat-A8QURED CHAAGES m PAOCEDUREN, DC# SPECT, I armurmacocauxNis, anAwrarmurDe conDmans.

{

Ir vEs 70 m usr QUunow, HAvE m ORGANIZADONS RESPONNELE FOR1 HOSE CHANOEs O Yes BEEN Nor!EED7 N/A l

(m RuroNaraLE ORGANIZADONS IWUST BE NOTIFIED PRIOR 1D DOUMDmNG THE CHANOE.)

l l Signatures and Approvals of Attached Safety and Environinental Evaluation EvRluated: g g 4 *g '

, gA f 9,.gg (PRDrr NAMEf ORIGINATOR (830NA1URE) DAll Reviewed: hk T,% , j h/g%

i (PRINT NAME) INDEFENDENT REVIEWER (330NA1URI) DATE Reviewed: yk , f (PRDrrNAME) OT'f.R REVIEWER (IF REQUUtED) (530 NATURE) DArd Plant Safety Review Committee Review QA .n y qf &

CHADLMAN C DAM i

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l SE No. 96-0071-R00 Page_2. af 6 I

! B. Exe*ive Summarv (Auo SERVES As omrr To NRC

SUMMARY

REPORT) l BlUEF DEsCRFnON OF CHANGE, Err OR EXPERBaDrr

! of raxfoedhe materief h aroes odside me con 6cCod access area, protoded area, l and structwes. These areas hciude me warehouses, onste merking area buidngs, storage arees outside of plant strudwes bd heide me protected area, Unit H, and storage

} buildings W redoecitwo components such as the High Presswo and Low Pressure Turbines.

This evaluation aino indudes a short dura 60n holdng aree Wredoedhe weste in kansmon l Rom processing to konsporthy ohNo.

REASON FOR CHANOE, TEST OR EXPE3SEDrr I l

1 This material may anE how useM nk and k not behg decarded, or is swaning thaf re l vnemnon. h the cose of redoncthe waste, t in M a Vanatory period between processing i andshepha andis M #: that shephp contahors. The normal dration

the weste spends M en area outsch or me protected aren is uss man one week, j we me at expected of abod two weeks. The e=vaan to mis is me wwege

sludge is beme new tbr decey and subsequent emtonmental deposmon, and is bekw any pondne amus wnhin the axe boundary. The normal dwanon me waste

! spends me protected area hoidhe area is has man one cycde.

j SAFETY EVALUADON sUh64ARY AND CONCLUSMMS

sedian 12.2.1.0 of me SAR descrees me storage of materief odain of plant structures, and l 1 states met adequate measures are to be taken by heaEh physics W materialis to '

i be stored outside of stuctures. SAR Secdon 11.4.7 descrees the use of sh@php l contahars and outside stuctaaes bd hside the restided area (protected

area). The restided area at GuWis denned k severaf Econee basis documents, l boeudha Todanical Specdcanons and en SAR. For me pwposes of todadon protec60n, he j restided area is the same as me protected area, as outtnedh Secdon 2.1.1.3 of the as the see boundary b Todanicef i SAR.

Specmce60n's Foremuents Ohne Dose and otherpurposes Caicula00ns Manual Ris cdoerty(OOCM) Appenen A 1.K Furth l CFR 20.1003 delbes a confroned area as being odside of a testined area bd Maide the  ;

sMe boundary, wth conkoned access by the noensee. The storage areas outside the

}

protoded area are maintahed as conkoned areas, endoned as neoe , Joded, and under

! control of HeaNh Ph Alredoedhe material storage armes of piant stuctures j hsw been as a surwy , and aren 7LDs are used Jbr each storage area to ensure the requhments of 10 CFR .1301 are met. 7he same adnhistadwe, todadon 1 protodion, and secdon procedwes that are used M the contoned access area wit also be 1 usedM the protoded area and 3nkoned erwes outside ofpiant stuctures.

GL80061 and GL81038 ofwes dethElons and s W radoecdwe weste storage tkacindes. The high presswo twbhe rotors and redoedhe materiais M storage anr not weste s>ce they are not being decarded, but may be ghen the same canaldaranons as such ibrthe purposes ofradoecthe materief conkof andregulatory comp 0ance.

10 CFR 20.1301(s) descrbes the dose and dose rate Ents to members of the pub 0c. SAR

\ secdon 2.1.2.1.1 paragraph 3 states that Enterw OperaNone wit a00w access to parts of the

! sne odside of the exclusion area. 10 CFR 20.1301(b) states that if members of the ,

i are perm 6hd access to contmEed areas, the nmts of 10 CFR 1301(a) anE apply, i Since Wese conkoned areas are enc 60 sed and Joded under the conkof of HeaNh Physics, 1 pubEc access is contoned. The dose rates at the contoned area tience are below the Emts of 10 CFR 20.1301(a), mereby ensuring the done nmts at the sEe .

l 10 CFR 50.75 ofws requirements Jbr toccrds of hfbrmadon knportant to decommissioning of i

the plant. Surveys of mese areas are currently persbrmed on a monthly basis, and retained j for use Abr decommissionhg 7his is curren#y covered h sodion procedwo 08-3 01-1 11,HeaNh ysics Document Handng and Conkol.

1 M is conciuded that me storage of radandhe materials is acceptab6e outside plant studwes 1 and hside Grand Gut's sNe boundary pnmided that t is M a conto 0ed area or hside the j protoded area.

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SE No. 96-0071-R00

Pap Laf 6 1 t i IL Safety Evaluaties Safety Evanestion not requimd por Completed Sahey and  !

e Envimamental Pre Sassning or ApplPntality Rayww. Pmoned to l jl ection m for Envuommental Evaluatic .

j A. Technical Specifications 4 1. F;'- d= or A -- e of the action described in the evaluesed doomment will u Yes  ;

j regare a change to the GGNS Unit 1 Technical Wh El No l i ,

i l I BASIS: The aforage of racfoedhe maferief ostsMe of the controsed access area or '

profecfed aren k not docussed h the Gramt Gut Tech Spoon sec6an. However, the i requirements of Tech Spoo seasons &&4 anct E7 ads appfy. Als change h the Tech Spec ,

'l Ewould be requirost access area bdas a resut wehh of adorage the ase bouncery. or racioecehe maforfef oWaMe the l

1 1 l B. Unreviewed Safety Ouestion i, I namentrAMON OR PERPGsMANCE OF ME ACDON DESCREED IN MIE IVAWATED DoCWElcr:

)

i 1. May inmonde probability af oocurrenos of an acculent previously evaluated in the O Yes l i SAR. 4 No .

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BASIS: Secdon 12.2.f.6 of the SAR h==* aerect radosodhe mafariale outaMe the l

piant. SecWon f f.4.7 docusses the use of sh@php contahars udhh the reeHofed area i t (profecfed area) bd odattie plant stuchaos. Ins neukt McGude the area outsMe of the j RacBoecehe Weafe Busting truck boy, and other arses heide the profeded aros.

Redoecthe weafe behp theti hsMe the aree bd odade piant adtuttures is

normafy M shtiping or satahip shrope that most the guidednes of GL8trf and 1.

81038.

I Radoedhe wasfe h the radosothe neafe hoking arse octaids of the profeofed area is heti

temporarty pencing sh> ment, usue0y wihh hee weefts. R k k Rnal contahers and 1 readed itir bedtire Jesuing the prohcied arse, confterms 10 T and ARC l spedNcedern, ' y preventhg any unmentored rainene, the twinese of radoecekdy abose l the knts of Tech Spec LK4, and ensu'hg the requirements of 10 CFR 20.1301 are met.

i' 7he prusure turbhe infore and omer radoecebe maferinde h are not weste i shoe are not behg decanied, bd may be ghen the same as docussed

l. M 1 and 01030 itar the pauposes of ragsoecthe maferial contof and i compfence. The raciosothe materiais are packaged, houssct, or encfosed 10 the

. ediscts of and provide confahment of any contamh&nts, provide meferief

contois, anst premature decey of profedhe measures. For arses outsMe the i protoded arran, a contoned area wm De establehed to aEuefe dhe proper boundary and

( ernce be knts 1410 CFR 20.f301 are met. Consequendy, the storage of redoncthe

}

a materird does not aneet orincrease the probabEy of occasrance of any accMent previouefy evaluated h the SAR.

I j 10 CFR 50.75 ghes inquirements Atir reocnis of indtirmedian important do

! decommissioning of the piant. Sanveys of these aroes are cummsyprvtymed on a 3 monthfy baade, ancf refaheaf a use itar piant decommissionirs Ins is surtsnffy j covened h season prooeckse os.01 ff, Henan physics oocument Herushy and i Cantof.

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! SE Ne. 96-00H-R00

Page i. d i.

2. May imorenes the y of an accidset prwously evalasted in the SAR.

O Yes

! o No BASIS: The remocWee matarief kr aforage consists of meterief h soAW Arm and h bw 3 hvain of immecdhey money som sued contamhanan. 7he remeceve mafersain are i pedaged, housed, a enchoed to reduce the esects of weedherhg and prewdde contahment of any contamhants, mafwief coneois, and prevent premerwe deoey of profeme meeswee outhsed M GL80061 and 81030. For the redoeceW matarial storage armes outside the i

?

profected area but mehh the 800 boundary, a contoned area nG be estabEnhed, such as n '

buksng teorrt or Annche wth toded gefes. This afowe access to be contofed by Heath physion, proper aQue6cn of the bee'ery, and ensures me knEn of 10 CFR 20.1301 are met.

, As a resuf, no adveras aSects #are. /s aforage nG ocor, and m8 have tw emoct or heresse

he consequences of an accident previouafy evaluefed h me SAM.

3. May imorense the ,,2 d'y of occunemos of a -IA=*ia= of areipensme important to O Yes safety pewiously evaluated in the SAR. ><

No BASIS: 7here are no system ceniponents or hiportant to sehty used h the aforage of redoac#we materief outade of access areas and profocfod aross, nor are any aSected by this acnon. Therekre, there k no horeene a the probabEy of occurrence of a mammcdon of egukment hiportant to samty prevdounfy evedusfo.1 h the SAM.

4. May increase the cocaequences of a malfunction ofeguir==d important to safety 0 Yes previousiv waluated in the SAR. ><

No BASIS 7te shrage of radosodwo maferials outside of contoAnd access aross ancf -

profeown uroes k outside of piant studwes, equ> ment, and syafees, and does not eSed thew cpera60n k any way. Therekte, there k no heneene h the consequences of a mammcdon of a component or agupment hiportant to samty.

. May croses the possibility for an accidset of a daftsrent type than any seviously D Yes

5. evaluaand in the SAR. El No BASIS: sec6cn 2.4.13.3 of the SAM docusees the nooident en> cts k the event of a ape or toissas of Eguld recfosceW materiet. 7he radoecWwe materials are housed, or enclosed to reduce me emocts of weatherbg and provide of any contamhants, provide materief conenin, and prevent dooey of profoceW meeswee. For aross outside me protected area, e aree wm be estabnehed to aEuste the boundary and answa the Amts of 10 CFR 20.f301 are met. The todoeceW weado eres is not a aforage aree, and the wests k appropriefoly kr sh> ment conktmhg to oor and twee m mersey preven 6ng arr/ rudesse, em redesse of redeemey above me knes of Tech spec 5.U, avut enewho the roounments of 10 CFM 20.f301 are met. Hence, em aforage of redoacew meenrief does not cronie me posemary Mr en accident of a derent type man anypreviouery evaluend h en sAR.

6. May crones the possibility for a malfuncnon of equipment important to safety of a O Yes different type than any prwiously evaluased in the S AR. ><

No BASIS: For the aforage of radoacek materief or me holdng of radoscWwe weafe, no equ> ment important to sannty wit be used or asected h any way. As a resut of aforkp ramecfhe maferief orholdng radoscWwe wealt no equpment hiportant to sehty k used or eSocted. Accort8ngfy, mere k no crea60n of a posaDEy of a meNuncnon of equQment important to sannty of a GSerent type than anyo- 2ry evaluated h SAM.

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j Page Lof 6

7. Will redmos the margia af safety as danned in the BASIS: for any Technical O Yes W 3 No

BASIS

Tod Spec &&4 detags the enredons amt requirements of racsoecWw Wauents andmahanhhe done m #w ALARA. The shmtradoecWe maferW i was decontamhama betro conensed acons aree e remow a much

!' acthey as posande andis housed, or encioned to pm4de contahment of anycontamhants. Therettre there no twdudion h the margin of saflety as denned M the BASIS str any Technicef SpecdNce6en i IIL EnviressmestalEvaluation Environmental Evaluation act required per Saduy and Eavuomeental Pre-Scnemaag or 'ty

{, Rayww.

j IMPLEMENTATION On FEaFOaMANCE OF THE AC110N mmm IN THE EVALUATED DOCUhG2rr:

f l A. Environmental Protection Plan I

i

{ 1. Will require a changs in the Envuonmental Protection Plan. D Yes

! E No j BASIS: SecWon 2.4.13.3 of the SAR addosses accidenfat redesses of Aguld radoecifw material. Liguld radoecWw materief is not arosent h sliptincent amounts M any of the

! componsnts or roms bomp how or stored. The redoadhe materiais are Asmer omfocied i kom rah and incidental water or Equide by being autat& packaged, housed, or enckeed.

{ 7his asso seduce me esisde or weamerhe and am4 des contahment or any contamhants, i

paddes medeninf conenis, andprevents prematwo decey orprofeasw mesmass.

Technical 's ohne Done raa~hnnns neanuel (oOCht) Appener A 1.0 estahnahn resnoted aren inv eauents a me see boundary. The redoecem materiaf

storage messons outside orme protoded ares are mehh me see boundary and anchoed h a j conkoned area (eg, bu0dng room, or weh accked gefes) weh access conkoned by i Heath Physics. Tne done rates at the area barclar are benow the ents Ested h 10 i CFR 20.130116r done e the puth and are Asther ensured byperienc swwy and aron TLD i Therettre, storage ofradoscWw mehrfef oWalde orthe contofed access area or l aree cues notreguhr a change e me ErwtonmentalProtecnon Plan.

B. Unreviewed Environmental Ouestion 1.. Concerns a maner which may readt la a sigmticant incmass is any advens a Yes envuomeental impact pseviously evalueesd la the Final Eavuommental Statement (FES) 5 No as modiSed by the NRC stairs testunomy to the Atoadc Safay and Liceamag Bonni (ASLB), supplements to the FES, savunamental impact appraiani, or la any decisions of the ASLB.

BASIS: Sec6on K9.1.1.2(2) of the FES states that +hnnns show the cumulathe don to me aposed populacon . tom redacon nous som me reactor and as assodated components wouw be hatpincent when compared to netwat bettgrounct done, Jeu than 1 persorwomwar. n afoo storm met bw*ver resoecew meeerief shrase contahors wir coneewe an esamerat 0.1% orme aposure som naragon fs at me see boundary. Sedian 2.a f oram ooCurequires dehrminalfon orcumuisse dose e members orme pubsc som dired radisson, andisguires mis be made by our o'wtonmentaf 7LDs which are boefod near the ane h 8 of 16 mofoerologicef soofors. 7he pdecoment of these contahors is the radmkolcaf conetunon at me see boundary is noengene and we not strinoeney hereen any redanon Jevois already M nintence, and we be humer ensured by the readhos taken kom the emtonmental TLDn

4 SE No. 96-0071-R00 Page ,j_of 6

2. Concerns a sigmscent ctangs la efnuents or power level D Yes III No BASIS: The indoacWw materials behg aforud orhatt do not confah any redoac#w materief abow teos amomts ufsch wount how a negiDNe omsct M These maiorfals are aforod adhh the afe boundary. Theredbre, there are no enluonts or power Jewel changes assoodefod wilh the aforage o(these materiefs.

3. Concerns a matter not previously revweed and evaluated la the dunaeants specded la 0 Yes III.B.1 abme, which may have a agnifkant environmental impact. (Il No BASIS: The aforage ofredoads materfat wPhh the age boundary is addressed b FES Sec00n 5.9.1.1.2(2), and does not concem a matter not previouafy evefunded h Ill.B.1 ofihls fbrm.

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SE 96 00045.R00

%'? In !'il W L pegg 1 Q cn g l

• 4 GRAND GULF NUCLEAR STATION UNIT 1 CHANGES, TEST 8 OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM L

1. Safety Evaluation Overview A. Reference Data OR@NATOR: Bryan Ford dept / sect: NS&RA EVAL.#: 96-00065-R00 .

DOCUMENT EVALUATED: TRM/UFSAR Appendix 168 Change - Reduce survosilence frequencies for refuel platform, saxiliary platform and fuel handiing platform for LCOs 6.9.3 & 6.9.5.

SYSTEM NO: F11 REFERENCEst GGNS - Techmcal SW1.c Ms updated throu9h amendment #122, UFSAR Rev. 9, Ch. 7, 9,15 NUREG-0612, NEDO 31466.

FSAR CHANGE REQUIRED? 3 Yes O No CR# 96-111 FSAR SECDONS TO BE REVISED: CHAPTER 16, APPENDIX 168 TRM CHANGE REQUIRED a Yes a No TECH. SPEC. CHANGE REQUIRED 0 Yes a No CR # N/A IS THE VAUDITY OF THis SAFETY EvAtuAnON DEPENDENT ON ANY CHANGES OTHEROTHAN Yes THE CHANGE BEING EVALUATED (E.G. PROCEDURAL. OPERADONAL CONOmONS)?

g ExPLArc N/A l

IF vus TO THE LAsT OuEsTION, HAVE 1NE ORGANIZATION 8 RESPONSELE FOR THOSE CHANGES O Yes BEEN NOTIFIED?

(THE RESPON86BLE ORGANIZAT)ONS MUST BE NOTIFIED PRIOR TO EdPLEMENTING TH18 CHANG Signatures and Approvala of Attached Safety and Environmental Evaluation Evaluated: g QJ & g/gk P.HK NAME) om isNA W (SIGNATURE) 'oATE Reviewed: 3,% peg hh gh/g (PRINT 71AME) INDEPENDENT REvienkR (SIGNATURE) DATE Reviewed: ph (PRINT NAME)~ OTHER REV@NER (IF REQUIRED) (SIGNATURE) DATE Plant Safety Review Committee Review

%A-- 4, to/3k(o CHAIRMAN SRC DATE J

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' SE 9s 00045-R00 pose 2

,u s B. Executive Summary Brief Description of Change, Test or Experiment:

The surveillance test frequen,:y of interlock functions is being reduced to eliminate unnecessary testing. These interlocks have a reliable operating history. The proposed change will continue to provide the intended equipment protection currently provided by the interlocks. The proposed change in surveillance frequencies is listed below:

Refueling Platform:

For TRM/UFSAR Appendix 168 surveillences SR 6.9.3.3 through 6.9.3.5, the frequency will be changed from "within 7 days prior to handling fuel assemblies or control rods' to '31 days".

Fuel Handling Platform:

l For TRM/UFSAR Appendix 168 surveillance SR 6.9.5.1 the frequency will be changed from "Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of starting hoist operation AND 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />' to "12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />".

For TRM/UFSAR Appendix 168 surveillances SR 6.9.5.4 through 6.9.5.6 the frequency will be changed from "within 7 days prior to handling fuel assemblies or control rods" to "31 days."

There is no documented technical basis for the current surveillance frequencies of these interlocks.

Reason for Change, Test or Experiment Reduce surveillance test frequency of the refueling interlocks listed above to eliminate the unnecessary testing.

Safety Evaluation Summary and Conclusion The interlocks affected by this change primarily protect the refueling equipment, reactor vessel intomals, fuel storage racks or fuel assemblies during refueling operations. Some interlock functions apply to both the Refueling Platform (TRM/UFSAR Appendix 168 6.9.3) and the Fuel i Handling Platform (TRM/UFSAR Appendix 168 6.9.5). The interlock functions with a brief description and the cssociated Surveillance Requirement (SR) references are listed below:

l Jam Cutofflnterfock Protects reactor vessel intomals and fuel storage racks from excessive

! lifting force should they become inadvertently engaged during lifting operation and fuel l assemblies from excessive force should they become stuck. (SR 6.9.3.3 and SR 6.9.5.4) t l Primary & Redr/ndant OverfoedInterfock: Protects the hoists and limits the loads carried by i the hoists by assuring that loads in excess of their of the expected ranges in weight are not i moved by the hoists. (SR 6.9.3.4, SR 6.9.5.5, and SR 6.9.5.6) 1 A

. - . y , - . . - - y

~ *

- se es.oooss.noo

. pagesg}T 9D Downtravel Cutoff /nterfock: Protects the main hoist and the reactor intomals by assuring that the hoist does not reach the top guide and that the hoist cable ends are not pulled loose from the drum. (SR 6.g.3.5)

Monorail Auxillary Holst load Override: Protects the monorail auxiliary hoist by preventing l lifting of loads above its design capacity. (SR 6.g.5.1)

Accident Analyses The Fuel Handling Accidents (FHA) in the containment and a'uxiliary building were evakstad for the efect of proposed changes in the surveillance frequencies of refueling equipment. The l failurs of any of these interlock functions is not an event initiator or mitigator in the postulated I FHAs discussed in UFSAR Sections 15.7.4 and 15.7.6 or in any of the reactivity events discussed in UFSAR Section 15.4.1.1.

These interlocks, except for the Jam Cutoff interlock, are credited in the analyses presented in UFSAR Appendix 90, "GGNS Compliance with NUREG-0612, Control of Heavy Loads at Nuclear Power Plants. This analyses does credit the existence of these interlocks to prevent these hoists from being of concem for drops of heavy loads. But this analyses does not credit  ;

any specific surveillance intervals. The Jam Cutoff Interlock functions to protect the reactor intamals and fuel storage racks from damage and prevents separating the handle from a stuck fuel bundle. The change only reduces the surveillance frequency of the refueling equipment which is shown to operate reliatdy. Reliability of these interlocks has been established through their operating history over the last three refueling outages during which no surveillance test failures were experienced. Engineering judgment was used in conjunction with this operating i history of the interlocks to determine the proposed surveillance test frequencies. Due to the l demonstrated high reliability of the subject interlocks no discemible increase in the frequency of the failure of this equipment is expected at the revised intervals.

The proposed changes only reduces the surveillance frequencies of these interiocks. The  !

interlocks will continue to provide the intended equipment protection after the change since no I change is made in the design or operation of these interlocks or the associated refueling equipment and due to the established reliability of the interlocks.

The surveWances affected by this change were removed from the Technical Specifications via amendment #120. Therefore, this change does not require a change in the Technical Specifications.

The proposed change does not result in an unreviewed safety question. '

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SE 9s 00045 R00 <

page 4 ofQ l 91?h 11 Safety Evaluation A. TechnicalSpecifications

1. Implementation or performance of the action described in the evaluated document will require a change to the GGNS Unit 1 Technical Specifications.

, Yes O No E BASIS:

The surveillance requirements being changed are contained in the TRM and UFSAR Appendix 168 and not in the Technbal Specifications. This change has no impact on any current Techrilcal Specification.

B. Unreviewed Safely Question implementation or performance of the action described in the evaluated document:

i

1. May increase the probability of occurrence of an accident previously evaluated in the SAR.

Yes O No E BASIS:

The interlock functions affected by this change protect the refueling platform, the fuel handling platform, the reactor vessel, the reactor vessel intomais, the fuel storage rocks, and the fuel assemblies from overloading or overstressing. However, failure of any of these interlock functions is not an event initiator in the postulated FHAs discussed in UFSAR Sections 15.7.4 and 15.7.6 or in any of the reactivity events discussed in Section 15.4.1.1.

i The proposed change only reduces the surveillance frequencies of these interlocks. No change is made in the design er operation of these interlocks or the associated refueling equipment.

Reliability of these interlocks has been established through their operating history over the last '

three refueling outages during which no surveillance test failures were experienced.

judgment was used in conjunction with this operating history of the interiocks to determine the proposed surveillance test frequencies. Due to the demonstrated high reliability of the subject i interiocks no discemible increase in the frequency of the failure of this equipment is expected at the revised intervals. The interlocks will, therefore, continue to provide the intended equipment i protection described in UFSAR Chapter g after the change.

Therefore, the proposed change will not increase the probability of the accident previously evaluated in the SAR.

2. May increase consequences of accidents previously evaluated in the SAR.

Yes O No E BASIS:

The refueling interlocks affected by this change and their associated instrumentation are not assumed in the mitigation of the consequences of FHA analyses discussed in UFSAR Sechons 15.7.4 and 15.7.6 or the reactivity events discussed in Section 15.4.1.1. The proposed change, thereform, does not increase the consequences of the accident previously evaluated in the SAR.

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• SE 96 00065 R00 l

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' 3. May increase the probability of occurrence of malfunction of equipment important to safety previously evaluated in the SAR t'1s O No E BASIS:

The interlock functions affected by this change protect the refueling platform, the fuel l handling platform, the reactor vessel, the reactor vessel intamals, the upper containment

' fuel racks, and the fuel assemblies from overloading or overstressing. However, failure of any of these interlock functions is not an event initiator in the postulated FHAs discussed in UFSAR Sections 15.7.4 and 15.7.6 or in any of the reactivity events discussed in i Section 15.4.1.1.

j These interlocks, except for the Jam Cutoff Interlock, are credited in the analyses i presented in UFSAR Appendix 90, "GGNS Compliance with NUREG-0612, Control of j Heavy Loads at Nuclear Power Plants. This analyses does credit the existence of these interlocks to prevent these hoists from being of concem for drops of heavy loads. But this

analyses does not credit any specific surveillance intervals. The Jam Cutoff Interlock functions to protect the reactor intamals and upper containment fuel racks from damage

and prevents separating the handle from a stuck fuel bundle.

i The proposed change only reduces the surveillance frequencies of these interlocks. No l change is made in the design or operation of these interlocks or the associated refueling i

equipment. Reliability of these interiocks has been established through their operating j history over the last three refueling outages during which no surveillance test failures were {

experienced. Engineenng judgment was used in conjunction with this operating history of

~

the interlocks to determine the proposed surveillance test frequencies. Due to the i demonstrated high reliability of the subject interlocks no discemible increase in the

frequency of the failure of this equipment is expected at the revised intervals. The

[ interlocks will, therefore, continue to provfde the intended equipment protection discussed of UFSAR Chapter 9 after the change.

! The proposed change, therefore, does not increase the probability of malfunction of the j equipment important to safety.

4. May increase the consequences of malfunction of equipment important to safety previously evaluated in the SAR. Yes O No E BASIS:

The proposed change only reduces the surveillance frequencies of these interiocks. No change is made in the design or operation of these interlocks or the associated refueling equipment. Reliability of these interlocks has been established through their operating history over the last three refueling outages during which no surveillance test failures were experienced. Engineering judgment was used in conjunction with this operating history of the interlocks to determine the proposed surveillance test frequencies. Due to the demonstrated high reliability of the subject interiocks no discemible increase in the frequency of the failure of this equipment is expected at the revised intervals. The proposed change, therefore, does not increase the consequences of a malfunction of equipment important to safety.

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j, 5. May create the possibility of an accident of different type than any previously evaluated in the SAR. Yes O No E i,

BASIS:

4 1

The proposed change only reduces the surveillance frequencies of these interlocks. No

change is made in the design or operation of these interlocks or the associated refueling i

{ equipment. Reliability of these intadocks has been established through their operating' I

' history over the last three refueling outages during which no surveillance test failures were i experienced. Engineering judgment was used in conjunction with this operating history of

. the interlocks to determine the proposed surveillance test frequencies. Due to the 1 3

~ demonstrated high reliability of the subject intadocks no discemible increase in the l 1

frequency of the failure of this equipment le expected at the revised intervals. Therefore, no new equipment failure modes are introduced as a result of this change. Consequently, the change would not create accident of a different type than any previously evaluated in the SAR.

4 j 6. May create the possibility of a malfunction of equipment important to safety of a different l type than any previously evaluated in the SAR. Yes O No 5 i

j BASIS:

! The proposed change only reduces the surveillance frequencies of these intedocks. No i

change is made in the design or operation of these interlocks or the associated refueling

} equipment. Reliability of these interlocks has been established through their operating j history over the last three refueling outages during which no surveillance test failures were i experienced. Engineering judgment was used in conjunction with this operating history of 1

the interlocks to determine the proposed surveillence test frequencies. Due to the j demonstrated high reliability of the subject interlocks no discemible increase in the j

frequency of the failure of this equipment is expected at the revised intervals. Therefore, i the proposed change does not create possibility of a malfunction of equipment important j to safety of different type than any previously evaluated.

{ 7. Will reduce margin of safety as defined in the Basis for any technical specification.

Yes O No E I BASIS:

j The interlocks on which the surveillance intervals are being extended are not required by the Technical Specifications. Reliability of these interlocks has been established through their operating history over the last thnee refueling outages during which no surveillance test failures were experienced. Engineering judgment was used in conjunction with this operating history of the interlocks to determine the proposed surveillance test frequencies. Due to the demonstrated high reliability of the subject interlocks no discemible increase in the frequency of the failure of this equipment is expected at the revised intervals and as a result the proposed changes are not expected to affect the bases for any Technical Specification limit or surveillance requirement. The proposed change, therefore, does not reduce the margin of safety as defined in the basis for any technical specification.

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SE 96 00005 R00 b

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- ~bh wb Ill. Environmental Evaluation O Not applicable per Environmental Evaluation jl Applicability Review

. IMPLEMENTATION OR PERFORMANCE OF THE ACTION DESCRIBED IN THE EVALUATED DOCUMENT:

l A. Environmental Protection Plan 3 1. Will require a change in the Environmental Protection Plan. O Yes E No

Basis

i 4

B. Unreviewed Environmental Question

! 1. Concems a matter which may result in a significant increase in any adverse i

O Yes environmental impact previously evaluated in the Final Environmental 5 No j Statement (FES) as modified by the NRC staffs testimony to the Atomic Safety and Licenslng Board (ASLB), supplements to the FES, environmental impact appraisal, or in any decisions of the ASLB.

l

. BASIS: l l

2. Concems a significant change in emuents or power level.. O Yes E No ,

BASIS: l

3. Concems a matter not previously reviewed and evaluated in the documents O Yes specded in ll.B.1 above, which may have a significant environmental E No impact.

Basis: I 1

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Wod- PSE ADMINISTPATIVE PROCEDURE 01-S-06-24 Revision: 102

' - gg,  :

i Attachment V Page 1 of 5 kI hi/4 -]

' I GRAND GULF NUCLEAR STATION UNIT 1 l CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM l L Safety Evaluation Overview Page]_ofd A. Reference Data ORIGCJATOR: RON GREEN JErr/ SECT: PASE EVAL. #: 96-0087-R00 DOCUMENT EVALUATED: WO 174234 SYSTEM NO' B33 REACTOR RTCIRCULADON SYSTEM REFERENCES 1

TECHNICAL SPECIFICATION 3.4.2 ft.OW CONTROL VALVES i SAFETY EVALUADON 86-095 FSAR SECDON 15.3.2 RECIRCULATIONFLOW CONTROL FAR.URE - DECREAsm FLOWRATE FSAR SECDON 15.4,5 REC" :'4 4 DON FLOW CONTROL FAILURE nf7HINCREAsM FLOW FSAR 15.4.5.3.2.2 SLOW OPENING OF A R&lRCULADONFLOW CONTROL VALVE l

FSAR 5.4.1.1 SAFETYDEsIGNBAsis(REACTORRECIRCULADONSYsTEM) i FSAR 5.4.1.9 FLOW CONTROL SYSTEM DEsCRIPDON l

VENDOR MAMAL 460000787 REACTOR RECIRCULA DON HYORAUuc ANO AssOctA TED ELECTRON AND ELECTRICAL EQUIPMENT l

FSAR CHANGE REQUIRED 7 a Yes a No CR# CR or(n/a)

FSAR SECTIONS TO BE RIVISED:

• O Yes TRM CHANGE REQUIRED 7 a No l TECH. SPEC. CHANGE REQUIRED 7 m O Yes a No CR# CR or(n/a)

IS THE VAUDCTY OF THIS SAFETY E7ALUATioN DEPENDENT ON ANY CHANGES OMER O THAN Yes THE CHANGE BEING EVALUATED (E.O. PROCEDURAL, OPERATIONAL CONDITIONS)?

gg l

EXPLAIN: SYSTEM OPERATION PER Sol IS NOT CHANGED BY THIS EVALUATION IF YES TO THE LAST l BEEN NO11FILD7 " QUESTION, l. AVE THE ORGANIZATIONS RESPONSIBLE FOR THOSE CHANGES D Y (THE RESPONSIBLE ORGANIZATIONS MUST BE NOTIFIED PRIOR TO IMPLEMENTING THIS CHANGE.)

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GRAND GULF NUCLEAR STATION ADMINISTRATIVE PROCEDURE 01-S-06-24 Revision: 102 Attachment V Page 2 of 5 RD&5 Signatures and Approvals of Attached Safety and Environmental Evaluation Evaluated: "5 ( hgg ,

f p 22-f4 (PRINT NAME) ORIGINATORL (SIONATURE) DATE Reviewed: 08 l D. A. G AnWAw (PRINT NAME)

/h[ / 9-4-%

INDEPENDENT REVIEWER (SIONXTURE) DATE

! Reviewed: 09

/ /

(PRINT NAME) OTHER REVIEWER (if REQUIRED) (SIONATURE) DALE l

l Plant Safety Review Committee Review l 05) I ODOyJa i5 nv L ci, Tu m T&lKtett LG 7 9!1klf 5

c PS C / '

DATE M I

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. . *0 RAND GULF BlUCLEAR STATION ADMINISTRATIVE PROCEDURE

, 01 S 06-24 Revision: 102 i

Attachment V Page 3 of 5 SE No. 96-0087-R00 Page 2 of X B. Executive Summary (A130 SERVES AS INPUT TO NRC

SUMMARY

REPORT)

BRIEF DESCRIPTION OF CHANCE, 'IT.ST OR EXPERIMENT: WO 174234 DOCUMENTS TMOUs:.ESH00TNG EPPoMTS FOM RECIRCULA MON FLOW CONTMOL VALVE *B" (1833F069. TM MSULTS OF TM0veL SHOWN THAT THEPOSMONFEEDEACK TMANSDUCEM, TMRyuTISNOTMESPohDNG ASMOUtMEO. TM RESPONSE LAST EVENNG (9-21-96) HAS BEEN NOISY AND EMMAnC AT DMES, TNUS THE RVDTIS CONScEMED To BE OPEMATINGINA DEGMADED CON 0m0N. TwS SAFETYEVALUA n0NDOCUNNTS THEEVALUADONOF VAL OPERA TION WITH A DEGRADED OR POTENTIALLY DEGMADED POSMON FEEDSACK SIGNAL FROM THE FCV REASON FOR CHANCE, TEST OR EXPERIMENT: THE REASON FOR TMS EVALUADONIS TO PMOVICE DOCUMENTADON OF THE MEVIEW OF PEM100CALLY UNLOCKMG THE HPU TO MEPOSMON THE VALVE AS NECESSARY TO RECIMCULADONFLOW. THr5 EVALUADONISNECESSARYBECAUSE THEFCVB MUSTat MEPOSm0NEO PERICOCALLY OUE TO ANINTEMNAL LOOP HYDMAUUC LEAK CAUSNG THE FCV B To SLOM.Y DNFT CLO ALc0 TO ACCOMPUSH FCV OPEMNG OM CLOSMG FOR NOMMAL RECIMCULADONPLOWCHANGES AND NOMM PLANT SHUTDOWN. his OPEMDON DOES NOTNVOLVE A SAFETVFUNCDON, AS THE RECIMCULAn0N SYSTEMIS A POWEM GENEMATION SYSTEM ANOIS NOTREQUtMED FOR SAFETY, NOR MEQUfMED TO OPERATEDURNG OM AFTEM ANYDESIGNBASIS ACCCENT, FUMTHEM, THE CONTMOLS AND NTEMLOCKS AMENOTMEQUtMED OM DEDGhnD TO COMPLY WITH ANY SNGLE-FAILUME CRITERIA. CONTMOL SYSTEM F LUMES MESULDNQ N COMPLETELOSS O CONTMOL SIGNAL WELL MESULTNELECTMICAUHYDMAUUC LOOGhG M TMFNAL CONTMOL VALVE ACTUATOM.'

SAFETY EVALUATION

SUMMARY

AND CONCLUSIONS: THISSAMTYEVALUAn0N ADOMESSESPEMOD OF THE HPU To CHANGE THE FCV POSMON UNDER THE CON 0m0NS OF A DEGRADED POSMON OPERAn0N OF THE FCV WITH A LOCKED-UP HPU HAS aEENPMEvs0USLYEVALUATED UeER SE SAFETY EVALUADONIS CONSDEMED STEL APPUCABLE, SNCE THE OM.Y CNANGE SNCE THATEVALUANON WAS COMPLETEDIS THE CHANGE TO NEWITS (IMPMOVEO 7lrCHwCAL SPECn9CAn0NS) CONCEMNNG FCV OPEMABlUTY AND MATE OF MOVEMENT, WHFCH HAS NO BEARNG ON HPU LOCKUP.

THE POSITION FEEDBACK SIGNAL, WMCHIS SUPPUED BY THE RVDT, IS COMPANEO TO THE POSm0N SETPONT ,

SIGNAL BY THE POSm0N CONTMOLLEM SUmnNG JUNCn0N. THE MSULTANT POSm0N DEvnAn0N\

AMPUFIED BY THE P0sm0N CONTMOLLEM AMPUREMS TO PMOVICE A VELOCITY DEMAND SIGNAL WHCl THE ACTUATOM TO TRAVEL N THE OMECDON REQUMED TO MEDUCE THEPOSm0NDEVEAn0N TO 2EMO. THUS,  ;

VALv:POSm0N MLL RESPONO TO THE POSm0N SET PONT SIGNAL. DEGRADAn0N OF THE RVDT O SIGNAL OR LOSS OF THE OUTPUT SIGNAL DOES NOT MESULTN UNCONTROLLAnlE VALVE MOVEMENTDUM

~

THE HPU iS UNLOCKEO, BECAUSE THE POSm0N CONTMolLEM LanTER UMtTS THE MAGMTUDE OF THE VELOCITY SET POINT SIGNAL SO THAT ACTUA TOM VELOCITY WILL NOTEXCEED PMESETUMTS EVEN 9 LAMGE MAPD CHAN POSm0N SETPONT OCCUM. IF THE PCSm0N SETPONT(DEMAND) SIGNAL BtCEEDS PMESETUMtTS, OM EXCEEDS DESIGN MATE OF CHANGE UMITS, OM 9 THE SIGNAL IS UNSTABLE, THE HPU SUELOOPS AME AUTOMAnCALLY ^

LOCKED, PMEVENTING VALVEM0 DON.

l FuttTHEM, ITIS NOT POSTULA TED THA T A FAILED OM DEGMADED SIGNAL FROM Tw RVOT WOLA.D MESULTN A MPD '

OPENNG OF THE FLOW CONTMOL VALVE, HOWEVEM, IF N THE UNUKELYEVENT THS WEM TO OCCUR DURNG VALVE POSm0N CHANGES, THE FLOW CONTMOL VALVE MA TE OF OPEMNG 15 UMtTED ELECTMCALLY AhD HYDMAUUCALLY (VtA VLLOCITY UMtT OM99CE) TO PREVENT POWER EXCURSIONS DUE TO EXCESSIVE MEACT.VITY AD0m II. Safety Evaluation a O Safety Evaluation not required per compieeed Safety and

< Erwironmental Pre-Screening or Applicabdity Review Proceed to Section III for Environmental Evaluation.

l A. Technical Soecifications

• i Implementation or performance of the action described in the evaluated docenent will D Yes

1.

require a change to the GGNS Unit i Technical Specifications. 5 NO '

t BASIS: TECH SPEC 3.4.2 ADOMESSES OPEMAhUTY OFFLOWCONTMOL VALVES. The TtCHSPEC l BASES IS CONCERNED MTH OPERAnlUTYIN MEGAMD TO VALVE STM0KE MTE. THE POSm0N

]

FEEDBACK SIGNAL, HNETHEM NOMMAL OM DEGRADED, DOES NOT AFFECTM TE OF VALVE STM0KE, THEMEFOMEPERIOOC OPEMAn0N OF A FCV MTH A POTENnAU.YDEGMDEO RVOT DOES NOT CHANGE OM AFFECT GGNS UMT 1 TECHMCAL SPEC 99 CATIONS.

i J:\ADM_SRVS\ TECH _ PUB \ REVISION \1\1SOE24.A5

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f GRAND GULF NUCLEAR STATION ADMINISTRATIVE PROCEDURE

, 01-S-05-24 Revision: 102 Attachment V Page 4 of 5 B. Unreviewed Safety Ouestion y 4 OS $

IMPLEMENTATION OR PERFORMANCE OF THE ACTION DESCRIBED IN THE EVALUATED DOCUMDrr:

1.

May increase the probability of occurrence of an accident previously evaluated in the Q Yes SAR.

5 No ,

) B ASIS: OPERAn0N OF THE FCV MTH A PO TENDAU.Y DEGMADEO RVDT 00ES NOTNCMEASE TH l

ANY ACCIDENTEVALUATED IN THE SAR. Tkt EXPECTED ACDON ONDEGMADADON OF THEFEEDSACK SIGN TRIPPNG OF THE ASSOCIATED HPU OUE TO EXCESS SERVO EMMOM ($% OkTEMENCERETHEEN PoSm0NDEM AND ACTUAL POSinON). EVEN MTH A COMPLETE FAM.UME OF DM ELECTM0heCS, Tw FCVIS DESIGNED TO UMrT STM0KNG MATE TO < 30 % PEM SECONOIN THE OPEMNG OMECDON AND < 60% N THE CLOSNG OMECDON.

l l 2. May increase the consequences of an accident previously evaluated in the SAR.

O Yes i

5 NO BASIS: THE ONLY POSSIBLE ACCIDENT THA T COULD BE CONSCEMD MMALD BE MEACTMTY ACC RAPIO OPENNG OF A FCV. SNCE FAILUME OM DEGMADEO OPERA n0N OF THE RVOT 00ES NOT AFFECT TH .

l RA TE OF OPENING OM CLOSING OF THE FCV, NO CONSEQUENCES OF ANY ACCJ0ENT EVALUA TED N THE SAR AME  !

INCMEASED BY PERIOUC OPERA TION OF THE FCV MTH A POTENnALLY DEGMADED RVDT.

3.

May increase the probability of occurrence of a analfunction of equipment important to O Yes safety previously evaluated in the SAR.

\

5 No

\ BASIS: THE RECIRCUU DON SYSTEM PEMFORMS NO SAFETY FUNCDON AND NO EQUFMENT ne IS AFFECTED BY UTUZNG A POTENDALLYDEGMADEO RVOT. T>nt MEuAavrYOF THERECaCULAn0NSYSTEM TO PERFORM ITS FUNCTION OF MANTAINNG MEACTOR COOLANT PMESSUME BOURDARY NTEGMF1YIS NOT L

ii 4. May increase the consequences of a malfunction ofequipment unportant to safety 0 Yes l previously evaluated in the SAR.

3 No i BASIS: SNCE NO EQulPMENTIMPORTANT TO SAFETYIS AFFECTED, THEM AME NO CONSEQUENCES l OF ANY MALFUNCTION TO INCREASE. \

5.

May create the possibility for an accident of a diff'erent type than any previously 0 Yes evaluated in the SAR.

B No BASIS: THE SAR EVALUATES BOTH SLOW AND FAST OPENNG AND CLOSNG OF ONE OR BOTH FCVS.

NO OTHEM ACCIDENT OTHER THAN VALVE MOVEMENT DUE TO AN NCOMMECT F6EDBACK SIGNAL IS POSTULATED As POSSIBLE.

6. May create the possibility for a malfunction of equipment important to safety of a O Yes different type than any previously evaluated in the SAR. 3 No BASIS: THE RECIMCULAMON SYSTEMS ONLYFUNCn0N MELATED TO SAFETYIS TO MANTAN THE REACTOM PMESSURE BOUNOARY. NO EQUIPMENTIMPORTANT TO SAFETYIS AFFECTED, THEME FORE NO MALFUNCTION OF ANY O THER EQUIPMENT IS POSTULATED.

i f

7.

Will reduce the margin of safety as defined in the BASIS: for any Ta4aie=1 O Yes Specification.

3 No BASIS: No margin of sal >ty is atFected by operation of the FCV wth a potentiecy degraded RVOT. The valve stroke rate is not einected by the RVOT.

I

3:\ADu_Savs\ TECH _ PUB \ REVISION \1\1S0624.A5

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Attachment V Page 5 of 5 1 l

QQ 3 e 5 of 5 l l  :

l IH. Environmental Evaluation" Environ =*atal Evaluation not requued per

( Safety and Environmental Pre-Screemag or 'ty Review. i IMPLEMENTATION OR PERFORMANCE OF THE AC110N DESCRIBED IN THE EVALUA17,D DOCUMENT

• A. Environmental Protection Plan "

1. Will require a change in the Environmental Protection Plan. D Yes 1 E No I BASIS: THE RECIRCULATION SYSTEM IS NOT ADDRESSED C4 THE EPP.

t B. Unreviewed Environmental Ouestion "

1. . Concerns a matter which may result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) a Yes g No as modified by the NRC staffs testunony to the Atomic Safety and Licensing Board (ASLB), supplements to the FES, environ =eatal impact appraisal, or in any decmons of the ASLB.

BASIS: THE RECIRCULATION SYSTEMISIN THEORYwELL. INTEGMfTY OF MREACTOM VESSELIS NOTIMPACTED AS ADOMESSEDIN THE ABOVE SAFETYEVALUATION, TMMSFOME TMMEISNOIMPACT

\ ON THE ENVIRONMENT.

i l

l

2. Concerns a significant change in effluents or power level. O Yes BASIS:

5 No OPERA TION WITH A POTENTIALLY CEGRADED RVDT HAS NO IMPACT ON EFFLUENTS OM FO%ER.

3. Concerns a matter not previously reviewed and evaluated in the documents WW in O Yes 111.B.1 above, which may have a significant environmental i=F B No BASIS: NO IMPACT TO THE ENVIRONMENT.

i i

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GRAND GULF NUCLEAR STATION r cuena INISTRATIp PROCEDURE

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• Miu. 5 -S g

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, a.m!a :t FAG 3 S l q Ccf to IniqL Attachment V Page 1 of 9 N a N " 1l} %  %

1 l

GRANDGULFNUCLEARSTATION UNIT 1 i CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM i I. Safety Evaluation Overview Page 1 of 9 A. Reference Data ORIGINATOR: ALAN J. MALONE DEPT / SECT: P&SE Ev4Ls: 96-0098-R00 l DOCUMENT EVALUATED: LICENSING DOCUMENT CHANGE REQUEST NO.96-123

(

SYSTEM NO(S): N/A ( twsERT N/A IF NOT APPUCASLE)

REFERENCES:

SEE ATTACHED LICENSING DOCUMENT CHANGE REQUEST l FSAR CHANGE REQUIRED 7 0 Yes E No ca # N/A FSAR SECTIONS TO BE REVISED: N/A l TRM CHANGE REQUIRED 7 m Yes a NO TECH. SPEC. CHANGE REQUIRED 7 0 Yes a No ca # N/A l IS THE VAUDITY OF DBS SAFETY EVALUATION DEPENDENT ON ANY CHANGES ODIER THAN O Yes THE CHANGE BEING EVALUATED (E.G. PROCEDURAL, OPERATIONAL CONDITIONS)?

E No l

EXPLAIN. .N/A i

r IF YES TO THE LAST QUESTION, HAVE THE ORGANIZATIONS RESPONSIBLE FOR THOSE O Yes CHANGES BEEN NOTIFIED 7

! (THE RESPONSIBLE ORGANIZATIONS MUST BE NOTIFIED PRIOR TO IMPLEMENTING THIS CHANGE.) i l

i Signatures and Approvals of Attached Safety and Environmental Evaluation Evaluated: &y JQy , g, ,f0.y.yf (PRINT NAME) ORIGINATOR (Si ATURE) DATE I

Reviewed: ))j,._ p _ $ ,., , gg  ; yo,y,.yg (PIUNT NAME) INDEPENDENT REVIEWER (SIGNATURE) DATE i jo-17 %

Reviewed: g/A (PRINT NAME)

, M ,

DATE OTHER REVIEWER (IF REQUIRED) iSIGNATURE)

Plant Safety Review Comnpittee Review i

tY // 9S CHAIRMAN, PSRC ' 'DATE i

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,, , GRAND GULF NUCI, EAR STATION ADMINISTRATIVE PROCEDURE 01-S-06-24 Revision: 102 Attachment V. Page 2 of 9 SE No. 96-0098-R00 Page 2 of 9 B. Executive Summary (ALSO SERVES AS INPUT TO NRC

SUMMARY

REPORT)

BRIEF DESCRIPTION OF CHANGE, TEST OR EXPERIMENT

• This change revises Technical Requirements Manual (TRM) Surveillance Requirement (SR) 7.6.3.3.g.2 to allow the Standby Liquid Control pump relief valves to be tested at least once per 18 months during plant or system shutdowns to verify that they open within the specified 3% tolerance Currently, TRM SR 7.6.3.3.g.2 requires that the relief valves must be tested "[ alt least once per 18 months, during shutdown,"

without clarifying whether the " shutdown" is plant or system shutdown Although the TRM requirement could be interpreted as currently allowing the testing during system shutdown, we have conservatively chosen to address this change as an intent change, rather than an editorial, or non-intent, change.

This change is a specific application of a general clanficaten of the required intervals specified in ASME Code,Section XI, as stated in NUREG-1482. According to NUREG-1482, testing or examination of any valve in the IST Program can be deferred to a refuelmg outage interval, without rel sf, using the guidance in ANSI /ASME OM Standard, Part 10. Additionally, it is allowable to test or examme those valves that are currently tested / examined on a refueling outage frequency during system outages, provided that the testing or examination is done at the same intervals as the refueling outage intervals.

REASON FOR CHANGE, TEST OR EXPERIMENT Performsnce of valve set pressure testmg, exeressing testmg and disassembly of valves for inspecten of intomals is frequently on the entical path schedule during refueleg outages and contributes to lengthened outages, thereby requiring purchase of more expensive replacement power, as well as requiring additional personnel resources. Performance of this testing and inspectens during system outages while the plant is 4 operating at power contributes to shorter refueling outages and better use of plant staff personnel, resulting l in lower outage costs. Although it is antic; pated that these tests and examina6ons will be performed only  ;

when a system is being taken out of service Ibr other reasons, the potential exists that the schedule for i performing a test or examination may require that the system be taken out of service solely to comply with i the required test or examinaten intervals specified in ASME Code Section XI.

SAFETY EVALUATION

SUMMARY

AND CONCLUSIONS Testing or examination of any valve in the IST Program can be deferred to a refueling outage interval,  ;

without relief, using the guidance in OM-10. Additionally, it is allowable to test those valves that are currently tested or exammed on a refueling outage frequency during system outages, provided that the testing / examination is done at the same intervals as the refueling outage intervals.

Since the systems in which these ceinfi.E,eiits are installed must be declared inoperable while they are removed from service for tests and examinations, the systems must be evaluated for the effect on the plant before they are taken out of service, as required by Technical Specht;eiis. In fact, the Tech Specs require that the plant enter a limiting condition for operaton (LCO) whenever such safety-related systems are taken out of service. Since the systems are out of service, they cannot influence the operaton of the plant, increase the probability of accidents or malfunctons of equipment, nor can they significantly affect the consequences of any accidents or malfunctons of equipment.

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', 01-S .06-24 Ravihlon: 102

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} Attachment V Page 3 of 9 i

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! II. Safety Evaluation Safety Evaluation not required per Completed Safety and l Environmental Pre-Screening or Applicability Review. Proceed to l J) Section III for Envvonmental Evaluation. )

~! I j A. Technical Specifications i 1

l. Implementation or performance of the action described in the evaluated document O Yes will require a change to the GGNS Unit i Technical Specifications. 3 No BASIS: Pump and valve inservice testing (IST) requirements are given in ASME Boiler and

. Pressure Vessel Code (ASME Code),Section XI, " Rules for Inservice inspection of Nuclear j

' Power Plant Components," which is invoked by GGNS Unit 1 Technical Specificebon (Tech 1 1 Spec) 5,5.6 and Technicc) Requiremonia Manual (TRM) Surveillance Requirement (SR)

! 7.6.3.3.

l Tech Spec 5.5.6 is a general testing requirement that simply requires that the testing and ,

j examinations specified in ASME Code Section XI shall be performed. it also specifies the

required frequencies in days for the intervals (e.g., weekly, monthly, quartetty, etc.) that are given in ASME Code Section XI. It does not get into the specdics of the individual testmg and examination requirements, nor does it address testmg components during shutdown. As discussed in "Brief Description of Change, Test or Expenment," above, and amplined in li Licensing Document Change Request No.96-123, the NRC has determined, in NUREG-1462, j that performance of cold shutdown testing during system outages meets the intent of the

ASME Code Seebon XI requirements.

1 The general requirements in TRM SR 7.6.3.3 to comply with Sechon XI were at one time j Technical SMuu,n SR 4.0.5, but they were removed from the Tech Specs and transferred l

! to the TRM in Tech Spec Amendment 120.- Also, the specific requirement to verifv that the  !

SLC

• pump relief valve opens within 3% of the system design pressure" was formen; in Tech '

Spec SR 4.1.5.d.2 but was removed and transferrM to the TRM in Tech Spec Amendment

120 i

j Although several other Tech Specs touch on the Sechon XI valve inservice testing requirements (for example, SR 3.5.1.4, SR 3.6.1.3.4, SR 3.6.1.3.6, SR 3.6.1.7.2, SR 3.6.2.3.2, SR 3.6.4.2.2, SR 3.6.5.3.3, etc., require surveillances to be performed "In accordance with the inservice Testing Program"), they do not affect the interoretabon of the meaning of shutdown as addressed in this Safety Evaluabon Also, this change does not conflict with any other current GGNS technicci spEJhik,6.

IMPLEMENTATION OR PERFORMANCE OF THE ACTION DESCPJBED IN THE EVALUATED DOCUMENT: i

1. May increase the probability of occurrence of an accident previously evaluated in the O Yes SAR. 3 No BASIS: As stated in Subarticle IWV 1100 (" Scope *) of ASME Code Section XI, the Pump and valve IST requirements are intended 'to venfy operational readiness of certain Class 1,2 and Ji\ADM_SR7S\ TECH _ PUB \ REVISION \1\1S0624.A5

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,: QRAND GULF' NUCLEAR STATION' ADMINISTRATIVE PROCEDUR3 01-S-06-24 R@ ion : 102 i

• j' Attachment V Page 4 of 9 no j SE No. 96-0098-R00 j Page 4 of 9 i'

1

1. May increase the probability of occurrence of an accident previously evaluated in the SAR(CONTINUED).

i BASIS (CONTINUED):

3 valves (and their actuating and position indicating' systems) in light-water cooled nuclear power plants, which are required to perform a specific function in shutting down a reactor to the cold shutdown condition or in mitigating the consequences of a accident." This change affects only the times at which these IST requirements are performed, it does not change the -

j frequency at which they are performed, nor does it change the types of tests and examinations i that are performed or the acceptance cnteria for the tests and examinabons.

I Since the systems in which these components are installed must be declared inoperable while they are removed from service for tests and exammations, the systems must be evaluated for -

the effect on the plant before they are taken out of service, as required by Technical Sporificebons. In fact, the Tech Specs require that the plant enter a limstmg condition for operation (LCO) whenever such safety-related systems are taken out of service. Since the systems are out of service, they cannot influence the operation of the plant or increase the probabihty of occunence of any accident.

2. May increase the consequences of an accident previously evaluated in the SAR. O Yes BASIS: IB No As stated in Subarticle IWV-1100 (" Scope") of ASME Code Section XI, the Pump and valve .

IST requirements are intended "to verify opershonal readiness of certam Class 1,2 and 3 valves (and their actuating and position indicabng systems) in light-wster cooled nuclear power plants, which are required to perform a specific function in shutting down a reactor to the cold shutdown condition or in mitigatmg the consequences of a accident.* This change affects only the times at which these IST requirements are performed. It does not change the -

frequency at which they are performed, nor does it change the types of tests and examinations that are performed or the acceptance entena for the tests and examinsbons.

Since the systems in which these ce,npenents are installed must be declared inoperable while they are removed from service for tests and exammations, the systems must be evaluated for the effect on the plant before they are taken out of service, as required by Technical Specificabons, in fact, the Tech Specs require that the piant enter a limiting condition for operation (LCO) whenever such safety-related systems are taken out of service. The LCO conditions usually limit the length of time that the plant may be operated with the systems out of service, and they may also impose additional requirements, such as increased venficebon of the availability of other mitigation syvems. The NRC has evaluated and approved the operation of the plant with the systems out of service under these LCO conditions, thereby indicating (whether stated or not) that the consequences of any accidents and malfuncbons of equipment that may occur while the systems are out of service are within the capabilities of the remaining operable systems to mitigate All systems and parts of systems which are takers out of service under LCOs have been evaluated for their absence during an accident. As long as the terms of the LCO, including time limits and other mquirements, have been complied with, the consequences to the plant have been found to be acceptably low.

J:\ADM_,SRVS\ TECH _ PUB \ REVISION \1\150624.A5

4 QRAND GULF NUCLEAR STATION ADMINISTRATIVE PROCEDURE 3: 01-S-06-24 Revision: 102

$ Attachment V Page 5 of 9 4 .

1

] SE No. 96-0098-R00 l Page 5 of 9 I

l 2. May increese the consequences of an accident previously evaluated in the SAR i (CONTINUED).

l BASIS (CONTINUED):

Although the unavailability of the system may increase the consequences of an accident by its

! not being capabie of performing its function during and following the accident, the NRC's j approval to operate the plant under LCO with the system out of srvice indicates that the increase is not "significant' within the meaning of 10CFR50.5g. T> .refore, this change does

, not increase the consequences of any accidents i

4 l

3.

, May increase the probability of occurrence of a malfunction of equipment important O Yes j to safety previously evaluated in the SAR. 3 No j BASIS: As stated in Suberticle IWV-1100 (" Scope') of ASME Code Section XI, the Pump and j valve IST requirements are intended "to verify operabonal readiness of ce,tain Class 1, 2 and

• j 3 valves (and their actuatmg and position indicating systems) in light-water cooled nucieer j power plants, which are required to perform a specific function in shutting down a reactor to

! the cold shutdown condition or in mitigabng the consequences of a accident." This change affects only the times at which these IST requirements are performed it does not change the i frequency at which they are performed, nor uoos it change the types of tests and exarrunabons i that are performed or the acceptance criteria for the tests and examinabons.

1 Since the systems in which these components are installed must be declared inoperable while i

they are removed from service for tests and exammations, the systems must be evaluated for the effect on the plant before they are taken out of service, as required by Technical i

Spec.ibins. In fact, the Tech Specs require that the plant enter a limiting condition for i operation (LCO) whenever such safety-related systems are taken out of service Since the

systems are out of service, they cannot influence the opersbon of the plant or increase the i probability of occurrence of any equipment malfunctions.

i

)

3 j 4. May increase the consequences of a malfunction of equipment important to safety 0 Yes i previously evaluated in the SAR. 3 No l BASIS: As stated in Suberticle IWV-1100 (" Scope') of ASME Code Sechon XI, the Pump and

) valve IST requirements are intended "to verify cpersuonal readiness of certain Class 1, 2 and

3 valves (and their actuating and posMa indicating systems) in light water cooled nucioar power plants, which are required to perform a specific function in shutting down a reactor to the cold shutdown condition or in mitigating the consequences of a accident." This change  !

affects only the times at which these IST requirements are performed it does not change the l frequency at which they are performed, nor does it change the types of tests and examinations i that are performed or the acceptance enteria for the tests and examinations. I Since the systems in which these components are installed must be declared inoperable while they are removed from service for tests and examinsbons, the systems rnust be evaluated for the effect on the plant before they are taken out of service, as required by Technical

.J:\ADM_SRVS\ TECH,, PUB \ REVISION \1\1S0624.A5

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Attachmenc V Page 6 of 9 l

1.

SE No. 96-0098-R00 Page 6 of 9 i

j 4. May increase the consequences of a malfunction of equipment important to safety j

0 Yes previously evaluated in the SAR (CONTINUED). 3 No j- BASIS (CONTINUED):

Speedicabons in fact, the Tech Specs require that the plant enter a limiting condition for j operation (LCO) whenever such safety-related systems are taken out of sonnce The LCO l j

conditions usasily firnit the length of time that the plant may be operated with the systems out l

of service, and they may also impose additional requirements, such as increased venfication 4 ]

of the availatality of other mitigation systems The NRC has evaluated and approved the '

operation of the plant with the systems out of service under these LCO conditions, thereby indicating (whether stated or not) that the consequences of any accidents and malfunctions of equipment that may occur while the systems are out of service are within the capatzlities of the j remaming operable systems to mitigate r

j All systems and parts of systems which are taken out of service under LCOs have been evaluated for their absence during malfunctions of equipment. As long as the terms of the LCO, including time limits and other requirements, have been complied with, the

{ consequences to the plant have been found to be acceptably low.

1

Although the unavailability of the system may increase the consequences of an accident by its not being capable of performmg its function dunng and following the accident, the NRC's l

approval to operate the plant under LCO with the system out of service indicates that the increase is not "significant" within the meaning of 10CFR50.59. Therefore, this change does not increase the consequences of malfunction of equipment previously evaluated in the SAR.

1 i

l 5. May create the possibility for an accident of a different type than any previously 0 Yes j evaluated in the SAR. 3 No BASIS: As stated in Subarticle IWV-1100 (" Scope") of ASME Code Section XI, the Pump and valve IST requirements are intended "to venfy operational readiness of certain Class 1, 2 and 3 valves (and their actuating and position indicating systems) in light-water cooled nuclear power plants, which are required to perform a specific function in shutting down a reactor to the cold shutdown condition or in mitigating the consequences of a accident." This change affects only the tir es at which these IST requirements are performed it does not change the frequency at which they are performed, nor does it change the types of tests and examinations that are performed or the acceptance enteria for the tests and exammations.

Since the systems in which these components are installed must be declared inoperable while they are removed from service for tests and examinations, the systems must be evaluated for

. the effect on the plant before they are taken out of service, as required by Technical Specifications. In fact, the Tech Specs require that the plant enter a limrtmg condition for operation (LCO) whenever such safety-related systems are taken out of service Since the systems are out of service and are not relied upon for a safety function, they cannot influence the operation of the plant or create the possibility for an accident of a different type than any previously evaluated.

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._QRAND GULF NUCLEAR STATION ADMINISTRATIVE PROCEDURE 01-S-06-24 Rsvision: 102 e .

Attachment V. Page 7 of 9

, SE No. 96-0098-R00 Page 7 of 9

6. May create the possibility for a malfunction of equipment important to safety of a O Yes different type than any previously evaluated in the SAR. 3 No BASIS: As stated in Suberncle IWV-1100 (" Scope") of ASME Code Section XI, the Pump and valve IST requirements are intended "to venfy opersoonal readiness of certam Class 1, 2 and 3 valves (and their actuatmg and posson indicabng systems) in light-water cooled nuclear power plants, which are required to perform a specafic funchon in shutting down a reactor to the cold shutdown condition or in megebng the consequences of a acciatant." This change affects only the times at which ther,.e IST requirements are performed. It does not change the frequency at which they are performed, nor does it change the types of tests and exammations that are performed or the acceptance criteria for the tests and exammations Since the systems in which these components are installed must be declared inoperable while they are removed from service for tests and exammations, the systems must be evaluated before they are taken out of service, as required by Technical Specificabons. Since the systems are out of service and are not relied upon to perform a safety function, they cannot influence the opersbon of the plant nor create the possitulity for a malfunction of equipment different from any previously evaluated

7. Will reduce the margin of safety as defined in the BASIS: for any Technical O Yes Specification.. 3 No BASIS: The Pump and valve IST requirements given in ASME Boeler and Pressure Vessel Code Section XI are intended to provide assurance that pumps and valves which perform a specific funcbon in shuteng down the reactor or mitigating the conesquences of a accident are capable of performing their functions when called upon to do so. This change affects only the times at which these IST requirements are performed it does not change the frequency at which they are performed, nor does it change the types of tests and examinations that are performed or the acceptance entens for the tests and examinabons Since the systems in which these components are installed must be declared inoperable while they are removed from sonnce for tests and examinations, the systems must be evaluated before they are taken out of service, as required by Technical Specificebons Since the systems are out of service, they cannot influence the operation of the plant, act as precursors to accidents or equipment malfuncbons or affect the consequences of any accidents or equipment malfuncbons Therefore, this change cannot affect any margin of safety defined in the Tech Specs or their Bases I

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a , GRAND GULF NUCLEAR STATION ADMINISTRATIVE PROCEDURE 01-S-06-24 Ravilion s 102 Attachment V Page e of 9 SE No. 96-0098-R00 Page 8 of 9 IIL EnvironisestalEvaluaties Environmental Evaluanon not required per Completed Safety and Environmental Pre-Screening or Applicability Review.

IMPLEMENTAT10N OR PERFORMANCE OF THE ACTION DESCRIBED IN 1NE EVALUATED DOCUMENT:

A. Environmental Protection Plan i

1. Will require a change in the Environmental Protection Plan. U Yes E No BASIS: The Environmental Protecten Plan does not address the performance of tests and examinabons to comply with ASME Code Sedion XI, nor does it address the intervals at which such tests and exammatons are performed, the acceptance cntena for such tests and examinanons, or the plant conditions necessary for such tests and exammanone.

B. Unreviewed Environmental Question l

1.. Concerns a matter which may result in a significant increase in any adverse O Yes environmental impact previously evaluated in the Final Environmental Statement (FES)

W No as modified by the NRC staffs testunony to the Atomic Safety and Licensing Board (ASLB), supplements to the FES, environmental impact appraisal, or in any decisions of ,

the AC' B. j BASIS: The Final Environmental Statement does not address the performance of tests and examinations to comply with ASME Code Section XI, nor does it address the intervals at which such tests and exammations are performed, the acceptance cntena for such tests and examinatens, or the plant conditions necessary for such tests and examinabons.

2. Concerns a significant change in effluents or power level. O Yes 3 No BASIS: These tests and examinsbons can be performed only when the system they are

, connected into is out of sonnce Since the system is out of service, performance of these

! tests and examinations has no effect on affluents or power level.

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3. Concems a matter not previously reviewed and evaluated in the documents specified in O Yes i III.B. I above, which may have a significant environmental impact. g No BASIS: The Final Environmental Statement does not address the performance of tests and l examinatens b comply with ASME Code Section XI, nor does it need to. This change I concems only the times at which tests and exammations specified in ASME Code Secten XI l are performed The Pump and valve IST requirements given in ASME Bosler and Pressure  !

Vessel Code Section XI are intended to provide assurance that pumps and velves which 1 perform a specdic functon in shutting down the reactor or ratigatmg the consequences of a l l accdont are capable of performmg their functions when called upon to do so. This change does not change the frequency at which they are performed, nor does it change the types of tests and examinatens that are performed or the acceptance cnteria for the tests and examinations.

Although not Socifically stated in the Final Environmental Statement or the Environmental Protection Plan, the otwious expectaten and assumed requirement in both documor.ts is that the plant will continue to be operated in accordance with the Technical Speedications and I Technical Requirements Manual. As discussed in ll.A.1 above, this change does not require a change to the GGNS Technical Specificatens, and all tests and examinations will continue to be performed under the provisens of the Tech Specs. l I

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