ML17265A766

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1999 Rept of Facility Changes,Tests & Experiments Conducted Without Prior NRC Approval for Jan 1998 Through June 1999, Per 10CFR50.59.With
ML17265A766
Person / Time
Site: Ginna Constellation icon.png
Issue date: 06/30/1999
From: Mecredy R
ROCHESTER GAS & ELECTRIC CORP.
To: Vissing G
NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9910290070
Download: ML17265A766 (99)


Text

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J REGULA'JQ!Y INFORMATION DISTRIBUTI~YSTEM (RIDE)

ACCESSION NBR:9910290070 DOC.DATE: 99/06/30 NOTARIZED: NO DOCKET FACIL:.50-244 Robert Emmet Ginna. Nuclear Plant, Unit 1, Rochester G

05000244 AUTH.NAME AUTHOR AFFILIATION MECREDY,.R.C.

Rochester Gas

&, Electric Corp.

RECIP.NAME RECIPIENT AFFILIATION VISSING,G.S.

SUBJECT:

"1999 Rept of Facility Changes, Tests 8 Experiments Conducted Without Prior NRC Approval For Jan 1998 througn June 1999,"

per 10CFR50.59.With 991020 ltr.

NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).

DISTRIBUTION CODE:

IE47D COPIES RECEIVED:LTR

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ENCL

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SIZE:

TITLE: 50.59 Annual Report of Changes, Tests or Experiments Made W/out Approv 05000244

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RECIPIENT ID CODE/NAME VISSZN INTERNA : FILE CENTER EXTERNAL: NOAC COPIES LTTR ENCL 1

1 RECIPIENT ID CODE/NAME NRC PDR 1

1

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RGN1 FILE 01 COPIES LTTR ENCL 1

1 1

1 NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROI.

DESK (DCD)

ON EXTENSION 415-2083 I

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LTTR 5

ENCL 5

AHD A Subsidiary of RGS Energy Group, Inc.

ROCHESTER GAS ANDElECTRIC CORPORATION

~ 89EASTAVENUE, ROCHESTER, N.Y. Id6d9CXNI

~ 716 5d6.2700 www.rgo.corn ROBERT C. MECR EDY Vice President Nvdeor Operations October 20, 1999 U.S. Nuclear Regulatory Commission Document Control desk Attn:

Guy S. Vissing Project Directorate I-1 Washington, D.C. 20555

Subject:

Report ofFacility Changes, Tests, and Experiments Conducted Without Prior Commission Approval R.E. Ginna Nuclear Power Plant Docket No. 50-244

Dear Mr. Vissing:

The subject report is hereby submitted as required by 10 CFR 50.59(b). The enclosed report contains descriptions and summaries ofthe safety evaluations conducted insupport ofproposed changes to the facilityand procedures described in the UFSAR and special tests, from January 1998 through June 1999, performed under the provisions of 10 CFR 50.59.

Very ly yours, Robert C. Mecredy Attachment 9910290070 990630 PDR ADOCK 05000244 R

PDR

P

Mr. Guy S. Vissing (Mail Stop SC2)

Project Directorate I-1 Division ofReactor Projects I/II Office ofNuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King ofPrussia, PA 19406 U.S. NRC Ginna Senior Resident Inspector

1999 REPORT OF FACILITYCHANGES, TESTS, ANDEXPERIMENTS CONDUCTED WITHOUTPRIOR NRC APPROVAL

'I FOR JANUARY 1998 THROUGH JUNE 1999 UNDER THE PROVISIONS OF 10 CFR 50.59 R.E. GINNANUCLEARPOWER PLANT DOCKET NO. 50-244 ROCHESTER GAS ANDELECTRIC CORPORATION DATED OCTOBER 20, 1999

..9910290070

SEV-1090 TECHNICALSPECIFICATION BASES CHANGE FOR SCREENHOUSE BAY LOWER TEMPERATURE LIMIT The purpose ofthis safety evaluation is to address changing the Technical Specification Bases for LCO 3.7.8.

This change is being made to better correlate the lake (i.e., ultimate heat sink) environmental conditions with plant operations.

Specifically, the minimum screenhouse bay operability requirements willbe changed. Revision 1 ofthis evaluation changed the screenhouse bay temperature from "Temperature z 35 F..." to "Temperature a 32'F..." inaccordance with a sensitivity analysis. Revision 2 ofthis evaluation supports a change inthe minimum operating temperature ofthe service water system from 32'F to 30'F.

The probability ofoccurrence ofan accident previously evaluated inthe SARis not increased, because the change does not impact the capability to meet the accident analysis nor does it introduce any effects that could increase the probabilityofan accident. Inaddition, the reduction inthe temperature does notadversely impact the abilityofany equipment to perform their intended safety function.

The consequences ofan accident previously evaluated inthe SAR are not increased, because the radiological consequences meet the required acceptance criteria, thus the consequences are acceptable. This change does notintroduce the possibility ofan accident or equipmentmalfunction ofa different type than previously evaluated inthat the change affects onlythe parametric value used by current analyses.

The probability ofoccurrence ofa malfunction ofequipment important to safety previously evaluated inthe SARis not increased, because the change affects aparameter associated withthe SW system fluid and is minor in nature. A design analysis evaluated the impact ofthe fluid temperature change from32'F to 30'F. The results did not show a reduction ofthe structural integrityofany components relied upon and hence the designbasis would notbe affected bythe reduction in temperature limitto 30'F.

The consequences ofa malfunction ofequipment important to safety previously evaluated inthe SARare not increased, because the change proposed does not affect the ability ofanyequipmen relied upon to mitigate consequences fromperforming their functions. The structural integrityof criticalcomponents and their capability is not impacted bythe SWtemperature.

Revision 1 ofthis evaluation examined the impactofa30'F SW temperature on containment fan cooler performance and the affect on PCT, withthe limitingcase ofPCT'remaining less than 2200'F which is the approved criteria for PCT. This is documented in the UFSAR, Section 6.2.2.1.

The possibility ofan accident ofa different type than evaluated previously in the SAR is not created, because the change does not introduce any new initialconditions, or make any change to

the actuation ofaccident mitigating equipment.

The possibilityofamalfunction ofequipment important to safety ofa differen type than evaluated previously inthe SARis not created, because the temperature reduction from32'F to 30'I will not reduce any ofthe performance characteristics ofcomponents, such as valves, pumps, or heat exchangers, and willnot affect the structural integrityor stress levels ofpiping or pipe supports.

Components receiving service water flowforthe purpose ofremoving heat fromthe other fluid medium, are not adversely impacted and not subject to freezing, since the fluidmedium is air, oil or treated water.

This change does not reduce the margin of safety as defined in the basis for the Technical Specifications, because the slight impact upon PCT does not result in a PCT above the criteria basis.

Since all acceptance criteria are met there is no reduction in the margin ofsafety.

SEV-1094 REPLACEMENT OF RTD INPUT MODULES INTHE REACTOR PROTECTION RACKS The electronic components used to generate the T, and b,T signals inthe Reactor Protection System (RPS) are going to be changed to replace the aging loop modules which have no available replacements. This willrequire the removal of20 Foxboro H-linemodules whichwillbereplaced with 24 modules manufactured by NUS. The new module arrangement willconsist of 16 Resistance-to-Current(R/Q converters, and eight TimeDomain Modules (all safety grade analog devices). Each protection channelwill have fourMconvertors thatwillbeused forthe conversion ofHot leg and Cold leg temperatures in the Reactor Coolant System (RCS), and two Time Domain Modules that willbe used to condition the RCS temperature inputs into T,, and b,T signals. One additional function ofthe TimeDomain Modules willbe to provide the required lag time associated withthe temperature signal. The insertion ofinstrument loop lag time provides a compensating factor forthe extremely fast responding loop RTDs withrespect to the rest ofthe instrument loop. The lag timefactor was part ofthe original instrument loop response calculation forboththe T, and hTsignals. Thesignaloutputsofthe TimeDomainModuleswillbeidentical to the outputs ofthe existing modules being removed, including lag time, and therefore willhave no impactonthefunctionoftheloop downstreamofthenewmodules.

The T, anddT temperature loop forChanne12 oftheReactorProtection System have been modified under the firstphase of PCR 97-026.

The probabilityofoccurrence ofan accident previously evaluated inthe SARis not increased by this proposed modification. The change does not introduce any new failure modes or effects into the affected instrument loop nor does it functionally modifytheloop(including delaytimes, setpoints and uncertainties) or associated RPS and control systems in any way.

The consequences of an accident previously evaluated in the SAR is not increased by this proposed modification. The proposed change does not create any new equipment interactions.

Because there are no changes in loop failure modes and effects (note that the replacement equipment is also analog) and no new equipment interactions are added, the change cannot lead to a new type ofmalfunction.

The probability ofoccurrence ofa malfunction ofequipment important to safety previously evaluated inthe SARis not increased. The change does not introduce any new failure modes or effects into the affected T,,or 4T instrument loops nor does itfunctionally modifythe loop or associated RPS and control systems in any way not originally designed for.

The consequences ofamalfunction ofequipment important to safety previously evaluated inthe SARis not increased. The change does not introduce any new failure modes or effects intothe affected T, and hT instrument loops nor does itfunctionallymodifythe loop or associated RPS and control systems in any way not originally designed for.

The possibilityforan accident ofa diferent typethan any evaluated previously inthe SARis not created.

The proposed change does not create any new equipment interactions.

The possibilityofa malfunction ofequipment important to safety ofadifferen typethen evaluated previously inthe SARis not created. The proposed change does not create any new equipment interactions.

The margin ofsafety as defined inthe basis forany technical specification is not reduced by this proposed modification. The Overpower and Overtemperature setpoints, the process bywhich they are generated, and the total RPS delay time are all unaffected by the change.

SEV-1100 RWST ACCIDENTANALYSISUPPER TEMPERATURE LIMIT The accident analysis assumes aRWST temperature range of60'o 80'F. Recent temperature measurements inthe AuxiliaryBuildingindicate theupper temperature limitshould be increased.

This evaluation documents the efforts done to increase the upper limitfrom 80'o 104'F.

Increasing the assumed water temperature Rom 80'o 104'F does not change the functionofthe RWST, SI system or spray system. The effect ofthtemperature increase on the SI system and spray system interms ofavailableNFSH has been evaluated and determined not to be a concern.

The probability ofoccurrence of an accident is not increased by the assumption ofRWST temperature.

The RWST is not an accident precursor and therefore the change in maximum allowable temperature willnot affect the probability ofoccurrence for any accident analysis described in the UFSAR.

The consequences ofan accident have not increased because the acceptance criteria for the accident are stillmet. The peakcontainmentpressure as a result ofthis change remains below the limitof60 psig and therefore the control room and off-site dose radiological consequences due to the increase inRWST temperature stillsatisfy the limitsestablished by GDC 19 and 10CFR100.

The probability ofoccurrence ofa malfunction is not increased by the assumption ofRWST temperature.

The temperature increase from80'o 104'F is withinthe design ofthe affected systems and therefore there is no change in the likelihood offailure.

The consequences ofa malfunction have not increased because the acceptance criteria forthe accident are stillmet. The peak containmentpressure as a result ofthis change remainsbelow the limitof60psig and therefore the control room and off-site dose radiological consequences due to the increase inRWST temperature stillsatisfy the limitsestablished by GDC 19 and 10CFR100.

Increasing the assumed RWST temperature by24'F does not cause a differen type ofaccident than previously evaluated. The temperature change slightly affects thethermal hydraulic roperties ofthe water which would not cause a new type ofaccident.

Increasing the assumed RWST temperatureby24'F does not cause a differen type ofmalfunction than previouslyevaluated.

The temperature change slightly affects thethermal hydraulic properties ofthe water which would not cause a new type equipment malfunction.

The margin ofsafety is between the acceptance criteria and the ultimate failure point. 60 psig is the acceptance criteria forcontainment. This value has not been exceeded by increasing the upper limiton RWST temperature.

Therefore, there is no chang'e in the margin ofsafety.

SEV-1102 PCN ¹ 97-4346 SAFETY EVALUATION This Safety Evaluation describes proposed changes to test procedure PT-60.4. This procedure is used to test the performance ofthe ADiesel Generator Lube Oil and Jacket Water coolers coincident withthe monthlyADiesel Generator run doneunder PT-12.1. The foulinginthe Diesel Generator Aheat exchangers is determined analytically fromPT-60.4 test measurementsusing a

welldeveloped methodology. Theuncertaintyin the determination offoulingis stronglydependent on the service water temperature difference acoss the coolers. Inorder to reduce theuncertainty inthe fouling,the service water willbe throttled to approximately 250 gpm. The followingchanges are evaluated:

PCN ¹ 97-4346 adds steps to PT-60.4 to unlock and throttle globe valve 4671 during testing of the Diesel Generator A coolers.

Diesel Generator A will be declared INOPERABLEfor the duration oftime that valve 4671 is unlocked and throttled.

PCN¹ 97-4346 adds a precaution to PT-60.4 to have ari observer continually monitor the lubricating oil and jacket water outlet temperatures fromDiesel Generator A, and record the values on a ten-minute frequency, whenever the engine is running and the service water is throttled. Inthe event that thejacket water temperature rises above the alarm setpoint of182'F or the lubricating oiltemperature rises above the alarm setpoint of195'F, the HCO is informed and test personnel immediately open valve 4671. Test personnel also immediately open valve4671 iftheHCO receives a high-temperature alarm on the MCB.

Allother proposed changes toPT-60.4 are inconsequential.

They involveinstallation of additional non-intrusive instrumentation (surface-mounted RTDs) and changes to the frequency and duration at which data is taken. These changes are intended to further improve the accuracy ofthe tests.

The proposed changes do not increase the probability ofoccurrence ofan accident previously evaluated inthe SAR. The emergency dieselgenerator is not an accident initiator, and temporarily throttlingservice water to the diesel generatorcoolerswillnotchange the configurationofany othe system in such a way as to impact the probability ofanother system initiating an accident.

The proposed changes do not increase the consequences ofan accident previously evaluated in the SAR. Diesel Generator A,although INOPERABLE,is expected to function normally, and can be returned to OPERABLE status by opening and lockingvalve 4671. Inaddition to the normal MCBalarm, Diesel Generator Awillbe continually monitored locally to verifythat thelube oiland jacket water temperatures do not exceed the alarm setpoint values. Inthe event that temperatures reach alarm setpoints, test personnel willtake immediate action to open valve 4671. Therefore,

the probabilityoffailureofDiesel Generator Aisno higher than itisduring theregular monthlyPT-12.1 Surveillance Test.

Theproposed changes do notincrease the probabilityofoccurrenceofamalfunction ofequipment important to safety. The tested emergency diesel generator can be restored to operable status immediately by opening and locking valve 4671.

Since this corresponds to the analyzed configuration ofthe plant, there is no increased probability ofmalfunction ofthe diesel generator or any other equipment.

The proposed changes do not increase the consequences ofa malfunction ofequipment important to safety previously evaluated inthe SAR. Accident analyses already assume the loss ofa diesel generator.

The proposed changes do not increase theprobability ofan accident ofa different type than any evaluated previously inthe SAR. The proposed changes involveminor modifications t a test that is routinely carried out. The most severe occurrencewould be the tripping ofDiesel Generator A to prevent itfrom overheating.

Contingent actions stemming from a diesel generator trip are already available.

The proposed changes do not increase thepossibility ofa malfunction ofequipment important to safety ofa different type then evaluated previously inthe SAR. The Diesel Generator Alube oil and jacket water temperatures willnot be allowed to rise above the currently established alarm setpoints. Ithas been established by the vendor that these are acceptableoperating temperatures for the diesel engines.

The margin ofsafety as defined inthe basis forany technical specification is not reduced, since no Technical Specifications are violated.

Since the normal configuration ofthe system can be immediately re-established as necessary to provide adequate cooling to the diesel generator, there is no reduction in any safety margins.

SEV-1103 VACUUMFILLOF THE REACTOR COOLANT SYSTEM Industry wide use ofthe vacuum fillmethod ofincreasing thereactor coolant system(RCS) level from mid loop to the narrow range on the pressurizer is to be'evaluated.

This procedure isto beused during mode 5 priorto and during the final RC loop fillprocess. It willbe installed onlyduring this process and willbe removed when RCS refillis complete. The vacuum fill process willbeincorporated intprocedures 0-2.3.1 and 0-1B. The present method ofRCS system fillrequires along and complicated vent procedure. This modification wilallow a vacuum to drawn on the RCS when atmidloop inorder to allowtheRCS to be filledwithoutthe need for venting.

The initialconditions forRCS vacuum fillare established during RCS lowloop conditions. The RCS level is to be maintained between 10-12 inches indicated loop level and RCS temperature willbe maintained <85'F throughout the vacuum venting process. Lowloop procedure 0-2.3.1 willbe in effect, the level band restricts RHR flowto 800 gpm.

Thevacuum operation willconsist ofa vacuum pump connected via2 inch diameter vacuum rated hose to the pressurizer reliefoutlet piping to the pressurizer relieftank(PRT). There willbe an option vacuum hose forthereactor vessel head vent. The pressurizer PORV andBlock valves will be open to allowthe PRT gas space and pressurizer relieftailpipes to be connected to the RCS.

The pressurizer and PRT vent manifolds willsupply the vacuum taps for reactor vessel level sightglass and RCS loop level instrumentation.

TheRCS vacuum vent and fillprocedure willmaintain positive control over the RCS vents and the lowtemperature overpressure system(LTOP) alignment. The procedure maintains control over all equipment that can inject intothe RCS and increase its pressure. This assures RCS boundary protection at lowtemperatures, therefore the initialconditions and probabilityofoccurrence for any accident analysis previously evaluated in the UFSAR have not changed and are valid.

The RCS vacuum vent and fillprocedure maintains control ofreactor coolant boron, density, or operating temperature. The procedure monitors the dilutionand boration paths to theRCS. The vacuum process willnotinfluence coolantboronconcentration, therefore the initialconditions and consequences ofan accidentpreviously described inthe UFSARforreactivityinsertion have not changed and are valid.

The wallthickness ofthe pressurizer, steam generators and U-tubes, reactor coolant pumps and associated components exposed to the vacuum is sufficient to maintain the integrityofthe systems during vacuum venting, and after the fillprocess is complete.

The integrityofthe reactor coolant pump seals is assured bymaintaining a positive pressure at the number 1 seal inlet area. The pressurizer relieftankis designed to withstand a fullvacuum. The tank is equipped withan internal support forthe rupture diskto prevent the damage to the disk.

Therefore the integrity ofthe RCS remains unchanged and the probability ofoccurrence ofa malfunction ofequipment important to safety is not increased.

The containmentisolation system willremainunaffected bythis change. The systemwill stillbe able to achieve containment closurewithin the allowed 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time period ofgeneric letter 88-17, and be capable ofpreventing a radiation release within 10 CFR 100 limits. Therefore the abilityto isolate containment during reduced RCS inventory operations remains unchanged and the probability ofoccurrence ofa malfunction ofequipment important to safety is not increased.

~

The abilityofthe ResidualHeat Removal system to provide forcore cooling when the RCS is in a reduced inventory condition willnot change. TheNPSH available fortheRHRpumps is greater than required, per Design Analysis DA-ME-97-080,Rev3, therefore the capability ofRHRsystem to provide core cooling willnot be adversely affected.

The WCAP-11916 (section 2.5) was reviewed to verifythat operating in midloop with the RCS at a vacuum did not invalidate its analysis. The analyses forvortexformation were most sensitive to fluidvelocitywiththe density and viscosity ofthe fluidas secondary affects. None ofthese parameters areaffected by the RC

- being under a vacuum.

The analysis therefore remains valid. There is no increase in the consequences ofa malfunction previously evaluated in the UFSAR.

TheRHR, charging, and safety injection systems willallbe lined up and controlled per Operations procedure 0-2.3.1 "Drainingand Operating atReduced Inventory intheReactor Coolant System".

This procedure implements RGB's response to generic letter 88-17 concerns.

The RCS is maintained inan analyzed condition per WCAP 11916. The RCS and mitigating systems are lined up and operating per established procedures. Therefore this system configuration and procedure does not create the possibilityforan accident ofa differen type than any evaluated previously in the UFSAR.

Withthe steam generator intact and the pressurizer manway installed, the criteria is met forthe RCS intact configuration. This configuration wa analyzed and is one ofthe configuration that WCAP-11916 and Generic Letter 88-17 reviews. Therefore, the possibilityofa malfunction of the RCS boundary ofa different type than evaluated previously in the UFSAR is not created.

AKYPIPE analysis (noted on "Expeditious Actions" response to the NRC, dated January 4, 1997) oftheRHRsystem verifiedthat the gravityfeed method would place approximately 7000 gallons ofwater in the RCS ifinitiated within 16 minutes ofthe event and assuming an intact, unvented RCS, that would pressurize according to the WCAP 11916 fi.3.3.1-1. This was based on the decay heat load at 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shutdown. The vacuum fillevolution is taking place at greater than 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> after shutdown, the estimated time to saturationis approximately27 minutes and thereis additional time needed to buildup pressurein the RCS. The openPORVs and having one steam generator filledwillfurther delay the increase inRCS pressure. Therefore additional time is available for the operators to increase RCS level using gravity feed. A pressure of approximately 42 psia was found to stop gravity feed fiowfromthe RWST. The finalrecovery action ofrestarting RHRwould occur after level is increased. Performing the RCS vacuum vent and fillunder these conditions does not reduce the margin ofsafety as defined inthe basis forany Technical Specification.

C'

SEV-1104 PCN 8 97-4347 SAFETY EVALUATION This Safety Evaluation describes proposed changes to test procedure PT-60.5. This procedure is used to test the performance ofthe Diesel Generator B Lube Oil and Jacket Water coolers coincident withthe monthlyDiesel Generator B run doneunderPT-12.2.

The foulingintheDiesel Generator B heat exchangers is determined analytically fromPT-60.5 test measurements using a welldeveloped methodology. Theuncertaintyin the determination offoulingis strongly dependent on the service water temperature difference across thecoolers. Inorder to reduce theuncertainty inthe fouling, the service water willbethrottled to approximately 250 gpm. The followingchanges are evaluated:

PCN 8 97-4347 adds steps to PT-60.5 to unlock and throttle globe valve 4672 during testing of the Diesel Generator B coolers.

Diesel Generator B will be declared INOPERABLEfor the duration oftime that valve 4672 is unlocked and throttled.

PCN 8 97-4347 adds a precaution to PT-60.5 to have an observer continually monitor the lubricating oil and jacket water outlet temperatures from Diesel Generator B, and record the values on a ten-minute frequency, whenever the engine is running and the service water is throttled. Inthe event that thejacket water temperature rises above the alarm setpoint of182'F or the lubricating oiltemperature rises above the alarm setpoint of195'F, the HCO is informed and test personnel immediately open valve 4672. Test personnel also immediately openvalve4672iftheHCO receives a high-temperature alarm on the MCB.

Allother proposed changes to PT-60.5 are inconsequential.

They involveinstallation of additional non-intrusive instrumentation (surface-mounted RTDs) and changes to the frequency and duration at which data is taken. These changes are intended to further improve the accuracy ofthe tests.

The proposed changes do not increase the probability ofoccurrence ofan accident previously evaluated inthe SAR. The emergency diesel generator is not anaccident initiator, and temporarily throttlingservicewater to the diesel generatorcoolerswill not change thecon6guration ofany other system in such a way as to impact the probability ofanother system initiating an accident.

The proposed changes do not increase the consequences ofan accident previously evaluated in the SAR. Diesel Generator B, although INOPERABLE,is expected to function normally, and can be returned to OPERABLE status by opening and lockingvalve 4672. Inaddition to the normal MCBalarm, Diesel Generator Bwillbe continuaHymonitored locallytoverifythat the lube oiland jacket water temperatures do not exceed the alarm setpointvalues. Inthe event that temperatures reach alarm setpoints, test personnel willtake immediate action to open valve 4672. Therefore, 0

the probabilityoffailureofDiesel Generator B is no higher thanit is during the regular monthly PT-12.2 Surveillance Test.

Theproposed changes do not increase theprobability ofoccurrenceofamalfunction ofequipment important to safety. The tested emergency diesel generator can be restored to operable status immediately by opening and locking valve 4672.

Since this corresponds to the analyzed configuration of the plan, there is no increased probability ofmalfunction ofthe diesel generator or any other equipment.

The proposed changes do not increase the consequences ofa malfunction ofequipment important to safety previously evaluated inthe SAR. Accident analyses already assume the loss ofa diesel generator.

The proposed changes do not increase the probabilityofan accident ofa difFerent type than any evaluated previously inthe SAR. The proposed changes involveminormodification to a test that is routinely carried out. The most severe occurrencewould be the tripping ofDiesel Generator B to prevent itfrom overheating.

Contingent actions stemming from a diesel generator trip are already available.

H The proposed changes do not increase the possibilityofa malfunction ofequipment important to safety ofa di6erent type then evaluated previously inthe SAR. The Diesel Generator B lube oil and jacket water temperatures willnot be allowed to rise above the currently established alarm setpoints. Ithas been established bythe vendor that these are acceptable operating temperatures for the diesel engines.

The margin ofsafety as defined inthe basis forany technical specification isnotreduce, since no Technical Specifications are violated.

Since the normal configuration ofthe system can be immediately re-established as necessary to provide adequate cooling to the diesel generator, there is no reduction in any safety margins.

SEV-1105 VACUUMEFFECTS ON RCS INSTRUMENTATION DURINGRCS VACUUMVENT ANDFILL The effect ofhaving a vacuum on the Reactor Coolant System(RCS) instrumentation during the RCS vacuum vent and fillevolution are to be evaluated. The instrumentation willbe exposed to RCS temperatures of70- 85'F. The pressure willrange from atmospheric to 28 inches ofHg vacuum or 0.948 psia. The RCS loop willbe initiallyat the midloop level. This level is 10 inches using local level indication and is at the 246'10" elevation. The time duration ofthe exposure to vacuum is less than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Once the RCS level is inthe 80% (180 inches) wide range in the pressurizer, the vacuum willbe removed and the system willbe returned to normal operational pressures.

The RCS vacuum vent and fill procedure wilmaintain positive control over the RCS vents and the lowtemperature overpressure system(LTOP) alignment. The procedure maintains control over all equipment that can inject into the RCS and increase its pressure. This assures RCS boundary protection and RCS instrument operability at lowtemperatures per UFSAR chapter 5.2.2. The RCS instrument system willcontinue to accurately monitor and display the process variables needed to verifyRCS parameters.

Therefore the initialconditions and probabilityofoccurrence for any accident analysis previously evaluated in the UFSAR have not changed.

The RCS vacuum vent and fillprocedure maintains control ofreactor coolant boron, density, and operating temperature. The procedure monitors the dilutionand boration paths to the RCS. The RCS instrument system willcontinue to accurately monitor and display the process variables needed to verify RCS parameters.

The vacuum process willnot influence coolant boron concentration, therefore theinitial conditions and consequences ofan accidentpreviouslydescribed inthe UFSARforreactivity insertion in chapters 15.4.4.2.2 or 15.4.4.2.6 have not changed and are valid.

The wall thickness of the RCS process instrumentation and sensing lines and associated components exposed to the vacuum is sufficient to maintain the integrityofthe systems during vacuum venting, and after the fill proces is complete. The integrityofthe reactor coolant pump seal instrumentation is assured by maintaining apositive pressure at the number one seal inlet area.

The pressurizer relieftank instrumentation is designed to withstand a fullvacuum. The tank is equipped withan internal support forthe rupture diskto prevent the damage to the disk. Therefore the integrityofthe RCS instrumentation remains unchanged and the probabilityofoccurrence of a malfunction ofequipment important to safety is not increased.

The containment isolation system and its associated instrumentation willremainunaffectedby this change. The system willstillbe ableto achieve containment closure withinthe allowed2 hour time period ofgeneric letter 88-17, and be capable ofpreventing a radiation release within 10 CFR100 J

t't 0

limits. Therefore, the abilityto isolate containment during reduced RCS inventory operations remains unchanged and the probabilityofoccurrence ofamalfunction ofequipment important to safety is not increased.

The abilityofthe Residual Heat Removal system to provide forcore cooling when the RCS is in a reduced inventory condition willnot change. TheNPSH available forthe RHR pumps is greater than required, therefore the capability ofRHRsystem to provide core coolingwillnot be adversely affected. The WCAP-11916 (section 2.5) was reviewed to verifythat operating inmidloop with the RCS at a vacuum did not invalidate its analysis. The analyses forvortex formationweremost sensitive to fluidvelocity with the density and viscosity ofthe fluid as secondary affects.

TheRCS and RHRinstrument systems willcontinue to accurately monitor and display the process variables needed to verifytheir parameters. None ofthese parameters are affected by the RCS being under a vacuum.

The analysis therefore remains valid. There is no increase in the consequences ofa malfunction previously evaluated in the UFSAR.

TheRHR, charging, and safety injection systems willall be lined up and controlled per Operations procedure 0-2.3.1 "Drainingand Operating at Reduced Inventory intheReactor Coolant System".

This procedure implements RGB's response to generic letter 88-17 concerns.

The RCS is maintained inan analyzed condition per WCAP 11916. The RCS and mitigating systems are lined up and operating per established procedures. Therefore this system configuration and procedure does not create the possibilityforan accident ofa differen type than any evaluated previously in the UFSAR.

Withthe steam generator intact and the pressurizer manway installed, the criteria is met forthe RCS intact configuratio. This configuration wa analyzed and found acceptable inWCAP-11916.

Therefore, the possibilityofamalfunction ofthe RCS boundary ofa differen type than evaluated previously in the UFSAR is not created.

The RCS vacuum vent and fillprocess does not require a changeto Ginna Technical Specifications.

RCS pressure and temperature limitsas stated inthe Pressure TemperatureLimits Report(PTLR)

're not exceeded. The RCS and RHR instrument systems willcontinue to accurately monitor and display the process variables needed to verifytheir parameters.

The shutdown requirements and PORV operability limitsforthe RCS are maintained. The margin ofsafety forthe reactor coolant pressure boundary as defined by the ASME code forwall thickness, stress limits, integrity of systems and components is maintained.

SEV-1108 CYCLE 27 RELOAD Cycle 27 consists of41 new fuel assemblies from feed regions 29A, 29B, 29C, and 29D. This safety evaluation is valid for an end-of-cycle 26 burnup of15,200 to 16,200 MWD/MTUand Cycle 27 burnup not to exceed 16,517 MWD/MTUwithout additional analysis.

Cycle 27 characteristics are described inmore detail in the "Reload Safety Evaluation-Cycle 27, Redesign".

The fuel assemblies forCycle 27 aremechanically the same as the Cycle 26 fuel assemblies except for the following.

The use ofannular pellets in the axial blankets, Areduction inbackfil pressure inIntegral Fuel Burnable Absorber gFBA)rods to 100 psig, Grooved top and bottom fuel rod end plugs, 3-'tab inconel grids, 5.

New top nozzle spring pack design.

The Cycle 27 reload willnot increase the probability ofoccurrence ofan accident because the reload core does not affect accident initiators or equipment operation. The reload core does not cause a pipe to break or equipment to malfunction. Therefore, the reload core can not increase the probabilityofan accident. The fuel design change satisfies existing design criteria; therefore, the probability offailure does not increase.

Gap reopening does not acct accident initiators.

The Cycle 27 reload does not increase the probabilityofamalfunction ofequipment because the reload core does not acct equipment operation. The reload core does not cause equipment to malfunction. The fuel design change satisfy existing design criteria; therefore, the probabilityof failure does not increase.

Gap reopening is not expected to lead to fuel failure. Violatingthe gap reopening SAFDL criteria does not result in exceeding the 17% oxidation limit.

The Cycle 27 reload does not increase the consequences of an accident because the core characteristics are bounded by parameters assumed inthe accident analysis. When deviations occurred, reanalysis was performed to show the acceptance criteria was stillsatisfied. The fuel, assembly changes do not degrade fuel performances.

The resulting changes are still within acceptable ranges.

Gap reopening could affect the 17% oxidation limit; however, this is not possible until the screening limit has been reached.,Analysis has been performed which demonstrates compliance with the limitfor all ofCycle 27.

The Cycle27 reload does not increase the consequences ofmalfunction ofequipmentbecause the core characteristics areboundedby parameters assumed inthe accident analy'sis. When deviations occurred, reanalysis was performed to show the acceptance criteria was stillsatisfied. The fuel assembly changes do not degrade fuel performances.

The resulting changes are still within acceptable ranges.

Gap reopening does not acct the consequences ofequipment malfunc'tion.

For example, the consequences ofa pump failure is not afFected by gap reopening.

The Cycle 27 reload and fuel assembly changes do not cause a new type accident because the core parameters are boundedby those assumed inaccident analysis and design parameters are still within the'assumed ranges.

Gap reopening is not an accident initiator.

The Cycle 27 reload and fuel assembly changes do not cause anew type ofmalfunction because the core parameters are bounded by those assumed inaccident analysis and design parameters are still within the assumed ranges.

Previous analysis assumed no gap reopening for simplicity.

Analyses with gap reopening show acceptable consequences.

Therefore, this condition is acceptable provided continued compliance with the 17% oxidation limitis maintained.

Sincethe assumptions in the safety and accident analysis including those related to the core design arebounding forthe Cycle 27 reload, the conclusions inthe GinnaUFSARremainappropriate and the regulated acceptance criteria for the accident analysis has not been violated. There is no reduction in the margin ofsafety as defined in the basis for any Technical Specification.

No gap reopening is a Westinghouse design criteria used to simplifythe design process. Analyses withgap reopening show all aspects ofplant safety analyses remain bounding. The screening criteria provides the point at whichcompliance withthe 17% oxidation needs to bere-evaluated.

The plant specific analysis demonstrates continued compliancewith the 17% oxidation criterion throughout Cycle 27.

SEV-1109 NEW PROCEDURE PT-60.3 "CONTAINMENTRECIRCULATION FAN COOLER PERFORMANCE TEST" This Safety Evaluation describes new procedure PT-60.3A. This procedure was developed to provide an simplifiealternative to procedure PT-60.3.

Simplification wa desired to reduce the number ofpeople and amount ofequipment that would be required incontainment to facilitate on-power testing. The new procedure onlyprovides information necessary to determine the fouling ofthe Containment Recirculation Fan Coolers (CRFC). ItDOES NOT test the CRFC motor coolers.

The actions in the procedure that have potential safety-significance include:

1.

Throttlingthe servicewater flowtoeachCRFC downto-300 gpm fromthe usual value of-1200 gpm. This is only done to one CRFC at a time, and the CRFC is declared inoperable.

2.

Isolationofservicewaterflowto the fan motor cooler ofthe CRFC being tested. Again, the CRFC is declared inoperable when the motor cooler flow is isolated.

3, Installation and removalofintrusivetestinstrumentation(differentialpressurecells).

This willperiodically cause the control room operators to get lowflowalarms on FIA-2033, FIA-2034, FIA-2035, and FIA-2036.

The operators are informed before these manipulations are done.

4.

Positioning and repositioning ofA-3.3 Containment Isolation Boundaries.

PT-60.3A does not increase the probability ofoccurrence ofan accident previously evaluated in the SAR. The CRFCs and associated containment HVACequipment are not accident initiators.

PT-60.3A does not increase the consequences ofan accident previously evaluated in the SAR since accident analyses already assume the loss of a train of containment HVAC, and the inoperable duration ofany CRFC willbe much less than the LCO 3.6.6 allowed time of7 days.

PT-60.3A does not increase theprobability ofoccurrence ofamalfunctionofequipment important to safety. PT-60.3A does make a train ofcontainment HVACinoperable, which is already assumed inaccidentanalyses.

Manipulations on other systems, other than the servicewater supply to the inoperable train ofCV HVAC, are not performed as part ofthe PT-60.3A procedure.

PT-60.3A does not increase the consequences ofa malfunction ofequipment important to safety previously evaluated in the SAR. Accident analyses already assume the loss of a train of containment HVAC,and the environmental qualification profile met witha containmentHVAC train out ofservice. Moreover, manipulations ofother systems and equipmentimportant to safety are not performed as part ofthe PT-60.3A procedure, so there is no associated increase in probability or consequences.

PT-60.3A does not increase the probabilityofan accident ofa different type than any evaluated previously inthe SAR. The procedure involves manipulation of servicewater system valves in the supply to an inoperable CRFC, entry into the enclosure ofthe inoperable CRFC, and installation oftest equipment only.

PT-60.3Adoes not increasethepossibility ofa malfunction ofequipment important to safety ofa diFerent type than evaluated previously inthe SAR. The procedure involves manipulation of service water system valves inthe supply to an inoperable CRFC, entry into the enclosure ofthe inoperable CRFC, and installation oftestequipment only. No other equipment is manipulated or expected to malfunction as a result ofthis procedure.

The margin ofsafety as defined in the basis for any technical specification is not reduced. PT-60.3A onlyaffects a single train of containment HVAC (including the associated post-accident

- charcoal system), which is declared inoperable under LCO3.6.6 fortesting. Withthe exception

'ofservice water supply to the inoperable train, no other systems are affected by the testing.

The inoperable CRFC operates during the testing, and the reduced service water flowrate does not have a significant effect on the heat removal capability ofthe CRFC. The operable CRFCs are also available to maintain containment temperature below the normal operating Technical Specification limitof 120'F as defined in LCO 3.6.5.

SEV-1111 FUEL ASSEMBLYREPAIR PROCEDURE RF-73.1 Inorder to repair (reconstitute) selected fuel assemblies the preferred technique is to remove the top nozzle which allows access to the fuel pins. This differsfrompast methods ofreconstitution which involved turning the fuel assembly upside down and removing the bottom nozzle. The removable top nozzlehas been incorporated into Ginna fuel designs and itis desirable to utilizethis method ofreconstitution.

Fuel reconstitution is accomplished byremoving defective rods and replacing them with "dummy" stainless steel rods. The acceptability ofusing a reconstituted fuel assembly inthe reactor is not covered by this safety evaluation as that willbe covered by a revision to the reload safety evaluation.

This evaluation covers the process ofreconstitution only.

The general process forreconstitution is as follows: Once a fuel assembly has beenidentifie as a leaker and the defective pin(s) identified by aUT inspection the fuel assembly is transported to the new fuel elevator. The new fuel elevator willbe outfitted witha special reconstitution basket that is compatible withthe reconstitution tooling. Once the fuel assembly has been placed inthe elevator the elevator willbe raised to a height where the top nozzle locktubes can be removed.

This elevation is approximately 9 feet below the water surface. The locktubes and top nozzle are then removed and the fuel assembly lowered to the rack elevation. Next the defective fuel pins are removed and placed inthe existing failed fuel storage container. Dummy rod(s) are inserted inthe location(s) previously occupied bythe defective pins and the fuel assembly raised again to the 9 footelevation and the top nozzle and locktubes are reinstalled. The assembly is then lowered and transferred to its desired location.

The GinnaUFSAR states that the new fuel elevator isused fornew fuel only. Since this procedure willdeviate &omthat description this safety. evaluation is being prepared to describe the additional use ofthe elevator for fuel repair activities.

Since the assembly to be reconstituted is contained in systems designed to handle its associated geometry and weight theprobability ofa fuel handling accident or any other accident inSARis not increased.

Since the fuel assemblywill bethe only assembly in transit orbeing worked on during reconstitution activities and the activities performed at less than 23 feet ofwater coverage are limitedin scope so as to not damage any fuel pins the consequences of afuel handling accident remain bounded by the evaluated accident.

The probabilityofamalfunction ofequipment important to safetyis notincreasedbecause multiple layers ofadministrative and physical controls areinplaceto maintain sufficient water leve above the fuel assembly at all times.

The consequences ofa malfunction ofequipment important to safety are not increased because sufficient controls have been put inplace to preclude overexposure ofplant personnel as well as the public from reconstitution activities.

The possibility ofanaccident ofa differen type than any previously evaluated inthe SARhas not

, been created because the new fuel elevator has sufficient controls in place to prevent the inadvertentwithdrawal ofa spent fuel assembly fromthe water. Anypossiblebreakage ofasingle fuel rod during the reconstitution process is bounded bythe fuelhandling accident analysis which assumes all rods in a single assembly are failed.

The use ofthe new fuel elevator willnot create the possibility ofa malfunction ofequipment important to safety because the adjusted elevator stop willbe tested priorto placing a spent fuel assembly into it. Since the elevatoris designed forthe weight and geometry ofthe component that is being inserted into it this change does not create the possibility ofits malfunction.

Since fuel handling; water level, boron concentration specifications are allmaintained withintheir Technical Specification limitsthis procedure does not decrease the margin ofsafety as defined in the basis for spent fuel pool technical specifications.

I 0

SEV-1112 ACTIONREPORT 97-1846 DISPOSITION FOR MAINSTEAMLINEA ANDB CRACKREPAIR ATPENETRATION 401 AND 402 As a result ofnew ISIinspectionmethods forintegral attachments to piping/components cracks were discovered inthe gusset welds ofMS penetrations 401 and 402 inside containment. The purpose ofthis safety evaluationis to review the root cause and corrective action taken as a result ofthe cracks and determine ifthe affected systems are operable.

This revision ofthe safety evaluation was performed to update the references to the supporting analysis.

The root cause ofthe cracks was found to be due to poor weldjointdesign, referred to as a tee joint, which caused high residual stresses inthe heat affected zone ofthe weld. Heavy presence

~ of oxides is evidence that the cracks have existed for a long time, possibly from original construction initiation. Cracking inteejoints is a wellknow phenomena(Lamellar tearing) which was identifie inthe late 1960s forlarge sectionstructural members. The literaturereviewed shows cracks starting fi'omthe weld toe and propagating down into the base metal along the heat affected zone. Based on the report, further cracks should not develop since the initiatingcause was the welding stresses, not service induced stresses (fatigue). Allcracks were found at the outer toe of the weld.

The repair process removed gussets which were located adjacent to the cracked weld to allow access to the pipe wallfordefect removal. Cracks were not found in any ofthe area between the outer toes ofthe two filletwelds on either side ofthe three gusset which were removed. The cracks were excavated down to "defect free" base metal and thenrewelded to restore the required pipe wall. Allrepairs were done in accordance withthe original plant construction code. The maximum crack depth was found to be less than 5/8" inall cases and started at the weld toe on the pipe. The removed gussets was not re-installed over the repaired pipe area per PCR 97-089, since they were not required to meet the design basis loads.

The FW system was found to have the same penetration design as the MS except withthinner members and smaller filletwelds. The inspections did not reveal any cracks. Areview was also done ofthe remainder oftheMS and FWsystem forother potential teejoint configurationswhic have the potential forcracks. No other attachments were found whichwerehighlyrestrained and had weld sizes large enough to generate high residual stresses. Athird review was done ofthe remainder oftheplant piping systems and theresults showed that the systems did not have a large enough pipe wall thickness or attachment welds to create the high residual stresses.

The probabilityofoccurrence orthe consequences ofan accident ormalfunction ofequipment important to safety previously evaluated inthe UFSAR are not increased by the proposed repair since the capability oftheMS linepenetrationsto resist design loads has not been reduced beyond what was originally assumed.

The possibilityforan accident or malfunction ofa differen type than evaluated previously inthe UFSAR willnot be created by the proposed repair.

Since the repair meets the original code requirements and design basis, and willnot change the functionofthe penetrations, no new types ofaccidents or malfunctions would be introduced.

The margin ofsafety, as de6ned inthe basis ofany Technical Specification, is not reduced bythe proposed repair since it meets the original design basis and codes.

SEV-1114 CONTAINMENTRECIRCULATlNGFAN COOLER COILREPLACEMENT Theoriginal Westinghouse Sturtevant Containment Recirculating Fan Cooler(CRFC) coils were replaced under EWR 5275 withenhanced design Marlocoils during the 1993 refueling outage.

Piping modifications were als madein the vicinityofthecoils to ease inspection and maintenance ofthe coils.

Alargenumber ofUFSARchanges were made as a result ofthe CRFC coilreplacement since the heat removal from these units a6ects the relevant analyses for high-energy line breaks inside containment (e.g. LOCAs and MSLBs).

.The original safety evaluation forthis EWR was taken to Revision 1, but this later revision was never approved by PORC.

This deficiency was discovered during the Service Water (SW)

System Safety System Functional Inspection (SSFI) performed by Sargent 2 LundyEngineers LLCinApriland May 1997. This deficiency wa documented by an ActionReport. The purpose ofRevision1 was to close out open items identified bythe original safety evaluation. Although a number ofthe Revision 0 open items were addressed byRevision 1, a number ofopen items were still identified by Revision

. Close out ofthese additional open items was documented by inter-oKce correspondence prior to start-up from the 1993 Refueling Outage.

This document serve as the final 10CFR50.59 Safety Evaluation of record for this plant configuration change. As such, itdocuments the actions taken in 1993 priorto plant start-up to close-out allofthe open items identified inthe originalEWR5275 safety evaluation. This safety evaluation willbe applicable to the modification asitwa completed in 1993; itwillNOTattempt to reconcile issues discovered between the time the modification wa completed and the present day. Alladditional changes to the plant subsequent to the CRFC replacement in 1993 would have had their own 10CFR50.59 review/evaluation.

The replacement ofthe CRFCs byEWR 5275 does not increase theprobability ofoccurrence of an accident previously evaluated in the UFSAR.

The CRFCs are used to mitigate the consequences ofdesignbasis piperuptures inside containment. Additionally,during normal plant operations the replacement CRFCs are capable ofperforming the same heat removal function and ventilation function as is performed by the original CRFCs. As such normal operation ofthe CRFCs does not initiate any design basis accidents presently described in the UFSAR.

The replacement ofthe CRFCs byEWR 5275 does not increase the consequences ofan accident previously evaluated in the UFSAR. The replacement CRFCs have enhanced heat remo'val capability when compared to the original CRFCs. Consequently, containment pressurization transient response to design basis accidents is improved. The peak clad temperature analysis is not affected by the CRFCs since the minimum containment back-pressure curve used for the cladding analysis included margin which allows itto stillbebounding when compared to theanalysis withthe replacement CRPCs. The control room and off-site dose radiological consequences due to the reduction inCRFC air flow rateunde design basis accident conditions stillsatisfy the limits established by GDC 19 and 10CFR100.

The replacement ofthe CRPCsbyEWR 5275 does not increase the probabilityofoccurrence of a malfunction ofequipment important to safety as previously evaluated in the UFSAR. The operation ofthe CRFCs under normal operating and design basis accident conditions has notbeen altered and does notdirectlyimpact the probability ofequipmentmalfunction forother components.

Since the normal operating and designbasis containmentpressure and temperature profile are not adversely affected bythe CRFC replacement, the EQ pressure and temperature profiles forsafety related equipment incontainment is stillbounding. The SWflowto safety related loads supplied in parallel withthe CRFCs is not adversely affected by the CRFC replacement.

The electrical loading ofthe SW pumps, the CRFC fans and consequently the EDGs are not increased by the CRFC replacement.

The replacement ofthe CRFCs byEWR 5275 does not increase the consequences ofoccurrence ofa malfunction ofequipment important to safety as previously evaluated inthe UFSAR. The functional heat removal and iodine removal capability ofthe CRPCs following design basis accidents has not been adversely affected bythe CRFC equipment. Therefore, the operation of the CRFCs does not impact equipment mal-functions discussed in the UPSAR.

The replacement ofthe CRFCs byEWR 5275 does not create the possibility ofan accident ofa differenttype than those previously evaluated inthe UFSAR. The operation ofthe CRFCs does not initiate any design basis accidents. The replacement CRFCs are similarinfunction, design and operation to the original CRFCs. The change inCRFC coildesign and tube material have resulted in an enhancement in CRFC functional capability when compared to the original CRFCs.

The replacement ofthe CRFCs byEWR 5275 does not create the possibility ofa malfunction of equipment important to safety ofa differen type than those previously evaluated inthe UFSAR.

The operational characteristics ofthe replacement CRPCs is similarto the original CRFCs. Both CRFC designs utilized finned tubing coils to cool containment air.

The basic SW piping configuration to and from the CRFC coils is unchanged as is the air side flow train inside containment. No automatic control features arebeing added to the replacement CRPC coildesign.

The onlychanges to the coils is enhanced tubing materials forcorrosion and erosion concerns and increased heat removal characteristics due to a different tube bundle design.

None ofthe enhancements incorporated into the new CRPCs can cause a new type ofCRFC malfunctionwhen compared to the original CRFCs.

The replacement ofthe CRPCs byEWR 5275 does not decrease the margin ofsafety as defined '

inthe basis forthe Ginna Technical specifications. No changes to the Technical Specifications were identified as a result ofthe CRFC replacement.

The peak fuel cladding temperatures still satisfy the 10CFR50.46 requirement ofnot exceeding 2200'F. OfF-site doses due to a design basis LOCAstillsatisfy the requirements of1OCRF100. Control Room doses due to a design basis LOCAstillsatisfy the requirements ofNRC General Design Criteria 19 related to Control RoomHabitability. Peak calculated containment pressures during design basis pipe ruptures are still below the containment design pressure of60 psig.

SEV-1115 REMOVALOF CRDMGAAND CRDMGB REVERSE POWER PROTECTION UFSAR Section 7.7.1.2.5.1 "Alternating Current Power Connections" takes credit fortrippingout an MGset on a reverse power condition. The Control Rod DriveSystem original design included reverse power protection. This protection was removed (reference TSR 91-167, TM93-031, and EWR 10322) due to several occurrences ofundesired inadvertent trippingofthe MGsets.

However, this TSR, TM,and EWRneglected to adequately document the 10CFR50.59 evaluation ofthe removal and to update the UFSAR.

The probabilityofoccurrence ofan accident previouslyevaluated inthe SARis not increased. The change does not introduce any new failure modes or effects into the Control Rod Drivesystem(not already reviewed in the accident analysis).

The consequences ofan accident previously evaluated inthe SARis notincreased by this change.

The change does not introduce any new failure modes or effects into the Control Rod Drive system.

The probability ofoccurenceofamalfunctionofequipmentimportant to safetypreviously evaluated inthe SARis not increased. The change does not introduceany new failure modes or effects into the Control Rod Drive System.

The consequences ofa malfunction ofequipment important to safety previously evaluated inthe SARis not increased.

The change does not introduce any new failure modes or effects into the Control Rod Drive System.

The possibilityforan accident ofa differen type than any evaluated previously inthe SARis not created.

The proposed change does not create any new equipment interactions.

The possibility ofa malfunction ofequipment important to safety ofa differentype then evaluated previously inthe SARis not created. The proposed change does not create any new equipment interactions.

The margin ofsafety as defined inthe basis forany technical specification is not reduced by this change. Reactivity control orthe abilityto drop the rods into the core(ifrequired) isunaffecte by this change.

SEV-1116 CHANGE MOV7443 ANDMOV7444 FROM MOTOR ACTUATIONTO MANUALACTUATION MOV7443 and MOV7444 are motor operated containment leak test isolation valves. The valves do not require electrical actuation to perform their design function. Due to the increased maintenance associated withmotor operated valves, the added cost ofmaintaining the motor actuators on valves 7443 and 7444 has no benefit foGinna Station and increases the competitive price ofproduct.

Electrical power forMOV7443 and MOV7444 willbe removed perPCR98-012.

The motors willbe abandoned in place.

Electrical cables, conduit and components willbe removed as practicable. Avalvehandle willbeinstalled to allowmanual actuation and ameans forlockingthe handle willbe provided to prevent tampering and/or mispositioning.

The probability ofoccurrence ofan accident previouslyevaluated inthe SARis not increased due to removal ofthe control power and position indication forvalves 7443 and 7444. The valves are not individuallyevaluated inaccidentmitigation. The valves are currently maintained in a closed position above mode 5 withposition indication provided on the MainControl Board. Afterthe modification the valveswil continue to be maintained inthe closed position however thevalves will be locked closed due to the removal ofthe position indication. Plant configuration and piping remains unchanged and the containment integrity boundary is unaffected.

The consequences ofan accident previously evaluated in the SAR willnot be increased due to removal ofthe control power and position indication forvalves 7443 and 7444. The valves are currently maintained in a closed position above mode 5 and this willnot be changed by this modification. New locking valve handles willbe installed to prevent mispositioning.

Plant configuration and piping willnot be changed by this modification therefore the integrity ofthe containment boundary is unafFected.

The probability ofoccurrence ofa malfunction ofequipment important to safety previously evaluated in the SAR willnot be increased due to removal ofthe control power and position indication forvalves 7443 and 7444. The valves willbe placed in a locked closed configuration during operation above mode 5, which is consistent withcurrent operational position. There is no change to the mechanical properties ofthevalves or pipingtherefore no new malfunctions arebeing added to the configuration.

The consequences ofa malfunction ofequipment important to safety previously evaluated inthe SARis not ere'ated due to removal ofthe control power and position indication forvalves 7443 and 7444. Since there is no change inmechanical properties and thevalves willbe maintained in a locked closed position foroperation above mode 5 there are no new malfunctions to consider.

The possibilityofan accident ofa difterent type than any evaluated previously inthe SARis not created due to removal ofthe control power and position indication forvalves 7443 and 7444.

The valves willbeprocedurally maintained ina locked closed position foroperation above mode 5, the same position which the valves are currently positioned.

The possibilityofa malfunction ofequipment important to safety ofadi6erent type then evaluated previously inthe SARis not created due to removal ofthe control power and position indication forvalves 7443 and 7444. The valves willbe placed in a locked closed position for operation above mode 5. New valve handles willbe installed which have been evaluated by Mechanical Engineering withthe determination that the additional weight(approximately three pounds) is negligible and willnot cause a component failure during an earthquake. Inaddition, the handle will be welded to the stem therefore no seismic interactions exist.

The margin ofsafety as defined inthe basis forany technical specification is not reduced due to removal ofthe control power and position indication forvalves 7443 and 7444. Previous technical specification requirements which applied to the valves were per surveillance requirement SR 3.6.3.6 which required verification ofproper actuation ofthe automatic containment isolation function inthe control circuitry. This functionwillbe removed and the valves willbemaintained in a locked closed position above mode 5.

SEV-1117 INSTALLATIONOF SPENT FUEL STORAGE RACKS ANDRELATEDMODIFICATIONSTO SPENT FUEL POOL The proposed changes to the facility are as follows:

P (a)

RemovethreeoldrackswithnoneutronabsorberthatcurrentlyconstituteRegion 1 ofthe spent fuel pool.

(b)

Install seven new racks having Borated Stainless Steel as a neutron absorber.

Two ofthe racks willbe assigned to increase the capacity ofRegion 2 and the remaining five racks willbe designated as the new Region 1.

(c)

Remove obstructionsasneeded.

Obstructions currentlyidentified forremova are as follows:

Four lightfunnels attached to the liner(twolocated on the northwall; one located on the east wall, and one located on the south wall). These light funnels willbe shortened to approximately 1/4 in.

Stubs attached to the liner (several stubs are located on the north, east and south walls ofthe spent fuel pool). These stubs willbe shortened to approximately 1/4 in.

On removal ofthe old racks, other obstructions maybeidentifie. These potential obstructions willbe removed using the same procedures, tools, and administrative controls that areutilized to remove the above obstructions. Thiswillensure that the probability ofpuncturing the spent fuel pool liner is as low as reasonably achievable. In the event ofa puncturing,ofthe liner, there are procedures and administrative controls necessary to promptly inspect and repair any potential leaks.

(d)

Install metal strips withasetof letter/number coordinates called "X-YIndexing" on the edge ofthe pool to aid positioning ofthe spent fuel bridge during fuel shuBling. The X-YIndexing plates willbebolted to the top ofthe concrete wall surrounding the spent fuelpool, on the north and south sides. The area at the top ofthe wallis that between the rail and the liner. Implementation guidelines will ensure that no rebar is cut. There willbe tack welds applied on the outer edges ofthe bolts and the X-Yindexing.

(e)

Relocate the support for the spent fuel handling tool further along theeastwall to a position closer to the south wall. This modification willentail removing the existing support forthe spent fuel assembly handling toolthat is welded to the liner and installing a new support that consists ofa horizontal plate supported by a bracket over the curb. The horizontal plate that supports the spent fuel handling tool is identical to the existing one. The bracket willhave a bolt on the outer side ofthe curb.

The scope of this safety evaluation is to primarily address any of the possible temporary configurations ofthe racks during the installation (a temporary configuration is defined as the geometrical arrangement ofany number ofracks on the pool floorthat is different fromthe final layout achieved after the end ofthe installation).

Ingeneral, temporary configurations are not explicitlydescribed intheNRC Safety Evaluation (NRC SE) issued bythe U S. NRC toRG&Eon July 30, 1998. The NRC SE addresses the final configuration an establishes safety requirements applicable during the installation(e.g. criteria for heavy loads, criticality,radiological, summary ofocc'upational exposure during the installation).

This safety evaluation willprovidethe basis fordetermining that the conclusions intheNRC SE are bounding withrespect to any ofthe possible temporaiy configurations that coul develop during the installation, and willalso provide the basis that there are no additional unreviewed safety questions by implementing the modifications described above.

Removal ofOld Racks and Installation ofNew Racks:

TheNRC SE documents the evaluation ofdesign basis accidents applicable during and after the installation. Training prior to the installation, adherence to procedures, and administrative controls willensure that the probability ofoccurrence ofthe applicable design basis accidents, including drop ofheavy loads, willnot increase. The probability ofoccurrence ofany ofthe designbasis accidents already documented inthe SAR and the NRC SE has not been increased.

This evaluation provides thebasis fordetermining that the consequences ofthe designbasis accidents documented in the NRC SE are bounding withrespect to any ofthe possible temporary configurations that could develop during removal of the old racks and installation of the new racks.

All limits and requirements will be met during the modification. The consequences ofaccidents previously evaluated inthe SAR and the NRC SE have not been afFected.

TheNRC SE outlines therequirements formovements ofheavy loads during and after the installation. Theserequirements willbe met during the installation. There is no impact on the malfunction ofequipmentimportant to safety. Therefore, the probability ofoccurrence ofamalfunctionofequipmentimportant to safety previously evaluated inthe SARremains unchanged.

TheNRC SE outlines therequirements formovements ofheavyloads during and after the installation. Theserequirements willbemet during the installation. There is no impact on the malfunction ofequipment important to safety. Therefore, the consequences ofa malfunction ofequipment important to safety previously evaluated inthe SAR remain unchanged.

The NRC SE documents the evaluation ofdesign basis accidents applicable during and after the installation. This evaluation providesthebasis fordetermining that the design basis accidents documented intheNRC SEarebounding and stillapplicable withrespect to any ofthe possible temporaryconfiguration that could develop during removal ofthe old racks and installation ofthe new racks.

There are no new accidents introduced during the modification. Therefore, the possibilityofanaccident ofa differen type than any evaluated previously in the SAR and in the NRC SE is not created.

TheNRC SE outlines therequirements formovements ofheavy loads during and after the installation. These requirements willbemet during the installation. Equipment important to safety willnotbephysically affecte by removal ofthe old racks and installation ofthe newracks. There is no impact on the malfunction ofequipment important to safety during the modification. Therefore, the possibility, ofa malfunction ofequipment important to safety ofa different type than evaluated previously in the SAR is not created.

TheNRC SE documents the evaluation ofdesign basis accidents applicable during and afier theinstallation. This evaluationprovides thebasis fordetermining that the evaluation ofthe basis accidents documented in the NRC SE is bounding and stillapplicable with respect to any ofthe possibletemporary configurationstht could develop during removal ofthe old racks'and installation ofthe new racks. Allregulatory requirements and limits set forth inthe SAR, the NRC SE, and the Technical Specifications are met during the modification. Therefo're, the margin ofsafety as defined in the basis for any technical specification is not reduced.

Removal ofObstructions:

Training, procedures, and administrative controls are established to ensure that the probability ofpuncturing the spent fuel pool lineris as lowas reasonably achievable. The probability ofoccurrence ofa breach ofthe liner resulting in a damage similar to that a tornado missile puncturing the liner documented in the UFSAR has not increased.

Training, procedures, and administrative controls are established to ensure that the

probabilityofpuncturing the spent fuel pool lineris as lowas reasonablyachievable.

The consequences ofany potential breach ofthe liner during removal ofobstructions are bounded by the consequences ofa hypothetical tornado missile puncturing the liner as documented inthe tornado missile designbasis accident. The consequences ofaccidents previously evaluated in the SAR have not been affected.

Maintaining the structural integrity ofthe spent fuel pool liner does not impact the malfunction ofequipment related to safety. Therefore, the probabilityofoccurrence ofa malfunction ofequipment important to safety previously evaluated inthe SAR remains unchanged.

Maintaining the structural integrity ofthe spent fuel pool liner does not impact the malfunction ofequipment related to safety. Therefore, the consequences ofa malfunction ofequipment important to safety previously evaluated inthe SAR remain unchanged.

Anypotential breach ofthe spent fuelpool linerduring removal ofobstructions is bounded bythe consequences ofa hypothetical tornado missile puncturing the lineras documented in the tornado missile design basis accident.

The proposed modification does not introduce anew failure mode not documented inthe SAR. Therefore, the possibility ofan accident ofa di6erent type than any evaluated previously in the SAR is not created.

Maintaining the structural integrity ofthe spent fuel pool liner does not impact the malfunction ofequipment related to safety. Therefore, the possibility ofamalfunction of equipment important to safety of a diferent type than evaluated previously inthe SARis not created.

Anypotential breach ofthe spent fuelpool linerduring removal ofobstructions is bounded bythe consequences ofahypothetical tornado missile puncturing the liner as documented inthe tornado missile design basis accident. Therefore, the margin ofsafety as defined in the basis for any technical specification is not reduced.

X-YIndexing:

The concrete structure ofthe spent fuel pool willnot be degraded by installing the X-Y Indexing plates. The probabilityofoccurrenceof the design basis accidents documented in the SAR and the NRC SE for the spent fuel pool structure has not increased.

The concrete structure ofthe spent fuel pool willnot be degraded by installing the X-Y Indexing plates. The consequences ofaccidents previously evaluated inthe SAR and the NRC SE have not been affected.

The concrete structure ofthe spent fuel pool willnot be degraded by installing the X-Y Indexing plates. There is no impact on the malfunction ofequipmentimportant to safety.

Therefore, the probability ofoccurrence ofamalfunction ofequipment important to safety previously evaluated in the SAR remains unchanged.

The concrete structure ofthe spent fuel pool willnot be degraded by installing the X-Y Indexing plates. There is no impact on the malfunction ofequipment important to safety.

Therefore, the consequences ofa malfunction ofequipmentimportant to safety previously evaluated in the SAR remain unchanged.

The X-YIndexingplates are an attachment to the spent fuel pool structure. The spent fuel pool structure has been evaluated under normal and abnormal conditions as documented in the SAR. The proposed modification does not introduce a new failure mode not analyzed inthe SAR. Therefore, the possibility ofan accident ofa different type than any evaluated previously in the SAR is not created.

The concrete structure ofthe spent fuel pool willnot be degraded by installing the X-Y Indexing plates. There is no impact on the malfunction ofequipment important to safety.

Therefore, the possibilityofamalfunction ofequipment important to safety ofa different type than evaluated previously in the SAR is not created.

The concrete structure ofthe spent fuel pool willnot be degraded byinstalling the X-Y Indexing plates. Therefore, the margin ofsafety as defined inthebasis forany technical specification is not reduced.

Relocation ofthe Support for the Spent Fuel Handling Tool:

The design ofthe proposed support is similarto the existing one. The probabilityofa drop ofthe spent fuel handling tool on the racks has remained unchanged. Inthe conservative direction, the tool support has been positioned further away from spent fuel racks. The probability ofoccurrence ofthe design basis accidents documented inthe SAR and the NRC SE for the spent fuel pool structure has not increased.

The consequences ofa drop ofthe spent fuel handling tool on top ofspent fuel racks are bounded bythe consequences oftheFuel Handling Accident(FHA) documented inthe NRC SE.

The consequences ofa drop ofthe tool support on top ofthe spent fuel racks are bounded by the consequences ofthe Tornado Missile Accident documented in the NRC SE.

The consequences of accidents previously evaluated in the NRC SE have not been 0

afFected.

The drop ofthe spent fuelhandlingtool and/or its support has no impact on the malfunction ofequipment important to safety. Therefore, the probabilityofoccurrence ofamalfunction ofequipment important to safety previously evaluated inthe SAR remains unchanged.

The drop ofthe spent fuel handling tool and/or its support has no impact on the malfunction ofequipment important to safety.

Therefore, the consequences ofa malfunction of equipment important to safety previously evaluated in the SAR remain unchanged.

Identified accidents are the drop ofthe spent fuel handling tool and/or its support inthe spent fuel pool. These accidents are bounded by accidents documented inthe NRC SE.

Theproposed modification does not introduce anew failuremode not analyzed inthe SAR and the NRC SE. Therefore, the possibility ofan accident ofa different type than any evaluated previously in the SAR is not created.

The drop ofthe spent fuel handling tool and/or its support has no impact on the malfunction ofequipment important to safety. Therefore, the possibilityofamalfunction ofequipment important to safety ofadifFerent type than evaluated previously inthe SARis not created.

The consequences ofa drop ofthe spent fuelhandling tool on top ofthe spent fuel racks are bounded by the consequences ofthe Fuel Handling Accident(FHA) documented in the NRC SE. The consequences ofa drop ofthe tool support on top ofthe racks are bounded bythe consequences ofthe Tornado MissileAccident documented intheNRC SE. Therefore, the margin ofsafety as defined inthe basis forany technical specification is not reduced.

SEV-1118 SEVERE ACCIDENTMANAGEMENTGUIDANCE SAMG IMPLEMENTATION The purpose ofthis change is to implement the Severe Accident Management Guidance(SAMG) at Ginna Station. The SAMGs are designed foruse inextreme accident circumstances when the Plant EOPs are no longer effective an core damage is progressing. The new guidelines address the accident management changes necessaiy to mitigatethe consequences ofa severe accident that have progressed beyond the plant's design basis.

Therefore, the actual SAMGs are likewise considered to bebeyond designbasis documents and are not subject to 10CPR50.59 review. The procedures addressed by this safety evaluation outline the administrative guidance for implementation and maintenance ofthe SAMGprogram, as well as the EOP transitions to the SAMGs.

The SAMGs are not put into use untilan accident has progressed beyond the design basis ofthe plant. Because the SAMGs do not direct any plant alterations until after the normal accident mitigation procedures (EOPs) are exhausted, the probability ofthe occurrence ofa previously evaluated accident is not increased.

Because the SAMGs are actually designed to minimize the consequences ofan accident that has progressed beyond the design basis after the mitigation efforts directed by the'EOPs have been exhausted, the consequences of a previously evaluated accident willnot be increased.

The change addressed by this review simply establishes the administrative aspects ofthe SAMG program. The equipment configuratio, functions or methods ofperforming those functions as described in the UFSAR are not affected.

The consequences ofpreviously evaluated equipment failures arenot affected by the admiistrative aspects ofthe SAMGprogrambecause equipment operation or configuration is not addressed in these documents.

The purpose ofthe SAMG program is minimization ofthe public dose consequences from accidents that haveprogressed beyond the plant's design basis. Because the SAMGprocess does not change any normal, off norma or design basis event mitigation equipment configuration or fundamental interactions, the use ofthe process cannot lead to a previously unevaluated accident.

Ifapreviously unevaluated accident should occur, the SAMGs should provide some guidance in dealing withthe situation and thereby providing the plant staff wita toolto perform theirprimary function ofprotecting the public.

The administrative aspects ofthe'SAMG program does not deal with equipment operation, configuration or functionalityissues. Because the SAMGprogram does not result in any equipment design function changes, the possibilityofanunevaluated equipment failure is notincreased.

The SAMG's do provide equipment lineups and operational suggestions but onlyafter the design basis accident mitigation procedure set is determined to be ineffective. The SAMGs are considered beyond design basis documents and, as such, willbe maintained as guidelines and not subject to 50.59 review.

The implementation ofthe SAMGs is an industry commitment to theNRC and beyond the scope of Tech Specs.

As SAMGs deal with beyond design basis events, Tech Spec bases is not affected.

SEV-1119 EVALUATIONOF ADJUSTABLETRAVELSTOP SET POSITION FOR HCV-624 ANDHCV-625 Anadjustable valve travel stop willbe added to the actuators forRHR discharge control valves HCV-624 and HCV-625. The travel stop consists of atop mounted handwheel, mounted onto the existing actuator top cover. The handwheel has the capability to either manually close the valve orbe used as a limitto upward travel ofthe actuator, thereby limitingthe open position ofthevalve.

The handwheel willbeinstalled under PCR98-068, and the desired position ofthe valve set during aflowtest planned as part ofPT-2.10.10, during theinitial stage ofthe refueling cavityfillinduring the 1999 outage. Followingthe setting ofeach valve actuator inposition, the handwheel willbe chain locked in place. The modification willnot prevent the valves frombeing throttled in the closed direction. The valves'pen position willbe limited to a position less than fullopen as determined by analysis and be set during the flowtest. No further adjustment ofthe valve is needed for any mode ofoperation.

The probability ofoccurrence ofan accident previously evaluated inthe SAR is not increased, because the affected valves, HCV-624/625, do not change position following a postulated transient. They remain intheir open position. The extent oftheir open position is being changed, and the new position willstillensure the required lowhead safety injection flowforthe duration of the transient.

The consequences ofan accident previously evaluated inthe SAR are not increased, because the required flowrate listed inthe COLRwillcontinue to be maintained during the injection phase.

Providing alimiton system flow wilalso ensure, under conditions resulting inmaximuminjection flow,that RHRpump runout conditions do not exist. Inthe longer term, folio'wingswitchover to the sump recirculation phase, the modification provides a limitation on RHR flowrate, while assumingalossofinstrumentair,therebypreservingNPSHmargin.

Therefore,corecoolingcan continue with no loss offunction.

The probability ofoccurrence ofa malfunction ofequipment important to safety previously evaluated inthe SARis not increased, because the valve travel stop is a physical stop againstwhich the actuator stem rests. The valve actuator is not called upon to move followinga postulated accident so there is no increase in probability ofa malfunction. Should a loss ofinstrument air occur, the travel stop willprevent movement ofthe valve, since the stem ofhandwheel assembly rests against the diaphragm preventing further opening.

During non-accident modes ofthe RHR system when throttlingis necessary using HCV-624/625, the travel stop willnot interfere withthe throttlingofthe valves inthe closed direction. There is currently no need to throttle the valves more open than the travel stop position willbe set.

0

Administrative limitscurrently exist onRHR flo(1500 gpm) whichlimitthe lowrate to a value less than the travel stop would allow.

The consequences ofa malfunction ofequipment important to safety previously evaluated inthe SARis not increased, because the valves'unction willremain failopen on loss ofinstrument air.

Since the valves are normally maintained open whilethe plant is at power, there are no times when the failopen on loss ofinstrument airfunction would be called upon. There are no malfunctions that would cause the valves to failclosed since the spring inthe actuator is a passive device not dependent on external controls, and the valves are routinelytested and calibrated. The travel stop cannot cause the valves to move inthe closed direction, since its design onlyrestrictsmotion inthe upward direction. Therefore, LHSIflowrate willstillmeet the COLRvalues and no increase in consequences can occur due to reduced core cooling assumed in the accident analysis.

The possibilityofan accident ofa different type than any evaluated previously inthe SARis not created, because the travel stop does not interfere withthe operation ofHCV-624/625 over the range oftravel these valves are assumed to maintain.

The possibilityofamalfunction ofequipment important to safety ofa different type than evaluated previously inthe SARis not created, because the travel stop is designed to provide suflicientflow to preserve LHSIcapabilityunder thelimitingassumptions previously assumed, whilelimitingflow su6iciently to preserveRHRpump NPSHmarginduring the sump recirculation phase. The valve actuator handwheel willbelocked inplace so that manually opening the valvemore thanits setpoint cannot be inadvertently performed..Operationofthevalve fiomthe control room and operators use ofthe valves willbe unaffected.

The margin ofsafetyas define inthe basis forany technical specification is not reduced, because no changes are being made to the functions ofthe valves, and the LHSI system capability will continue to be maintained in excess ofCOLR flowrequirements in the limiting case.

SEV-1120 REMOVALOF DEWPOINT MEASURINGINSTRUMENTATIONFROM THE SEISMIC AND METEOROLOGICALINSTRUMENTATIONSYSTEM The Ginna Station Seismic and Meteorological Instrumentation System (SMI)is made up ofa variety ofcomponents.

Included is a dewpoint measurement system. The Instrumentation Ec Control Special Projects group has requested to remove the dewpoint measuring system because of the maintenance requirements of the system and lack of requirements for its use.

The environment inwhich the dewpoint transmitter is required to operate (increased frequency of airborne dirt particles due to a fairlyconstant breeze) is not conducive to the sensitivity ofthe dewpoint transmitter. The dewpoint transmitter senses humidityvia a lens which is frequently fouled with dirt and grime resulting in recurring problems and inaccurate data.

The dewpoint monitoring system does not interactwith any equipmentused to mitigate accidents or transients. Inaddition, the data gathered bythe dewpoint monitoring system is not used inthe decision process for mitigation ofaccidents or transients.

The dewpoint monitoring system is functionallyunrelated and physically independent ofany System, Structure or Component important to safety. Theindependence ofthe dewpoint measuring system from any System, Structure or Component important to safety ensures that the proposed modification can not introduce a failure mechanism which would increase the probability of occurrence ofan accident previously evaluated in chapter 15 ofthe UFSAR. The dewpoint monitoring system is not required per Reg. Guide 1.97. This modification willnot affect the meteorological monitoring system design limits nor reduce system reliability.

The dewpoint monitoring system is functionally unrelated and physically independent ofall equipment used forthe mitigationofaccidents and transients. The independence ofthe dewpoint measuring system from any System, Structure or Component important to the mitigation of accidents and transients ensures that the proposed modification can not introduce a failure mechanism whichwould increase the consequences ofan accident previouslyevaluated inchapter 15 ofthe UFSAR. The modification does not impact or increase the calculated radiological dose to the general publicforany event evaluated intheUFSAR. The dewpoint monitoring system is not presently required per Reg. Guide 1.97 and is not used as an input to other dose calculations.

The dewpointmonitoring systemis functionallyunrelated and physically independent ofany System, Structure or Component important to safety. Theindependence ofthedewpointmeasuring system from any System, Structure or Component important to safety ensures that the proposed modification can not introduce a failure mechanism which would increase the probability of occurrence ofa malfunction ofequipment important to safety previously evaluated inchapter 15 ofthe UFSAR. The modification willnot degrade the performance ofthe meteorological monitoring system. The dewpoint monitoring systemisnotinterconnected to any System, Structure

or Component important to safety.

The dewpoint monitoring system is functionally unrelated and physically independent ofall equipment used forthe mitigationofaccidents and transients. The independence ofthe dewpoint measuring system fiom any System, Structure or Component important to the mitigation of accidents and transients ensures the proposed modification wilnot introduce a failuremechanism which would increase the consequences ofa malfunction of equipment important to safety previously evaluated inchapter 15 oftheUFSAR. Themodification does not impact or increase the calculated radiological dose to the general public for any event evaluated in the UFSAR.

The dewpoint monitoring system is functionally unrelated and physically independent of all equipment used forthe mitigationofaccidents and transients. The independence ofthe dewpoint measuring system from any System, Structure or Component important to the mitigation of accidents and transients ensures that the proposed modification willnot introduce a failure mechanism which would increase the probability ofan accident of a diFerent type than any previously evaluated in chapter 15 ofthe UFSAR. There are no adverse affects upon other systems, nor any new failure modes induced.

The dewpoint monitoring system is functionally unrelated and physically independent ofall equipment used forthe mitigation ofaccidents and transients. The independence ofthe dewpoint measuring system from any System, Structure or Component important to the mitigation of accidents and transients ensures the proposed modification will notintroduc a failure mechanism whichwould increase the consequences ofa malfunction ofequipment important to safety ofa different type than previouslyevaluated inthe UFSAR. The power source forthe Ginna Station Seismic and Meteorological Instrumentation System is &omboth of-'site power and in-plantnon-1E sources.

The physical location is such that damage to the structure(s) itselfwillnot afFect equipment important to safety. The modification does notdegrade the meteorolgical monitoring system.

The dewpoint monitoring system is functionally unrelated and physically independent ofall equipmentused forthe mitigation ofaccidents and transients. The independence ofthe dewpoint measuring system from any System, Structure or Component important to the mitigation of accidents and transients ensures the proposed modification will not intoduce a failure mechanism whichwould reduce any margin ofsafety as defined inthe basis ofany Technical Specifications.

The required functions and characteristics ofthe Ginna Station Seismic and Meteorological.

Instrumentation System remain unchanged.

SEV-1121 PCN 8 98-4517 SAFETY EVALUATIONCHANGES TO ATT-2.1 ATTACHMENTMINSW TO ADDRESS ACTIONREPORT 98-1042 CONCERNS ACTIONReport98-1042 identified a concern withguidance provided inRevision 4 ofprocedure ATT-2.1,"ATTACHMENTMNSW". This attachment is used to align the service water system for the recirculation phase ofa LOCAwith one operable SW pump. ATT-2.1 instructs the operators to fullyopen the service water globe valve on the discharge side ofthe CCW heat exchanger to be aligned (V-4619 or V-4620).

During a reconstitution ofthe service water system hydraulic model, an error was found inthe hydraulic loss coefficient used to represent each CCWheat exchanger. The coefBcientused inthe calculationwas significantlyhigher thanthevalue that would beback-calculated fiomeither vendor supplied pressure drop data or actual test data.

Since the actual hydraulic resistance is lower than originallymodeled, the servicewater flowrate to the applicable CCW heat exchanger would be considerably higher than originallypredicted if V-4619 or V-4620 were opened completely witha single service pump in service. As a result, the flowrate to the CRFCs and EDG coolers could be significantly lower than predicted in previous hydraulic models, and the service water pump margin to runout would be reduced.

The following changes to Revision 4 ofATT-2.1 are proposed:

Delete step 3 which has operations request that the TSC evaluate isolation ofSW loads incontainment. The step willbe replaced withexplicitinstructions to isolate inoperable containment loads (CRFCs and Reactor Compartment Coolers) by closing the service water isolation valve on the discharge ofeach line. This step willbe preceded by a note stating that these isolations are to be performed as soon as possible after sump recirculation has been established.

Add anew step containing the guidancepreviously in step 3 regarding TSC evaluation of closure ofthe Bus 17-18 cross-tie and startup ofa second service water pump. This change is considered inconsequential and willnot be addressed inthis safety evaluation.

Breakout the step that isolates service waterto the SFP heat exchangers and place itprior to the step that adjusts service water flowto the applicable CCW heat exchanger. There is no reason that this step has to be done after restoring service water to the Auxiliary Building, and moving this step minimizes complications during alignment ofservice water to the applicable CCW heat exchanger, such as the effect ofservice water flowfrom SFP HXBon the FIA-2005 reading, which is used to set V-4619 orV-4620 position. Since the current attachment revision already isolates SFP cooling, and the attachment must be

completed in entirety prior to going into recirculation, relocation ofthis step has no implications on the timingofthe transfer to recirculation. Therefore, relocation ofthis step is considered inconsequential and willnot be addres'sed in this safety evaluation.

Modifystep 5 to throttle the SW outlet valve on the operating AuxiliaryBuildingservice water loop to between 2750 gpm and 3250 gpm.

Add a note to inform operators that EDG cooling may be aFected while adjusting the service water flowtothe CCW heat exchanger and to reduce load orrefer to ER-D/G.2, ALTERNATECOOLINGFOR EMERGENCYD/Gs, should an EDG temperature alarm occur. This change is considered inconsequential and willnot be addressed inthis safety evaluation.

Add new step to notifyTSC ofall loads that were isolated. This change is considered inconsequential and willnot be addressed in this safety evaluation.

Implementation ofthese steps willaddress the issue raised inACTIONReport 98-1042.

Additionally,explicitisolation ofinoperable containmentbuilding loads willincrease the heat removal rates fromcontainment, increase the margin to vapor lockinginthe CRFCs due to flashing inthe downstream service water piping, and provide increased service water flowrates to the EDG coolers.

TheproposedprocedurechangesapplyduringtherecirculationphaseofaLOCAonly.

Theywill not increase the probability ofoccurrenceofan accident previously evaluate'd inthe SARsince the accident willhave already occurred prior to usage ofthe procedure.

The proposed procedure changes willnot increase the consequences ofan accident previously evaluated in the SAR. The analyses outlined in the functional impact section ofthis safety evaluation provide adequatejustification thatthe changes have ony positive sects withrespect to the capability to deal with and the consequences ofa LOCA, which is the only impacted accident.

Theproposed changes arebeneficial withrespec to equipment reliabilityduring the recirculation phase ofaLOCAand therefore itis reasonable to conclude that the probabilityofoccurrence of a malfunction ofequipment important to safety is reduced.

Specifically:

The margin to overheating ofthe Emergency Diesel Generators is increased due to.the increase in service water flowto the EDG coolers.

The margin to vapor lockinginthe CRFCs is increased due to the increase inservice water flowthrough the CRFC.coolers.

This is also beneficial with respect to cooling the

atmosphere in containment.

The margin to runout ofthe. single operating service water pump is increased by the increase in system back pressure.

The EQ temperature and pre'ssure profiles which were used to qualify equipment in containment forpost-accident conditions are met. Isolation ofinactive containment loads increases overall containmentheatremoval so is marginallybeneficialin term ofequipment reliability.

The consequences ofa malfunction ofequipment important to safety are not increased by the proposed changes.

This statement is justified in the sub-sections below:

Note that the licensing basis single-active failurewillalready have occurred priorto usage ofATT-2.1, since this is required to get to one service water pump operation; therefore, any additional failure willbe beyond the design basis ofthe plant.

The consequences ofany failure that results in the loss ofthe sump heat sink willbe reduced. Loss ofthis heat sinkwillresult inan increase incontainment temperature; the increased flowtothe CRFCs provided by the proposed changes beneficiallyincreases th heat removal fiomthe containment atmosphere and increases the margin to vapor locking in the CRPCs.

The consequences ofthe loss ofa CRFC are reduced since the remaining CRFC(s) will have higher flow rates and therefore greater heat removal.

The consequences ofthe loss ofan EDG, assuming only one was originallyoperating, are no more severe than they would be ifthe proposed changes were not implemented since the end result in either case is a complete loss ofactive heat sinks.

The consequences ofthe loss ofthe operating service water pump are no more severe than they wouldbeifthe proposed changes were not implemented sincethe end result ineither case is a complete loss ofactive heat sinks.

The possibilityofan accident ofa different type than any evaluated previously inthe SARis not created. The proposed changes are intended to help mitigate the consequences ofan accident that has already occurred, and a second accident is not assumed to occur coincidentally during recovery from the first.

The proposed changes do not change the configuration ofthe plant priorto the occurrence ofa design-basis LOCAand therefore willnot create the possibility ofadifferent type ofmalfunction.

As discussed previously, the proposed changes ultimately havebeneficia impacts on equipment reliability during the recirculation phase ofthe LOCA.

The onlymargins that could potentiallybe challengedby the changes are themaximum containment pressure and the EQ profiles. The proposed changes willhave no impact on the peak pressure since this occurs priorto the transfer to recirculation. Further, ithas been shown that the proposed changes do not challenge the profiles assumed forequipment qualification. Since no margins of safety have been challenged, the margin of safety as defined in the basis for any technical specification willnot be challenged.

C

SEV-1123 SPENT FUEL PIT LEAKAGERELEASE PATH ASSESSMENT There havebeen numerous USNRCInspectionReports dealing withthe presenceofwater leakage into various plant structures. Analyses ofsome ofthe leakage has indicated the presence ofboric acid and radionuclides that are also present inthe spent fuel pool(SPF) and transfer canal. With this finding,theNRC has expressed a concern on the potential fora radionuclide release ofF-site.

USNRC Inspection Report 95-015-01 initiated the concern ofa radiological release ofthe Spent Fuel Pit (SPF) water into the environment.

Since that inspection, several measures have been initiated to (1) assess the leakage source, (2) determine the most probable groundwater flow direction, and (3) initiate a monitoring program for tracking any potential offsite releases.

Based on sampling and testing, ithas been determined that some leakage is occurring fromthe transfer canal.

This evaluation is to assess the potential forsuch a release and demonstrate that leakage fromthe transfer canal willbe controlled and processed as required to confoimto the appropriateNRC and EPA regulations.

Asudden increase in SFP Liner leakage would be the accident/event ofconcern which, is not presently addressed inthe UFSAR accident analysis. Therefore, the probabilityofoccurrence of an accident previously evaluated in the UFSAR is not increased.

As stated above, the accident in question is not evaluated in the UFSAR. Therefore, the consequences ofan accident previously evaluated in the SA is not increased.

The equipment inquestion would be the RHR and RCDT Pumps inthe AuxiliaryBuildingsub Basement. The suspected leakage ofSFP water intothe RHRroom is believed tobe originating through incomplete or defective seal welds ofthe linerto the embedment ofthe refueling canal.

Should there be a complete failure ofthesewelds, anunrestricted flowofthe canal inventory into the RHR room is precluded by the concrete/bedrock interface.

In addition, the increased frequency ofAuxiliaryBuildingsump Pump actuations would alert the Operators, providing an opportunity to take corrective actions.

Therefore, there is no increase in the probability of occurrence ofa malfunction ofequipment previously evaluated in the SA.

The consequences ofthe event described above, (failureofthe RHR/RCDTPumps) would remain the same regardless ofthe failure mechanism and therefore, would not be increased.

The flowpath described above, does not lend itselfto a rapid outflowofwater &omthe SFP. The leakage ofborated water into the RHRroom has beendetermined to be fromthe refueling canal and has been quantified to be very small (&.001 gaVmin). The leak path is through the interface ofthe refueling canal concrete foundation and bedrock. Both the concrete foundation and bedrock are impervious to water and, as such, erosion/failure ofeither, which could establish a potential flood path is not possible. This restrictionofflowwouldallowample time formitigating actions such as installing the weirgate, closing ofthe transfer tube gate valve and/or draining the transfer canal. Even inthe unlikelyevent ofa rapid outflow ofwate from a failure inthis area, the height ofthe weir gate path would preclude the uncovering ofspent fuel in the pit. Therefore, the possibility ofan accident ofa different type than that evaluated in the UFSAR is not created.

Based on the above discussions ofleak rates the operability ofthe RHR/RCDT pumps is not jeopardized bythis condition. Therefore, the possibilityofa malfunction ofequipment important to safety ofa different type than previously evaluated inthe SA is not created by this condition.

The issue ofSFP leakage is not addressed in any technical specification, therefore, there is no affect on margins ofsafety as defined in the bases ofthe Technical Specifications.

SEV-1124 VALVESTEM PACKINGIMPROVEMENTPROGRAM SAFETYEVALUATION CHANGES TO DESIGN CRITERIA-EWR 4859 This SafetyEvaluationwasprepared to replaceRevision1 to the Safety Analysis (Revision 1 was never approved) forEWR 4859 to evaluate the addition ofExpandable Valve Stem Packing (EVSP) ofthe "cup and cone" design as a packing system alternative.

This analysis covers the live-loading ofgland followers and/or replacement ofvalve stempacking ofcertain selected valves.

Valve stem packing leakage is a widespread problem that impacts overall nuclear power plant operation and maintenance.

In some cases, even minor stem packing leakage has far reaching implications interms ofradiation exposure, load reduction and housekeeping problems. In 1984, ElectricPowerResearchInstitute(EPRI) established a program to study the root causes ofvalve stem packing leakage and to identify, develop and evaluate means ofcorrective action. As a result, two improvements wereidentifie by recent EPM studies. These improvements, when retrofitte, have the potential to greatly alleviate the maintenance burden associated withthe valve steam packing leakage. These improvements are:

Replacement oftraditional woven asbestos packing with die-formed square flexiblegraphite packing orExpandable Valve Stem Packing(EVSP) ofthe "cup and cone" design.

Live-loading of gland followers to compensate for stress relaxation, aging, consolidation or thermal cycling ofthe packing material.

As part ofthe preventive maintenance program, Ginna Station Maintenance Department has decided to replace the asbestos stem packing withdie-formed square graphite stem packing or EVSP forseveral existing and new valves. Some ofthe valves shall also be retrofitted withlive-loading.

The proposed modification would not increase the probability ofoccurrence ofan accident previously evaluated in the UFSAR since this change only allows replacing approved packing materials and methods withimproved alternatives thatwillreducethe potential forpacking leakage.

The proposed modification would not increase the consequences of an accident previously evaluated intheUFSAR since the expandable valve packing is an improvement invalve packing systems with less potential and, subsequently, less consequences for leakage.

The proposed modification would not increase theprobability ofoccurrence ofamalfunctionof equipment important to safety previously evaluatedintheUFSARbecausethis program provides the criteriaforthereplacement and upgrade ofpacking materials and methodologyin safety-related valves resulting in an increase in reliabilityfor affected valve operation.

The proposed modificationwould not increase the consequences ofa malfunction ofequipment important to safety previously evaluated intheUFSARbecause the Mure ofpacking(existing or replaced by this program) would not violate the equipment's pressure boundary function.

The proposed modificationwould not create the possibilityofan accident ofa different type than any previously evaluated in the UFSAR because the potential for failure related to packing materials and methods currently exist inthelicensing basis and willremain with, althoughmitigated by, new improved packing systems.

The proposed modificationwould not create the possibilityofa different type ofmalfunction of equipment important to safety than any previously evaluated inthe UFSARbecause the change in packing material and methodology incorporates improved technology which willresult in a greater valve packing system reliability.

The proposed modificationwould not reduce any margin ofsafety as defined inthe basis ofany technical specification becaus equipinent reliabilitywillbeincreased upon modificationby the Valve Stem Packing Improvement Program.

~ '

SEV-1125

'TATIONARYBATTERYREPLACEMENT Battery A(BTRYA),Battery B (BTRYB) and the spare battery cells (BTRYSP) are being replaced during the 1999 refueling outage due to aging concerns initiallyidentified inACTION

'eport 97-1110. The batteries have not degraded to the point where discharge testing indicates replacement is required, however the physical signs ofaging, plus the need to replace the cells prior to 2009 have been factored in the decision to replace both batteries at this time.

ElectricalEngineering Specification EE-168 was prepared to outline the design and performance requirements forthe new battery cells. Nuclear Logistics Incorporated willbe providing new batteries manufactured by GNB Technologies meeting the design and performancerequirements ofEE-168. RGB requested quotes for 1200 amp-hour and 1495 amp-hour battery capacity in order to determine the marginal cost ofincreasing themargin between battery capacity and design basis load. The 1495 amp-hour battery was chosen as the replacement.

The existing batteries are GNB model NAX-1200 and NAX-17 (1200 amp-hour). The cells dimensions are: length 7.38 inches, width 14.5 inches and height 22.13 inches. Weight is 245 pounds.

The new batteries willbe GNB modelNCN-21 (1495 amp-hour). These cells are larger than the existing cells. Length 9.25 inches, width 14.5 inches and height 22.5 inches. Weight is 301 pounds which is 56 pounds heavier than the existing cells.

The added size and weight ofthenewbattery cellswould require modification ofthe existing racks.

Aninitialevaluation determined no costbenefitbetweenmodification orreplacent therefore new structural racks willbe installed, designed to meet the seismic forces ofthe Battery Rooms. The sparebattery cell racks willbe modified as necessary to accommodate the larger cells. The Battery Rooms are located in the basement ofthe Control Building.

The probability ofoccurrence ofan accident previously evaluated inthe SARis not increased by this modification. The stationbatteries areused to mitigate the consequences ofaccidents. They have no failure modes or effects which directly lead to the occurrence ofany accident previously evaluated inthe SAR. AAerthe proposed change is complete the batteries willcontinue to have the independence and separation which they are required to have, therefore there are no new functional interactions which affect the previously evaluated accidents.

The new batteries willbe seismically mounted on new racks and willbe operated in the exact configuration as the existing system. Several changes to Goat voltage and equalize time willbe placed into effect through procedural control, however these changes willnot result in the occurrence ofan accident. Other than the batteries, intercell connectors and racks, no new equipment is being added. No existing equipment needs tobe functionallymodifie as a result of this proposed change.

The consequences of an accident previously evaluated in the SAR is not increased by this modification. The newbatteries have a greater capacity than the existing batteries therefore they have the abilityto mitigate any design basis events whichthe old batteries have been qualified to mitigate. This includes the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> station blackout coping period. The increase in amp-hour capacity is based on a difference in battery design, however these design differences willnot increase the consequences ofany accident previously evaluated.

The probability ofoccurrence ofa malfunction ofequipment important to safety previously evaluated inthe SARis not increased by this modification. The newbatteries arebeing purchased as Class-1E equipment fi ornaqualifi supplier to the design conditions oftheBattery Rooms. The new. battery racks are being purchased fromthe same supplier withseismic qualifications to the requirements ofGinna's Battery Rooms. The batteries willbe bounded by the same maximum voltage(140 VDC)as the existingbatteries, therefore the operability ofall equipment connected to the DC distribution systemswillbemaintainedwiththe newbatteries installed. The floatvoltage willbe set at a new higher value inorder to minimize the amount ofequalize charges which have to be performed on the batteries. There is no increase inthe probability ofa malfunction ofany equipment important to safety connected to the DC distribution systems due to this modification.

The consequences ofa malfunction ofequipment important to safety previously evaluated inthe SAR is not increased by this modification.

There is no increase in the consequences of a malfunction ofany equipment due to this modification. Ifone ofthe batteries were to failthere would be no increase in the existing consequences resulting from the loss ofa battery.

The consequences of a malfunction of apiece ofequipment other than the batteries or racks will notbe increased bythis modificatio. Electrical evaluation including coordination and short circuit protection demonstrate that the protection ofthe electrical system willnot be degraded due to this modification. Anevaluation ofthe hydrogen generation capabilityofthenewbatteries and a change inthe Battery Room combustible load demonstrate that installation ofthe new batteries willnot exceed the abilityoftheHVACsystem to remove hydrogen fromthe Battery Rooms nor willthe amount ofcombustible load increasebeyond themaximum allowable combustible load forthe fire zones in which the batteries are contained.

The possibilityofan accident ofa differen type than any evaluated previously inthe SARis not created bythis modificatio. No new equipment is being added to the DC distribution system due to this modification, therefore there is no potential ofan accident ofa different type than any previously evaluated. The batteries arebeing replaced withnew cells withincreased capacity and new seismic racks are being installed. The new batteries are functionallyequivalent to the existing batteries and no new failure modes willbe introduced due to this modification.

The possibility ofamalfunction ofequipment important to safetyofadifferent type then evaluated previously inthe SARis not created bythis modificatio. This modification will installne Class-1E battery cells and new seismic battery racks to replace existing equipment. The new batteries willbelead calcium, which is differentthan the existingbatteries which are lead antimony. Antimony and calcium are metals added to the grid design to increase strength. There is no possibility ofa failure ofa different type due to differences in battery design than previously evaluated forthe batteries or racks.

Evaluations ofthe electrical system, HVACsystem and fireloading demonstrate that the new batteries willbe capable ofoperation withoutimpacting the operability'ofany systems supporting the Battery Rooms or the DC distribution system.

The margin ofsafety as defined inthe basis forany technical specification is not reduced by this modification. This modificationwillinstall new seismic battery racks designed forthe seismic conditions ofthe Battery Rooms and the loads ofthe new battery cells. The newbatteries have a greater capacity than the existing batteries being replaced therefore themargin ofsafetyis notbeing reduced by this modification. The impact of having a larger battery connected to the DC distribution system has been evaluated.

SEV-1127 DIESEL GENERATOR SUPPLY BREAKER TIMEDELAYRELAYS During safety injection, ifa single safeguards bushas anundervoltage actuation, either degraded or loss ofvoltage, its supplybreaker willtrip and the sister bus supplybreaker on that trainwillalso trip. Each bus willstart a 1.3 second timer that upon timing out closes the bus diesel generator (DG) supply breaker. The sister bus UVsystemwill not actuate since the minimumtime required is the loss ofvoltage relay definite time delay of2.4 seconds.

Ifthe sister bus is 14 or 16, all loads that were sequenced on prior to diesel generator closure would be block loaded. SI sequencer would not be reset.

Ifthe sister bus is 17 or 18, service water would be loaded out ofsequence. Itis also possible to load two service water motors onto the diesel generator. This would exceed design loading during SI.

This modificationwillinstal time delay relays inthe control circuits ofthe diesel generator supply breakers to the safety related 480 VAC busses 14, 16, 17 and 18.

The two time delay relays, set for 0.5 and 3.5 second delay pickups, in each safeguards diesel generator supplybreaker control circuitshall actuate the respective bus UVsystemupon coincident opening ofboth the normal and diesel generator supply breakers. Logic shall allowforlivebus transfers and bus restoration using bus tie breakers.

The new configuration willoperate as follows:

Upon coincident open normal and diesel generator supply breakers to a bus, the new relays begin timing. After0.5 seconds the firstrelay'sNO contacts close to actuate the bus UVsystem. After 3.5 seconds, the second relay times out and the 0.5 second delay relay is de-energized.

The DG supply breaker closes when the DG frequency and voltage are acceptable and the UVsystem resets. Ifthe existing 1.3 second delay relay times out and the DGsupply breaker closes, the 0.5 second relay is de-energized and the UV system resets.

The new configuration performs the following:

Activates a bus UVsystem before UVrelays actuate when abus is de-energized by a normal supply breaker trip.

The 3.5 second delaypickup bridges the gap between the supplybreaker opening and the UVloss ofvoltage relay time out, 2:75 second Technical Specification limit.

Allowsoperators to restorebus voltage through use ofbus-tiebreakers after a3.5 second delay.

Allowsbus transfer &omDGsupplybreaker to normal supplybreaker as currently performed in emergency procedures.

(The 0.5 second delay allows the DG breaker to open and the normal supply breaker to close without a UVsystem actuation.)

The probability ofoccurrence ofan accident previously evaluated inthe SARwillbe changed due to the implementation ofthis modification. The diesel generators are used to supply power to the 480 VAC busses to mitigate the consequences of accidents. After the proposed change is complete the diesel generator supply breakers willcontinue to have the independence" and separation which they are required to have to perform their safety related functions.

Afailure ofthe 0.5 second relay contacts to open introduces a new failure mechanism in the safeguards undervoltage systems. Arelay failure increases the frequency per reactor year ofa safeguards bus failureby 1.91E-6. Asingle relay failure renders abus inoperable by maintaining thebusundervoltage system inthe tripmode. The bus would be energized but the undervoltage actuation would prevent bus loading.

The failureofa relay's contacts to open is a single failure and does not affect the redundant train.

The potential failure ofa safeguards 480 volttrain due to the existing configuration is nota single failure. Combined with a single failure ofthe redundant train's diesel generator the existing configuration could result in a station blackout during SI.

The proposed control configuration significantlydecreases the.equency ofa loss ofa 480 voltbus safeguards train during an SI. However, it increases the frequency per reactor year of an inoperable safeguards bus. The net change infrequency ofa safeguards bus loss is a decrease of two orders ofmagnitude, lE-4 decrease versus 1.91E-6 increase.

Therefore the proposed configuration increases overall plant safety.

The consequences of an accident previously evaluated in the SAR is not increased by this modification. This modification willinstall new time delay elays inthe control circuits ofthe diesel generator supply breakers. Anyfailure ofthe breakers to actuate due to the new relays willbe bounded by previously assumed failures ofthe undervoltage system to actuate and failure of breakers to open/close. No new equipment is being installed and no plant configuration changes are being performed whichwillincrease the consequences ofany previously evaluated accidents.

The probability ofoccurrence ofa malfunction ofequipment important to safety previously evaluated in the SAR is not increased by this modification: As discussed above, there is a net overall decrease intheloss ofa safeguards bus due to this modification. Anewfailure mechanism willbe introduced to the undervoltage system due to this modification, however the increased probabilityofthe loss ofa safeguards bus due to the installation ofthe new time delay relays will be offset by a decreased probability ofthe loss ofa 480 voltbus safeguards train during an SI.

The consequences ofa malfunction ofequipment important to safety previously evaluated inthe SAR is not increased by this modification.

There is no increase in the consequences of a malfunction ofany equipment due to this modification. Ifabus supply breaker failed to close or an under voltage system failed to actuate there would be no increase inthe existing consequences.

The possibilityofan accident ofa different type than any evaluated previously inthe SARis not created bythis modification. No new equipment is being added to the diesel generator bus supply breaker control circuits which can result in an accident ofa different type than any previously evaluated. The new relays are being installed withinthe control circuits ofthe diesel generatorbus supplybreakers, inparallel withexisting time delay relays. The function performed bythebreakers is not being changed.

The possibilityofa malfunction ofequipment important to safety ofa diFerent type then evaluated previously inthe SARis not created by this modification. This modificationwillinstall new time delay relays in the diesel generator bus supply breakers. The relays are safety related and are seismically qualified foruse at Ginna Station. Afailureofany new equipment willnot create the possibilityofa malfunction ofequipmentimportant to safety ofadiFerent type due than previously evaluated for the breakers or undervoltage system.

The margin ofsafety as defined inthe basis forany technical specification is not reduced by this modification. This modification willinstall new time delay relaysin the control circuits ofthe diesel generator supply breakers. There is no impact on Technical Specifications due to this modification therefore there is no change in the margin ofsafety.

SEV-1128 SERVICE AIRSYSTEM UPGRADE PHASE B The scope ofthis modification is to replace the existing Service AirCompressor with a more efficient and reliable oil free, air cooled, two stage rotary screw compressor and heatless regenerative desiccant dryer. This modification also includes the addition ofa cross-connect between the instrument air system and the service air system downstream ofthe new dryer. A check valve and an automatic isolation valve on the service air side ofthe cross-connect will function to prevent service airbackflowinto the instrument air systemand prevent a loss ofservice air pressure fiomdegradin the instrument air system. Although no system or component name changes are involved, this proposed change realisticallyreconfigures the instrumet air system into a fourcompressor system whichtakes advantage ofits increased compressor capacity to supply air to the ser vice air system.

This modificationwillimprove the reliability fboth the service air and instrument air systems and thereby aid inmeeting theNRC requirements oftheMaintenanceRule.

The change installs avery dependable backup to the instrument air systemaswell providing a lowmaintenance source ofair capacity for use as service air.

This modification also includes reconfiguring door F28 such that itcanbe leftopen during periods ofhigh ambient temperature inthe turbine building. The door closer willbe redesigned so that a fusible linkwillallowautomatic closure inthe event ofa fire in the turbinebuildin or allvolatile treatment room.

The instrument and ser vice air systems have no failure modes or effects which are precursors to accidents evaluated inthe SAR. The proposed change does not introduce any newfailure modes or effects tthe air systems or any other system or component which is a precursor to an accident.

Because the proposed change has no interaction with any system which, iffailed, leads to an accident the proposed change can not increase the probabilityofan accident previously evaluated in the SAR.

Pneumatically operated components required foraccident mitigationhave backup systems which provide valve operator mode offorce inthe event the station air systems arelost or degraded. The proposed change has no functional interaction withthe backup systems thus the change cannot cause any equipment failures which would reduce the availabilityofequipment relied upon for accident mitigation. Because the change does not affect the equipment set used to mitigate accidents the change can not increase the consequences ofan accident.

As described the proposed change provides isolation between the service and instrument air systems.

Afterthe change the instrument air system willhave better capacity and equipment reliabilitythen before the change. It can be concluded that the proposed change reduces the

probability ofoccurrence ofa malfunction in a risk significant system (instrument air).

The accident analysis already assumes the unavailability ofthe station air systems.

The proposed change does not introduce any new failure modes or effects into the station air systems. This proposed change can not alter the consequences ofa loss ofinstrument air.

The accident analysis already assumes theunavailability ofthe station air systems.

The proposed change does not introduce any new failure modes or effects into the station air systems. Because the change does not affect any system that can act as an accident precursor itcan not create the possibility ofan accident ofa different type the previously reviewed.

Pneumatically operated components required foraccident mitigationhavebackup systems which provide valve operator mode offorce inthe event the station air systems are lost or degraded. The proposed change has no functional interaction withthebackup systems thus the change cannot cause any equipment malfunctions whichwould reduce the availabilityofequipment relied upon for

.accident mitigation. Because the proposed change onlyinteracts withnon-safety equipment itcan be concluded that the change does not increase the possibility ofa malfunction ofequipment important to safety.

The instrument and service air systems are not credited as inputs to the accident analysis nor are they factors in the basis for any margin ofsafety addressed in the technical specifications.

SEV-1129 CONTROL ROOM HVACUPGRADE PHASE 1

The purpose ofthis evaluation is to determine ifan unreviewed safety question exist withthe planned phase 1 modification to upgrade the reliabilityofthe Control Room HVACsystem by providing a redundant filtrationtrain for th ControlRoomEmergency AirTreatment Sub-system (CREATS). Phase 1 ofthe modification consist of:

1)

Addingtie inconnections to theexisting CRHVACductworkwithblindflanges to allow future connection ofa second charcoal filtertrain.

2)

Replacing MCC K spare breaker 1KKwith an upgraded breaker 3)

Swapping MCCKDCcontrol power fromtrain B to train A, removing 4KVtest cabinet load fromTrainAMainDC distribution panel and placing iton the Turbine BuildingDC distribution panel.

4)

Adding abackdraft damper to the outlet ofthe Control Room AirHandling unit Supply Fan(AKD27).

5)

Replace degraded CRHVACflex duct connectors 6)

Connect a spare cable to one ofthe spare contacts on SI relays SI-16X and SI-26Xfor future use. One cable to each relay.

7)

Disconnect or cap thesupplyand return ducts to the MUXroomand remove doors to the Relay Room 8)

Cutting the existing face plate inhalfand adding a support member on the AuxBench Board inthe ControlRoom to allowfuture on-line replacement fornew trainB controls.

Theprobability ofoccurrence ofanaccidentpreviouslyevaluatedin the SARis not increased since the source terms during a LOCAwillnot be increased bythis modification, itdoes not affect the RCS, Containment Filtrationorthe ECCS. The reliabilityofthe ACand DC systems willnot be decreased since the electrical changes meet the originalplant design and construction standards.

The SI system testing willbe done in a plant mode when the SI system is not required to be operable and procedural actions are inplace to prevent an inadvertent SI actuation. Theprovisions provided for the protection against fires willnot be impacted.

The consequences of an accident previously evaluated in the SAR is not increased by this

proposed change since theperformance ofthe CREATS system, ACand DC systems willnot be degraded from what is assumed in the accident analysis. The SI system functional willnot be affected by testing existing spare wiring and contacts.

The probability ofoccurrence ofa malfunction ofequipment important to safety previously evaluated inthe SARis not increased since the new backdraft damper is similarinconstruction to existing active dampers inthe system and therefore willhave the same type offailure mechanisms and probabilities. The new breaker, cables and flexibleduct connectors are the same or equal to the original equipment theare replacing and therefore their probabilityofoccurrence or malfunction willnot be increased over what was originallyassumed. No new equipment is added to the SI system.

The consequences ofa malfunction ofequipment important to safety previously evaluated inthe SARis not increased. Afailureofthe CREATS system filteringabilitywas already assumed inthe original accident analysis therefore the addition ofthe backdraR damper willnot affect any consequences.

The performance ofthe CREATs systemwill not be degraded and therefore, the consequences ofan radioactive release willnot be increased. The additional DC load on the A Batteries willnot reduce the systems abilityto cope witha loss ofallACforfourhours, therefore the consequences ofa SBO willnot be increased: The SI system testing willbe done in a plant mode when the SI system is not required to be operable and procedural actions are in place to prevent an inadvertent SI actuation.

The possibilityofan accident ofa different type than any evaluated previously inthe SARis not created. Theproposed change creates no new functional interactions withexisting plant equipment nor does itintroduce newfailure modes or mechanisms which could lead to reactor core damage or fission product releas. No new equipment is added to the SI system. The SI contact testing will not change any system configuration an the SI system testing willbe donein aplantmodewhen the SI system is not required to be operable with procedural actions in place to prevent an inadvertent SI actuation..

The possibility ofa malfunction ofequipment important to safety ofa differen type then evaluated previously in the SAR is not created since the cable, damper, breaker and duct material are all currently used at the station and therefore willnot create a differen type ofmalfunction. TrainA DC power is already used willallother trainAMCCloads and there are no trainB loads onMCC K. The loss ofMCC Khas already been considered and therefore, the use oftrain B control power is already bounded by the existing anaysis. The new spare SI cable willmeet exist seperation criteria and fireprotection requirementsan is therefore bounded by existing analysis.

The margin ofsafety as defined inthe basis forany technical specification is not reduced by this proposed modification since the capability or the requirement forthe CREATS system to detect

radiation, isolate the control room, filter 200 CIMoFcontrol room airwillnot be aFFected. The Atrain battery capacity willnot be reduced bythis modification. The SI contact testing willnot change any system configuration.

SEV-1130 NEW PROCEDURE AP-CVCS.3 LOSS OF ALLCHARGINGFLOW New procedure AP-CVCS.3, Loss ofAllCharging Flow, has been developed to deal withthe unique problems associated withthat event. Aloss ofcharging takes away the abilityforRCS makeup atnormal system pressure. Even ifCVCS Letdown is isolated, the RCS continues to lose inventory through the RCP seals. The RCP seals are protected &omhightemperature conditions by CCWflowthroughthe thermal barrier. However, ifaction is not taken, theRCS willcontinue

'to lose inventory untilthe pressurizer empties and pressure control is lost. Abriefdescription of the procedure's high level actions are as follows:

Attempt recovery ofCharging Pumps. Exit procedure ifsuccessful.

b.

Reduce load at 5%/minuteperAP-TURB.5, Transfer 4160 Voltloads, then trip the turbine at 15 Mw.

Shutdown the reactor.

CooldowntheRCS at<100F/hour to 530'F (provided two RCPs are operating),

to allowforRCS depressurization.

This is ashutdownmargin issue forsingle loop operation.

e.

Depressurize RCS to <1950 psig and block Safety Injection.

Restore pressurizer level by starting a Safety Injection Pump and further depressurizing the RCS to approximately 1400 psig.

Energize pressurizer heaters to maintain the pressurizer saturated.

h.

Maintain RCS at stable temperature, pressure and inventory.

The probability ofoccurrence ofan accident previously evaluated inthe SAR is not increased because the new procedureis designed to shutdown the plant and re-establish a means ofinventory control in a controlled manner.

Continued operation ofthe CCW System ensures RCP seal integrity.

The consequences ofan accident previously evaluated in the SAR are not increased because, without operator intervention, a loss ofcharging event would eventually terminate with auto safeguards actuation withthe RCS in a more degraded condition. The new procedure actually reduces consequences ofthis transient.

40-

The probability ofoccurrence ofa malfunction ofequipment important to safety previously evaluated in the SAR is not increased because the new procedure is actually mitigating the consequences ofjust such an event (loss ofcharging).

Other equipment is operated within previously established parameters, withthe exception ofthepressure/temperature limitcurve for RCP NPSH. Although not recommended forroutine operation, the limits established by this procedure for RCS pressure/temperature relationships are sufficient to support safe RCP operation.

The consequences ofa malfunction ofequipment important to safety previously evaluated inthe SARis not increased because the new procedure addresses this type ofevent withoutimpacting the failure consequences ofother plant equipment.

The possibilityofan accident ofa di6erent type than any evaluated previously inthe SARis not created because the intended function and operation ofplant systems is not afFected. Inaddition, the plant is not placed in a configuration which inot analyzed, and this evolution is intended to prevent a more significant transient from occurring.

The possibilityofamalfunction ofequipment important to safety ofa diFerent type than evaluated previously in the SAR is not created because the procedure addresses issues of equipment operational limits. For example, the operator is directed to observe starting duty ofSafetyInjection Pumps when cycling the pumps to stabilize RCS inventory.

The margin ofsafetyas defined inthebasi forany technical specification is not reduced because, although a loss ofallcharging may result inentry into certain LCOs, the procedure is designed to stabilize the plant and restore volume/pressure control, thereby maintaining themargin ofsafety.

41-

SEV-1131 CYCLE 28 RELOAD The reload forCycle 28 consists of44 new fuel assemblies labeled as feed regions 30Aand 30B.

This safety evaluation is validforan End-of-Cycle 27 burnup of15,509 to 16,517 MWD/MTU and for a Cycle 28 burnup that does not to exceed 18,160 MWD/MTUwithout additional analysis. The fuel assemblies to be loaded inthe core forCycle 28 are mechanically the same as the Cycle 27 fuel assemblies except for the following:

The fuel rod clad is fabricated with ZIRLO, an alloy similar to Zircaloy-4.

The thimble guide tubes are fabricated with ZIRLO.

The instrumentation tubes are fabricated with ZIRLO.

Compared withZircaloy-4, ZIRLOhas no Chromium(Cr) and reduced contents ofTin(Sn) and Iron(Fe). Inaddition, ZIRLOhas incorporated anominal amount ofNiobium(Ni). The purpose ofchanging the chemical composition ofZircaloy-4 to that ofthe ZIRLO alloy is to improve corrosion resistance and dimensional stability under irradiation.

Westinghouse has designated the design characteristics offuel assemblies with ZIRLO as VANTAGE+,and those offuel assemblies with Zircaloy-4 as OFA.

The Cycle 28 reload willnot increase the probability ofoccurrence ofan accident previously evaluated inthe SAR because the reload core does not affect accident initiators or equipment operation.

The reload core does not cause a pipe to break or equipment to malfunction.

Therefore, the reload core can not increase the probabilityofan accident previously evaluated in the SAR.

The change toZIRLOas the material forfuelrod cladding, guide tubes, and instrumentation tubes, is not directly related to the probabilityofany accidentpreviously evaluated inthe UFSAR. The use ofZIRLO as a material does not impact the mechanical integrity ofthe fuel rod, or the structural integrityofthe fuel assembly orthe coreunder normal or accident conditions. Alldesign criteria, applicable standards, and safety limits are met.

Because ofthis, there are no new challenges to components and systems that would increase the probability ofany previously-evaluated accident.

Furthermore, the use of ZIRLO as fuel cladding improves corrosion performance and dimensional stability under normal conditions.

The fuel design changes satisfy existing design criteria; therefore, the probabilityoffailure does not increase.

Gap reopening does not affect accident initiators.

The Cycle 28 reload does not increase the consequences ofan accident previously evaluated in the SAR because the core characteristics are bounded by parameters assumed in the accident analyses.

When deviations occurred, reanalysis was performed to show that the acceptance criteria was still satisfied.

The mechanical changes to the fuel assemblies do not degradeperformance.

The ZIRLOmaterial used in the fuel rod cladding, guide tubes, and instrumentation tubes has similar'hysical and mechanical properties to that ofZircaloy-4. Alldesign criteria, applicable standards, and safety limitsforthe fuelrod cladding, guide tubes, and instrumentation tubes using ZIRLOmaterial are met. Therefore, mechanical and structural integrityofthe fuel assemblies willbe maintained under normal and accident conditions.

Ananalysis ofallfuelwithgap reopening demonstrates compliancewiththecorrosion limit set forth in 10 CFR 50.46 for all ofCycle 28.

The radiological consequences ofaccidents withfuel assembliesusing ZIRLOmaterial are the same as those documented in the UFSAR for fuel assemblies using.

Based on the above, the radiological consequences ofaccidents previously evaluated inthe SAR have not increased.

The Cycle 28 reload and fuel assembly changes do not create an accident ofa different type than any evaluated previously in the SAR because (a) the core parameters are bounded by those assumed inaccident analyses, and (b) the design parameters are stillwithinthe assumed ranges.

The required SDMis met forallpower levels and at anytime during the core lifewiththe control rods above the insertion limits in the COLR.

The fuel assemblies ofthe Cycle 28 reload withZIRLOmaterial meet the same design criteria, applicable standards, and safety limits as those fuel assemblies withZircaloy-4 material in the remainder ofthe core. No new single failure mechanisms have been created under normal or accident conditions.

The Cycle 28 reload does not increase theprobability ofoccurrence ofa malfunction ofequipment important to safety previously evaluated in the SAR because the reload core does not acct equipment operation.

The reload core does not cause equipment to malfunction.

Alldesign criteria, applicable standards, and safety limitsare met forthe fuel assemblies withfuel rod cladding, guide tubes, and instrumentation tubes fabricated withZIRLOmaterial. Meeting design criteria and applicable standardsprecludes new challenges to components and systems that could increase theprobability ofmalfunction. No new failure modes orlimitingsingle failures have

been created. The fuel design changes satisfy all design criteria, applicable standards, and safety limits; therefore, the probability offailure does not increase.

Gap reopening is not expected to lead to fuel failure. Violatingthe gap reopening SAFDLcriteria does not result in exceeding the 17% oxidation limit.

The Cycle 28 reload does not increase the consequences ofa malfunction ofequipment important

" to safety previously evaluated in the SAR because the core characteristics are bounded by parameters assumed in the accident analyses.

When deviations occurred, a reanalysis was performed to show that the acceptance criteria was still satisfied.

The mechanical changes to the fuel assemblies do not degrade fuel performance. Alldesign criteria, applicable standards, and safety limits are met for the fuel assemblies with fuel rod cladding, guide tubes, and instrumentation tubes fabricated withZIRLOmaterial. Meeting design criteria and applicable standards precludes new challenges to components and systems and/or a challenge to the integrityofthe fuel rod cladding. No new failure modes orlimitingsingle failures have been created.

The doses documented in the UFSAR remain una6ected by the change in material to ZIRLO in the fuel assemblies.

Gap reopening does not a6ect the consequences ofequipment malfunction.

The Cycle 28 reload, withthe associated mechanical changes to the fuel assemblies, does not create the possibilityofanew type ofmalfunction ofequipment important to safety notpreviously evaluated inthe SARbecause(a) the core parameters are bounded by those assumed inaccident analyses, (b) the design parameters are stillwithinthe assumed ranges, and (c) the limitations imposed by the MSLB are within normal ranges ofoperation.

Alldesign criteria, applicable standards, and safety limitsare met forthe fuel assemblies withfuel rod cladding, guide tubes, and instrumentation tubes fabricated withZIRLOmaterial. Meeting design criteria and applicable standards precludes new challenges to components and systems and/or a challenge to the integrityofthe fuel rod cladding. No new failure modes orlimitingsingle failures have been created.

Previous analyses assumed no gap reopening forsimplicity. Analyses withgap reopening show acceptable consequences.

This condition is acceptable provided that continued compliance with the 17% oxidation limitis maintained. This compliance has been confirme witha cycle-specific corrosion analysis.

The Cycle 28 reload, with the associated mechanical changes to the fuel assemblies and the identified limitations, does notreduce the margin ofsafety as defined inthe basis forany technical specificationbecause itmeets all design criteria, applicable standards, and safety limitsset forthin 44-

the licensing basis.

Withrespect to the fuel assemblies, the two design types, OFA(Zircaloy-4) and VANTAGE+

(ZIRLO), meet all the design and safety limits.

SEV-1133 MINIMUMAUXILIARYFEEDWATER TEMPERATURE OF 32'F The purpose ofthis evaluation isto determine ifanunreviewed safety question exist withreducing the minimum temperature assumed in accident analysis forauxiliaryfeedwater from50'F to 32'F.

The only components and systems affected bythe proposed change are the Condensate Storage Tanks (CSTs) and AuxiliaryFeedwater System(~ whichprovide auxiliaryfeedwater to the steam generators(SGs).

Normally, this is during plant startup/shutdown orfollowinga reactor trip.

The function ofthe CSTs and AFWis to supply feedwater to the SGs fordecay heat removal or during periods oflowpower operation when the inain feedwater system is secured and/or the turbine is not latched. Reducing the minimum temperature increases the heat removal ofthe auxiliary feedwater which is beneficial except for over cooling transients, which have been evaluated.

Implementation ofthis change does not increase the probability ofoccurrence ofan accident previously evaluated in that the change does not introduce any effect that could increase the probability ofan accident. The systems affected by the change have been evaluated forthe new minimum temperature, and it is within there design.

The affected accidents have been reevaluated withthenewlowertemperature.

Implementation ofthis change does notincrease the consequences ofan accident previously evaluated inthat the consequences meet the required acceptance criteria; thus the consequences are acceptable.

Implementation ofthis change does not increase the probabilityofoccurrence ofa malfunction of equipment important to safetyin that the temperature reduction does not cause equipment orpiping to operate outside its design temperature range.

,t Implementation ofthis change d'oes not increase the consequence ofa malfunction ofequipment important to safety previously evaluated inthat the change does not impact the capability to meet the accident analysis nor does itadversely impact the abilityofany equipment to perform their intended safety function.

This change does not introduce the possibility ofan accident ofa different type than previously evaluated in that the change affects only the parametric value used by the current analysis.

This change does notintroducethe possibility ofan equipment malfunction ofadifferen type than

~ previously evaluated in that the change affects only the parametric value used by the current analysis.

This change does not reduce the margin of safety as defined in the basis for the Technical Specification inthat the accidents afFected stillmeet the required acceptance criteria. Since the acceptance criteria are met there is no reduction in the margin ofsafety.

SEV-1134 REACTOR 1NTERNALS BAFFLEBOLTREPLACEMENT The purpose ofthis safety evaluation is to examine the acceptability of1999 baEe-former-bolt replacement strategy.

Anewlydeveloped body ofinformation regarding bafHebolts concludes thatplants can be safely operated and shutdown, during normal, upset and faulted conditions, with fewer bafHe-former-bolts than are initiallyinstalledinthe formers provided that physically sound bolts are configured in a pattern which provides forthe correct structural resistance to the credible forces applied.

TheNuclear Regulatory Commission(NRC) has evaluated the acceptability ofperforming baffle-boltreplacements under the auspices oflOCFR50.59. (NRC Safety Evaluation ofTopicalReport WCAP-15029, "Westinghouse Methodology forEvaluating the AcceptabilityofBafHe-Former-Barrel Bolting Distributions Under Faulted Load Conditions", TACNo: MA1152.)

Inthe review ofWCAP-15029 theNRC concluded that the general methodology presented was acceptable provided that: 1) the limitingbaftle bolt loading willbe determined by analysis fora class ofplants and a specific break; 2) the noding to be used inthe representation ofthe loading is demonstrated to be adequate by performing nodalization sensitivity studies orby some other acceptable methodology. The review furthermorerequired the demonstration ofconservatism in projected bolting material properties and forthe accounting oflimitationsinthe inspection methods used to detect flaws.

The Rochester Gas and Electric(RG&E)bafHe-former-bolt replacement strategy differe slightly fromthe generic Westinghouse Owners Group (WOG) strategy (WCAP-15036, revision 1).

RG&Eperformed ultrasonictesting ofalltheinspectablebolts.

This examination identifie 59 bolts withflawindications. The replacement program changed 40 ofthesebolts along with9 bolts that could notbe tested and 7 bolts classified a suspect. All19 bolts withdefect indications inFormer Plate levels 2 through 6 were replaced regardless oftheir location withinthe WOGpattern. All bolts that exhibited defects inFormer levels 1 and 7 were left in service. Ofthe 6 bolts which exhibited flaws inpositions required to conform to the WOGpattern inlevels 1 and 7, sufficient adjacent bolts outside ofthe patternwere verifiedto be present to structurally compensate forthe potential defects.

The RG&E pattern ofanalysis thus incorporated both new and aged, but verified acceptable, bolts. The end result ofthis methodology yielded an output which provides reasonable assurance that allthe fasteners relied upon withinthe analysiswill function as intended.

The technical evaluations performed for RG&Es baffle-former-bolt replacement program demonstrate that the reactor internals package willfunction as required under faulted load conditions, as wellas comprehensively addressing theNRCs criteria formaking the changeunder 10CFR50.59, as delineated in their review ofthe WCAP-15029.

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The baffle-former-barrel and baffle-former-bolts react to the stresses caused by the accidents

, evaluated inthe SARbutdo not have a failure modethat is a precursor to any analyzed accident.

Because theunits do not have a failure mode that leads to an accident the former-bolt replacement can not increase the probability ofoccurrence ofa accident previously evaluated in the SAR.

Aftercompletion ofthe changethebafile-former-bolting arrangement willhave sufhcient structural integrity to resist the force loading caused by the analyzed accident set.

Because the as-left condition willperform itsfunction inaccordance withthe design requirements the consequences ofan accident previously evaluated in the SAR willbe unchanged.

The baffl-formers physically and functionallyinteract withthe reactor core barrel and provide lateral fuelsupport and reactor coolant flowdirection. Thebaffle-former-bolting replacement does not add any new functional interactions with equipmentimportant to safety. The analysis ofthe as-leftbaffle-former-bolting configuration demonstratestha theunits have sufBcientintegrityto resist all credible design basis loading conditions. Afterthe change application ofthe design accident loading forces to the units willnot cause them to physically fail,or otherwise deform, to an extent which could cause fuel damage or preclude the abilityto cool the fuel followingan accident.

Accordingly, itcan be concluded that the probability ofoccurrence ofamalfunction ofequipment important to'safety previously evaluated in the SAR is unchanged.

The analysis ofthe as-left baffle-former-bolting configuration demonstrates that the units have suf5cient integrityto resistall credible designbasis loading conditions. Afterthe change application ofthe design accident loading forces to theunits willnot cause them to physically fail,or otherwise deform, to an extent which could cause fuel damage or preclude the ability to cool the fuel followingan accident. Afterthe baffl-boltreplacements the reactor core internals willrespond to the stress of accidents and malfunctions the same way as before the change. Therefore the consequences ofamalfunction ofequipmentimportant to safety previously evaluated inthe SAR is unchanged.

Thebaffle-former-bolting replacement program does not alter the design, fit,form, or functionof the reactor internals. There are no new functional interactions created bybolt replacements. After completion ofthe changethe reactor internals packagewill function as before thechange.

Because this change does not introduce any new design or functional interactions it can not possibly introduce the potential foran accident ofa different type then previously evaluated inthe SAR.

Thebafile-former-bolting replacement program does not alter the design, fit,form, orfunctionof the reactor internals. There are no new functional interactions created bybolt replacements. The analysis ofthe as-leftbafHe-former-bolting configurationdemonstrates that theunits have suf5cient integrityto resist all credible design basis loading conditions. AAerthe change application ofthe design accident loading forces to the units willnot cause them to physically fail, or otherwise deform, to an extent which could cause fuel damage or pr'eclude the abilityto cool the fuel 49-

followingan accident. Aftertheba61e-bolt replacements the reactor core internals willrespond to the stress ofaccidents and malfunctions the same way as before the change. This change does not introduce any new design orfunctional interactions therefore it cannotpossibly introduce the potential for a malfunction ofa different type then previously evaluated in the SAR.

Technical specifications presumes that the reactor vessel internals are configure to function as designed and licensed. The ba61e-formers and ba61e-former-bolts are not specifically addressed inthetechnical specifications. Because, after thebolt replacements, thevessel internals willfunction as technical specifications assumes itcan be concluded that the margins ofsafety presumed inthe bases for technical specifications willnot be reduced.

10CFR50.59 SAFETY REVIEWFOR PCN 98-3013 TO PROVIDE GUIDANCEON THE TWO OPERATOR CONCUIUMNCERULE RG8cE responses to IEBulletinNo.79-06A, REVIEWOF OPERATIONALERRORS AND SYSTEM MISALIGNMENTSIDENTIFIED DURING THE THREE MILE ISLAND INCIDENT,state that administrative procedures were modified to require that two licensed operators shall agree on any overriding before the overriding action is executed on any safeguard system active component. The change modifies thatcommitmen to now require that the action be approved by the SRO that has the command function. This change stillmeets the intent ofthe NRC guidance which is to have the control board operators not stopping active components without thoroughly evaluating the conditions and consulting other individuals (ie, the SRO).

The change willresult in the procedural direction remaining in compliance with Technical Specification 5.1.2, which states that the Shift Supervisor or another SRO inhis absence shall be responsible for the control room command function.

The change willnot alter the description inthe UFSARwhich states that the Shift Supervisor is responsible forthe performance ofallpersonnel assigned to his shiftwho could afFect plant safety, regardless ofspecialty afhliation. The UFSAR also states that the operating shiftcrews conform to the requirements for shift complement as specified in 10 CFR 50.54 (l),where an individual licensed as a senior operator is to be responsible for directing the licensed activities oflicensed operators.

10CFR50.54 (x) and (y) also state that a licensee may take reasonable action that departs &oma license condition or a Technical Specification inan emergency when, as aminimum, it is approved by a licensed senior operator.