ML17265A766

1999 Rept of Facility Changes,Tests & Experiments Conducted Without Prior NRC Approval for Jan 1998 Through June 1999, Per 10CFR50.59.With 991020 Ltr

ML17265A766
Person / Time
Site:	Ginna
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

~~u ~~~~ J. 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.

DISTRIBUTION CODE: IE47D COPIES RECEIVED:LTR ~ ENCL ( SIZE:

TITLE: 50.59 Annual Report of Changes, Tests or Experiments Made W/out Approv NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244

(

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

RGN1 FILE 01 1 1 EXTERNAL: NOAC NRC PDR 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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AHD A Subsidiary of RGS Energy Group, Inc.

ROCHESTER GAS AND ElECTRIC 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 of Facility 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 in support ofproposed changes to the facility and 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 of Reactor Projects I/II Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U.S. NRC Ginna Senior Resident Inspector

1999 REPORT OF FACILITYCHANGES, TESTS, AND EXPERIMENTS CONDUCTED WITHOUT PRIOR NRC APPROVAL

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

..9910290070

SEV-1090 TECHNICAL SPECIFICATION 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..." in accordance with a sensitivity analysis. Revision 2 ofthis evaluation supports a change in the minimum operating temperature of the service water system from 32'F to 30'F.

The probability ofoccurrence ofan accident previously evaluated in the 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 probability ofan accident. In addition, the reduction in the temperature does notadversely impact the ability ofany equipment to perform their intended safety function.

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

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SARis not increased, because the change affects aparameter associated withthe SW system fluid and is minor in nature. A design analysis evaluated the impact of the fluid affect temperature change from 32'F to 30'F. The results did not show a reduction ofthe structural ability integrity ofany components relied upon and hence the designbasis would notbe affected by the reduction in temperature limit to 30'F.

ofanyequipmen The consequences ofa malfunction ofequipment important to safety previously evaluated in the SARare not increased, because the change proposed does not relied upon to mitigate consequences from performing their functions. The structural integrity of critical components and their capability is not impacted by the SWtemperature. Revision 1 ofthis evaluation examined the impactofa30'F SW temperature on containment fan cooler performance and the affect on PCT, with the limiting case 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 of an accident of a 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 of accident mitigating equipment.

The possibility ofamalfunction ofequipment important to safety ofa differen type than evaluated previously in the SARis not created, because the temperature reduction from 32'F to 30'I will not reduce any ofthe performance characteristics ofcomponents, such as valves, pumps, or heat exchangers, and willnot affect the structural integrity or stress levels ofpiping or pipe supports.

Components receiving service water flowfor the purpose ofremoving heat from the other fluid medium, are not adversely impacted and not subject to freezing, since the fluid medium 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 of safety.

SEV-1094 REPLACEMENT OF RTD INPUT MODULES IN THE REACTOR PROTECTION RACKS The electronic components used to generate the T, and b, T signals in the 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-line modules whichwillbereplaced with 24 modules manufactured by NUS. The new module arrangement will consist of 16 Resistance-to-Current(R/Q converters, and eight TimeDomain Modules (all safety grade analog devices). Each protection channelwill have fourMconvertors thatwillbeused for the conversion of Hot 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 with the temperature signal. The insertion ofinstrument loop lag time provides a compensating factor for the extremely fast responding loop RTDs with respect to the rest ofthe instrument loop. The lag time factor 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 for Channe12 oftheReactorProtection System have been modified under the first phase of PCR 97-026.

The probability ofoccurrence ofan accident previously evaluated in the 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 of malfunction.

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SARis not increased. The change does not introduce any new failure modes or effects into the affected T,, or 4T instrument loops nor does it functionally modify the loop or associated RPS and control systems in any way not originally designed for.

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

The possibility for an accident ofa diferent typethan any evaluated previously in the SARis not created. The proposed change does not create any new equipment interactions.

The possibility ofa malfunction ofequipment important to safety ofadifferen typethen evaluated previously in the SARis not created. The proposed change does not create any new equipment interactions.

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

SEV-1100 RWST ACCIDENT ANALYSIS UPPER TEMPERATURE LIMIT The accident analysis assumes aRWST temperature range of60'o 80'F. Recent temperature measurements in the AuxiliaryBuilding indicate theupper temperature limitshould be increased.

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

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

The probability of occurrence of an accident is not increased by the assumption of RWST temperature. The RWST is not an accident precursor and therefore the change in maximum allowable temperature will not affect the probability of occurrence for any accident analysis described in the UFSAR.

The consequences of an accident have not increased because the acceptance criteria for the accident are still met. 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 in RWST temperature still satisfy the limits established by GDC 19 and 10CFR100.

The probability of occurrence of a malfunction is not increased by the assumption of RWST temperature. The temperature increase from 80'o 104'F is within the design ofthe affected systems and therefore there is no change in the likelihood of failure.

thethermal The consequences ofa malfunction have not increased because the acceptance criteria for the accident are stillmet. The peak containmentpressure as a result ofthis change remainsbelow the affects limitof60psig and therefore the control room and off-site dose radiological consequences due to hydraulic the increase inRWST temperature still satisfy the limits established by GDC 19 and 10CFR100.

Increasing the assumed RWST temperature by 24'F does not cause a differen type ofaccident than previously evaluated. The temperature change slightly roperties of the water which would not cause a new type of accident.

Increasing the assumed RWST temperatureby24'F does not cause a differen type ofmalfunction than previouslyevaluated. The temperature change slightly affects thethermal hydraulic properties of the 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 for containment. This value has not been exceeded by increasing the upper limit on RWST temperature. Therefore, there is no chang'e in the margin of safety.

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 of the A Diesel Generator Lube Oil and Jacket Water coolers coincident with the monthlyADiesel Generator run doneunder PT-12.1. The fouling in the Diesel Generator Aheat exchangers is determined analytically from PT-60.4 test measurementsusing a well developed methodology. Theuncertaintyin the determination offouling is stronglydependent difference on the service water temperature acoss the coolers. In order to reduce theuncertainty in the fouling, the service water willbe throttled to approximately 250 gpm. The followingchanges are evaluated:

¹ PCN 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 INOPERABLE for the duration of time that valve 4671 is unlocked and throttled.

PCN ¹ 97-4346 adds a precaution to PT-60.4 to have ari observer continually monitor the lubricating oil and jacket water outlet temperatures from Diesel Generator A, and record the values on a ten-minute frequency, whenever the engine is running and the service water is throttled. In the event that thejacket water temperature rises above the alarm setpoint of 182'F or the lubricating oil temperature rises above the alarm setpoint of 195'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 to PT-60.4 are inconsequential. They involve installation 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 of the tests. configurationofany The proposed changes do not increase the probability ofoccurrence of an accident previously evaluated in the SAR. The emergency dieselgenerator is not an accident initiator, and temporarily throttling service water to the diesel generatorcoolerswillnotchange the othe system in such a way as to impact the probability of another 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 locking valve 4671. In addition to the normal MCB alarm, Diesel Generator Awillbe continually monitored locally to verify that thelube oil and jacket water temperatures do not exceed the alarm setpoint values. In the event that temperatures reach alarm setpoints, test personnel willtake immediate action to open valve 4671. Therefore,

the probability offailure ofDiesel Generator Ais no higher than itis during theregular monthlyPT-12.1 Surveillance Test.

Theproposed changes do notincrease the probability ofoccurrenceofamalfunction 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 in the SAR. Accident analyses already assume the loss ofa diesel generator.

The proposed changes do not increase theprobability ofan accident ofa different type than any modifications evaluated previously inthe SAR. The proposed changes involveminor t a test that is routinely carried out. The most severe occurrencewould be the tripping ofDiesel Generator A to prevent it from 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 in the SAR. The Diesel Generator Alube oil and jacket water temperatures willnot be allowed to rise above the currently established alarm setpoints. It has been established by the vendor that these are acceptableoperating temperatures for the diesel engines.

The margin ofsafety as defined in the basis for any technical specification is not reduced, since no Technical Specifications are violated. Since the normal configuration of the 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 VACUUMFILL OF 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. final fill This procedure is to beused during mode 5 prior to and during the process RC loop fillprocess. It willbe installed only during this process and willbe removed when RCS refillis complete. The willbeincorporated vacuum int procedures 0-2.3.1 and 0-1B. The present method modification ofRCS system fillrequires along and complicated vent procedure. This wil allow a vacuum to drawn on the RCS when atmidloop in order to allow theRCS to be filledwithout the need for venting.

The initial conditions for RCS vacuum fillare established during RCS low loop 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. Low loop procedure 0-2.3.1 will be in effect, the level band restricts RHR flow to 800 gpm.

Thevacuum operation willconsist ofa vacuum pump connected via 2 inch diameter vacuum rated hose to the pressurizer reliefoutlet piping to the pressurizer relieftank(PRT). There willbe an option vacuum hose for thereactor vessel head vent. The pressurizer PORV andBlock valves will be open to allow the PRT gas space and pressurizer relieftailpipes to be connected to the RCS.

The pressurizer and PRT vent manifolds will supply 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 low temperature 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 at low temperatures, therefore the initial conditions and probability ofoccurrence 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 dilution and boration paths to theRCS. The vacuum process willnotinfluence coolantboronconcentration, therefore the initialconditions and consequences ofan accidentpreviously described in the UFSAR for reactivity insertion have not changed and are valid.

The wall thickness ofthe pressurizer, steam generators and U-tubes, reactor coolant pumps and associated components exposed to the vacuum is sufficient to maintain the integrity ofthe systems during vacuum venting, and after the fill process is complete.

The integrity ofthe reactor coolant pump seals is assured by maintaining a positive pressure at the number 1 seal inlet area. The pressurizer relieftankis designed to withstand a fullvacuum. The tank is equipped with an internal support for the rupture disk to prevent the damage to the disk.

Therefore the integrity ofthe RCS remains unchanged and the probability of occurrence of a malfunction of equipment important to safety is not increased.

The containmentisolation system willremainunaffected by this change. The systemwill still be able by 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 ability to isolate containment during reduced RCS inventory operations remains unchanged and the probability of occurrence of a malfunction of equipment important to safety is not increased. the

~

The ability ofthe ResidualHeat Removal system to provide for core 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 verify that operating in midloop with the RCS at a vacuum did not invalidate its areaffected analysis. The analyses for vortexformation were most sensitive to fluidvelocity with the density and viscosity ofthe fluid as secondary affects. None ofthese parameters RC being under a vacuum. The analysis therefore remains valid. There is no increase in the consequences of a 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 "Draining and Operating atReduced Inventory intheReactor Coolant System".

This procedure implements RGB's response to generic letter 88-17 concerns. The RCS is maintained in an 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 possibility for an accident ofa differen type than any evaluated previously in the UFSAR. configuration With the steam generator intact and the pressurizer manway installed, the criteria is met for the RCS intact configuration. This wa analyzed and is one ofthe configuration that WCAP-11916 and Generic Letter 88-17 reviews. Therefore, the possibility ofa malfunction of the RCS boundary of a different type than evaluated previously in the UFSAR is not created.

A KYPIPE analysis (noted on "Expeditious Actions" response to the NRC, dated January 4, 1997) oftheRHRsystem verified that the gravity feed 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 fiowfrom the RWST. The final recovery 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 in the basis for any 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 of the Diesel Generator B Lube Oil and Jacket Water coolers coincident with the monthlyDiesel Generator B run doneunderPT-12.2. The fouling in theDiesel Generator B heat exchangers is determined analytically from PT-60.5 test measurements using a well developed methodology. Theuncertaintyin the determination offouling is strongly dependent on the service water temperature difference across thecoolers. In order to reduce theuncertainty in the 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 INOPERABLE for the duration of time 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. In the event that thejacket water temperature rises above the alarm setpoint of182'F or the lubricating oil temperature rises above the alarm setpoint of 195'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 involve installation 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 of the tests.

The proposed changes do not increase the probability of occurrence ofan accident previously evaluated in the SAR. The emergency diesel generator is not anaccident initiator, and temporarily throttling servicewater to the diesel generatorcoolerswill not change thecon6guration ofany other system in such a way as to impact the probability of another 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 locking valve 4672. In addition to the normal MCB alarm, Diesel Generator B willbe continuaHymonitored locally to verify that the lube oil and jacket water temperatures do not exceed the alarm setpointvalues. In the event that temperatures reach alarm setpoints, test personnel willtake immediate action to open valve 4672. Therefore, 0

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

the 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 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 in the SAR. Accident analyses already assume the loss ofa diesel generator.

The proposed changes do not increase the probability ofan accident ofa difFerent type than any evaluated previously in the SAR. The proposed changes involve minormodification to a test that is routinely carried out. The most severe occurrencewould be the tripping ofDiesel Generator B to prevent it from overheating. Contingent actions stemming from a diesel generator trip are already available.

H The proposed changes do not increase the possibility ofa malfunction ofequipment important to safety ofa di6erent type then evaluated previously in the SAR. The Diesel Generator B lube oil and jacket water temperatures willnot be allowed to rise above the currently established alarm setpoints. It has been established by the vendor that these are acceptable operating temperatures isnotreduce, for the diesel engines. specification The margin ofsafety as defined in the basis for any technical since no Technical Specifications are violated. Since the normal configuration of the 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 DURING RCS VACUUMVENT AND FILL 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 mid loop level. This level is 10 inches using local level indication and is at the 246'10" elevation. The time duration ofthe exposure to fill 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 in the 80% (180 inches) wide range in the pressurizer, the vacuum willbe removed and the system willbe returned to normal operational pressures. procedure The RCS vacuum vent and wil maintain positive control over the RCS vents and the low temperature 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 low temperatures per UFSAR chapter 5.2.2. The RCS instrument system willcontinue to accurately monitor and display the process variables needed to verify RCS parameters. Therefore the initial conditions and probability ofoccurrence 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 dilution and boration paths to the RCS. The fill RCS instrument system willcontinue to accurately monitor and display the process variables needed to verify RCS parameters. The vacuum process will not influence coolant boron concentration, therefore theinitial conditions and consequences ofan accidentpreviouslydescribed in the UFSAR for reactivity 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 integrity ofthe systems during vacuum venting, and after the proces is complete. The integrity ofthe 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 with an internal support for the rupture diskto prevent the damage to the disk. Therefore the integrity ofthe RCS instrumentation remains unchanged and the probability ofoccurrence of a malfunction of equipment important to safety is not increased.

The containment isolation system and its associated instrumentation willremainunaffectedby this change. The system willstill be 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 ability to isolate containment during reduced RCS inventory operations remains unchanged and the probabilityofoccurrence ofamalfunction ofequipment important to safety is not increased.

The ability ofthe Residual Heat Removal system to provide for core cooling when the RCS is in a reduced inventory condition willnot change. TheNPSH available for the RHR pumps is greater than required, therefore the capability ofRHR system 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 for vortex formationweremost sensitive to fluid velocity with the density and viscosity of the fluid as secondary affects.

TheRCS and RHRinstrument systems willcontinue to accurately monitor and display the process variables needed to verify their 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 of a 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 "Draining and Operating at Reduced Inventory in theReactor Coolant System".

This procedure implements RGB's response to generic letter 88-17 concerns. The RCS is maintained in an 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 possibility for an accident ofa differen type than any evaluated previously in the UFSAR.

configuration With the steam generator intact and the pressurizer manway installed, the criteria is met for the RCS intact configuratio. This wa analyzed and found acceptable in WCAP-11916.

Therefore, the possibility ofamalfunction 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 limits as stated in the Pressure TemperatureLimits Report(PTLR)

're not exceeded. The RCS and RHR instrument systems willcontinue to accurately monitor and display the process variables needed to verify their parameters. The shutdown requirements and PORV operability limits for the RCS are maintained. The margin ofsafety for the reactor coolant pressure boundary as defined by the ASME code for wall 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 of 15,200 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 for Cycle 27 aremechanically the same as the Cycle 26 fuel assemblies except for the following.

The use of annular pellets in the axial blankets, Areduction in backfil pressure in Integral 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 of occurrence of an 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 probability ofan 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 probability ofamalfunction 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 probability of 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 in the accident analysis. When deviations occurred, reanalysis was performed to show the acceptance criteria was still satisfied. 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 limit for all of Cycle 27.

The Cycle27 reload does not increase the consequences ofmalfunction ofequipmentbecause the core characteristics areboundedby parameters assumed in the accident analy'sis. When deviations occurred, reanalysis was performed to show the acceptance criteria was still satisfied. 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 of a 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 in accident 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 in accident 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 limit is maintained.

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

No gap reopening is a Westinghouse design criteria used to simplifythe design process. Analyses with gap reopening show all aspects ofplant safety analyses remain bounding. The screening criteria provides the point at whichcompliance with the 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 Simplification provide an simplifie alternative to procedure PT-60.3. wa desired to reduce the number ofpeople and amount ofequipment that would be required in containment to facilitate on-power testing. The new procedure onlyprovides information necessary to determine the fouling ofthe Containment Recirculation Fan Coolers (CRFC). It DOES NOT test the CRFC motor coolers.

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

1. Throttling the servicewater flowto eachCRFC downto-300 gpm from the 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 low flow alarms on FIA-2033, FIA-2034, FIA-2035, and FIA-2036. The operators are informed before these manipulations are done.

4. Positioning and repositioning of A-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 of an 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 of containment HVAC inoperable, which is already assumed in accidentanalyses. Manipulations on other systems, other than the servicewater supply to the inoperable train of CV HVAC, are not performed as part of the 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 profile containment HVAC, and the environmental met with a containmentHVAC qualification train out ofservice. Moreover, manipulations ofother systems and equipmentimportant to safety are not performed as part of the PT-60.3A procedure, so there is no associated increase in probability or consequences.

PT-60.3A does not increase the probability ofan accident ofa different type than any evaluated previously in the 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 of test equipment only.

PT-60.3Adoes not increasethepossibility ofa malfunction ofequipment important to safety ofa diFerent type than evaluated previously in the SAR. The procedure involves manipulation of service water system valves in the 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 of this procedure.

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

- charcoal system), which is declared inoperable under LCO3.6.6 for testing. With the exception

'of service 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 limit of 120'F as defined in LCO 3.6.5.

SEV-1111 FUEL ASSEMBLY REPAIR PROCEDURE RF-73.1 In order to repair (reconstitute) selected fuel assemblies the preferred technique is to remove the top nozzle which allows access to the fuel pins. This differs from past 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 it is desirable to utilize this method of reconstitution.

Fuel reconstitution is accomplished byremoving defective rods and replacing them with "dummy" stainless steel rods. The acceptability ofusing a reconstituted fuel assembly in the reactor is not covered by this safety evaluation as that will be covered by a revision to the reload safety evaluation. This evaluation covers the process of reconstitution only.

The general process for reconstitution 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 with a special reconstitution basket that is compatible with the reconstitution tooling. Once the fuel assembly has been placed in the elevator the elevator willbe raised to a height where the top nozzle lock tubes can be removed.

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

The GinnaUFSAR states that the new fuel elevator isused for new fuel only. Since this procedure willdeviate &omthat description this safety. evaluation is being prepared to describe the additional use of the 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 in SARis not water 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 limited in scope so as to not damage any fuel pins the consequences of afuel handling accident remain bounded by sufficient the evaluated accident.

The probability ofamalfunction ofequipment important to safetyis notincreasedbecause multiple layers ofadministrative and physical controls areinplaceto maintain 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 in place to preclude overexposure ofplant personnel as well as the public from reconstitution activities.

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

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

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

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

I 0

SEV-1112 ACTION REPORT 97-1846 DISPOSITION FOR MAINSTEAM LINE A AND B CRACK REPAIR AT PENETRATION 401 AND 402 As a result ofnew ISI inspectionmethods for integral attachments to piping/components cracks were discovered in the 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 of the cracks and determine ifthe affected systems are operable. This revision of the safety evaluation was performed to update the references to the supporting analysis.

The root cause ofthe cracks was found to be due to poor weld joint design, referred to as a tee joint, which caused high residual stresses in the 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 in teejoints is a well know phenomena(Lamellar tearing) which was identifie in the late 1960s for large sectionstructural members. The literaturereviewed shows cracks starting fi'om the 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 initiating cause 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 wall for defect removal. Cracks were not found in any ofthe area between the outer toes ofthe two filletwelds on either side of the 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 with the original plant construction code. The maximum crack depth was found to be less than 5/8" in all 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.

except with thinner The FW system was found to have the same penetration design as the MSconfigurationswhic members and smaller filletwelds. The inspections did not reveal any cracks. A review was also done ofthe remainder oftheMS and FW system for other potential teejoint have the potential for cracks. No other attachments were found whichwerehighlyrestrained and had weld sizes large enough to generate high residual stresses. A third 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 probability ofoccurrence or the consequences ofan accident or malfunction ofequipment important to safety previously evaluated in the UFSAR are not increased by the proposed repair since the capability oftheMS line penetrationsto resist design loads has not been reduced beyond what was originally assumed.

The possibility for an accident or malfunction ofa differen type than evaluated previously in the UFSAR will not be created by the proposed repair. Since the repair meets the original code requirements and design basis, and willnot change the function ofthe penetrations, no new types of accidents or malfunctions would be introduced.

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

SEV-1114 CONTAINMENTRECIRCULATlNGFAN COOLER COIL REPLACEMENT were Theoriginal Westinghouse Sturtevant Containment Recirculating Fan Cooler(CRFC) coils were replaced under EWR 5275 with enhanced design Marlo coils during the 1993 refueling outage.

modifications Piping als madein the vicinityofthecoils to ease inspection and maintenance of the coils. by Alargenumber ofUFSARchanges were made as a result ofthe CRFC coil replacement 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 for this EWR was taken to Revision 1, but this later revision was never approved by PORC. This deficiency was discovered during the Service Water (SW) deficiency System Safety System Functional Inspection (SSFI) performed by Sargent 2 Lundy Engineers LLCin Apriland May 1997. This Revision wa documented by an Action Report. The purpose ofRevision1 was to close out open items identified by the original safety evaluation. Although a identified number ofthe Revision 0 open items were addressed by Revision 1, a number ofopen items were still . Close out ofthese additional open items was documented by inter-oKce correspondence prior to start-up from the 1993 Refueling Outage.

asitwa This document serve as the final 10CFR50.59 Safety Evaluation of record for this plant modification configuration change. As such, it documents the actions taken in 1993 prior to plant start-up to modification close-out all ofthe open items identified in the originalEWR5275 safety evaluation. This safety evaluation willbe applicable to the completed in 1993; it willNOTattempt to reconcile issues discovered between the time the 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 by EWR 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 still bebounding when compared to theanalysis flow with the replacement CRPCs. The control room and off-site dose radiological consequences due to the reduction in CRFC air rateunde design basis accident conditions still satisfy the limits established by GDC 19 and 10CFR100.

The replacement ofthe CRPCsbyEWR 5275 does not increase the probability ofoccurrence of a malfunction of equipment 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 for other components.

Since the normal operating and designbasis containmentpressure and temperature profile are not adversely affected by the CRFC replacement, the EQ pressure and temperature profiles for safety related equipment in containment is still bounding. The SW flowto safety related loads supplied in parallel with the 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 of equipment important to safety as previously evaluated in the UFSAR. The functional heat removal and iodine removal capability of the CRPCs following design basis accidents has not been adversely affected by the 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 different type than those previously evaluated in the UFSAR. The operation ofthe CRFCs does not initiate any design basis accidents. The replacement CRFCs are similar in function, design and operation to the original CRFCs. The change in CRFC coil design 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 in the UFSAR.

The operational characteristics ofthe replacement CRPCs is similar to 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 coil design.

The only changes to the coils is enhanced tubing materials for corrosion and erosion concerns and increased heat removal characteristics due to a different tube bundle design. None of the enhancements incorporated into the new CRPCs can cause a new type ofCRFC malfunctionwhen compared to the original CRFCs.

The replacement ofthe CRPCs by EWR 5275 does not decrease the margin ofsafety as defined

in the basis for the 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 LOCA still satisfy the requirements of 1OCRF100. Control Room doses due to a design basis LOCA still satisfy 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 of 60 psig.

SEV-1115 REMOVALOF CRDMGA AND CRDMGB REVERSE POWER PROTECTION UFSAR Section 7.7.1.2.5.1 "Alternating Current Power Connections" takes credit for tripping out an MG set on a reverse power condition. The Control Rod Drive System original design included reverse power protection. This protection was removed (reference TSR 91-167, TM 93-031, and EWR 10322) due to several occurrences of undesired inadvertent tripping ofthe MG sets.

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

The probability ofoccurrence ofan accident previouslyevaluated in the SARis not increased. The change does not introduce any new failure modes or effects into the Control Rod Drive system(not already reviewed in the accident analysis).

The consequences ofan accident previously evaluated in the 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 in the 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 in the SARis not increased. The change does not introduce any new failure modes or effects into the Control Rod Drive System.

The possibility for an accident ofa differen type than any evaluated previously in the SARis not created. The proposed change does not create any new equipment interactions.

The possibility ofa malfunction ofequipment important to safety ofa differen type then evaluated previously in the SARis not created. The proposed change does not create any new equipment interactions.

The margin ofsafety as defined in the basis for any technical specification is not reduced by this change. Reactivity control or the abilityto drop the rods into the core(ifrequired) is unaffecte by this change.

SEV-1116 CHANGE MOV 7443 AND MOV 7444 FROM MOTOR ACTUATIONTO MANUALACTUATION MOV 7443 and MOV 7444 are motor operated containment leak test isolation valves. The valves do not require electrical actuation to perform their design function. Due to the increased benefit maintenance associated with motor operated valves, the added cost ofmaintaining the motor actuators on valves 7443 and 7444 has no fo Ginna Station and increases the competitive price of product.

Electrical power for MOV7443 and MOV7444 willbe removed perPCR98-012. The motors will be abandoned in place. Electrical cables, conduit and components will be removed as the practicable. Avalve handle willbeinstalled to allow manual actuation and ameans for locking the handle will be provided to prevent tampering and/or mispositioning.

The probability ofoccurrence ofan accident previouslyevaluated in the SARis not increased due to removal ofthe control power and position indication for valves 7443 and 7444. The valves are valveswil not individually evaluated inaccidentmitigation. The valves are currently maintained in a closed modification position above mode 5 with position indication provided on the Main Control Board. After the continue to be maintained in the 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 for valves 7443 and 7444. The valves are currently maintained in a closed position above mode 5 and this will not be changed by this modification. New locking valve handles will be installed to prevent mispositioning. Plant configuration and piping willnot be changed by this modification therefore the integrity ofthe containment boundary is unafFected.

The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR will not be increased due to removal of the control power and position indication for valves 7443 and 7444. The valves willbe placed in a locked closed configuration during operation above mode 5, which is consistent with current operational position. There is no change to the mechanical properties ofthevalves or piping therefore no new malfunctions arebeing added to the configuration.

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

The possibility ofan accident ofa difterent type than any evaluated previously in the SARis not created due to removal ofthe control power and position indication for valves 7443 and 7444.

The valves willbeprocedurally maintained in a locked closed position for operation above mode 5, the same position which the valves are currently positioned.

The possibility ofa malfunction ofequipment important to safety ofadi6erent type then evaluated previously in the SARis not created due to removal ofthe control power and position indication for valves 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 with the determination that the additional weight(approximately three pounds) is negligible and willnot cause a component failure during an earthquake. In addition, the handle will be welded to the stem therefore no seismic interactions exist.

The margin ofsafety as defined in the basis for any technical specification is not reduced due to removal ofthe control power and position indication for valves 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 in the control circuitry. This function willbe removed and the valves willbemaintained in a locked closed position above mode 5.

SEV-1117 INSTALLATIONOF SPENT FUEL STORAGE RACKS AND RELATED MODIFICATIONS TO SPENT FUEL POOL The proposed changes to the facility are as follows:

P (a) RemovethreeoldrackswithnoneutronabsorberthatcurrentlyconstituteRegion 1 of the 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 will be designated as the new Region 1. forremova currentlyidentified (c) Remove obstructionsasneeded. Obstructions are as follows:

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

Stubs attached to the liner (several stubs are located on the north, east and south walls of the spent fuel pool). These stubs will be 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. This willensure that the probability of puncturing the spent fuel pool liner is as low as reasonably achievable. In the event of a 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 of the pool to aid positioning of the spent fuel bridge during fuel shuBling. The X-YIndexing plates willbebolted to the top ofthe concrete wall surrounding the spent fuel pool, on the north and south sides. The area at the top ofthe wall is 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 of the bolts and the X-Y indexing.

(e) Relocate the support for the spent fuel handling tool further along theeastwall to a position closer to the south wall. This modification will entail removing the existing support for the spent fuel assembly handling tool that is welded to the liner and installing a new support that consists of a 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 of the 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 floor that is different from the final layout achieved after the end of the installation).

In general, temporary configurations are not explicitlydescribed in theNRC Safety Evaluation that (NRC SE) issued by the U S. NRC to RG&E on 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).

configurations This safety evaluation willprovidethe basis for determining that the conclusions in theNRC SE are bounding with respect to any ofthe possible temporaiy coul develop during the installation, and will also provide the basis that there are no additional unreviewed safety questions by implementing the modifications described above.

Removal of Old Racks and Installation of New 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 in the SAR and the NRC SE has not been increased.

This evaluation provides thebasis for determining that the consequences ofthe designbasis accidents documented in the NRC SE are bounding with respect 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 in the SAR and the NRC SE have not been afFected.

TheNRC SE outlines therequirements for movements 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 ofamalfunctionof equipmentimportant to safety previously evaluated in the SARremains unchanged.

TheNRC SE outlines therequirements for movements ofheavyloads during and after the installation. Theserequirements willbemet during the installation. There is no impact on the malfunction of equipment important to safety. Therefore, the consequences of a malfunction ofequipment important to safety previously evaluated in the SAR remain unchanged.

The NRC SE documents the evaluation ofdesign basis accidents applicable during and after the installation. This evaluation providesthebasis for determining that the design basis accidents documented in theNRC SEarebounding and still applicable with respect to any ofthe possible temporaryconfiguration that could develop during removal ofthe old racks and installation of the 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 for movements 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 of a 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 for determining that the evaluation configurationstht ofthe basis accidents documented in the NRC SE is bounding and still applicable with respect to any ofthe possibletemporary could develop during removal ofthe old racks'and installation ofthe new racks. Allregulatory requirements and limits set forth in the SAR, the NRC SE, and the Technical Specifications are met during the modification. Therefo're, the margin of safety as defined in the basis for any technical specification is not reduced.

Removal of Obstructions:

Training, procedures, and administrative controls are established to ensure that the probability ofpuncturing the spent fuel pool liner is as low as reasonably achievable. The probability of occurrence 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

probability ofpuncturing the spent fuel pool liner is as low as reasonablyachievable. The consequences of any potential breach of the liner during removal of obstructions are bounded by the consequences ofa hypothetical tornado missile puncturing the liner as documented in the tornado missile designbasis accident. The consequences ofaccidents previously evaluated in the SAR have not been affected.

Maintaining the structural integrity of the spent fuel pool liner does not impact the malfunction ofequipment related to safety. Therefore, the probability ofoccurrence ofa malfunction ofequipment important to safety previously evaluated in the SAR remains unchanged.

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

Any potential breach ofthe spent fuel pool liner during removal ofobstructions is bounded by the consequences ofa hypothetical tornado missile puncturing the liner as documented in the tornado missile design basis accident. The proposed modification does not introduce anew failure mode not documented in the SAR. Therefore, the possibility ofan accident of a di6erent type than any evaluated previously in the SAR is not created.

Maintaining the structural integrity of the 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 in the SARis not created.

Any potential breach ofthe spent fuel pool liner during removal ofobstructions is bounded by the consequences ofahypothetical tornado missile puncturing the liner as documented in the 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 probability ofoccurrenceof 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 in the 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-YIndexing plates 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 in the 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 possibility ofamalfunction 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 by installing the X-Y Indexing plates. Therefore, the margin ofsafety as defined in thebasis for any technical specification is not reduced.

Relocation of the Support for the Spent Fuel Handling Tool:

The design ofthe proposed support is similar to the existing one. The probability ofa drop ofthe spent fuel handling tool on the racks has remained unchanged. In the conservative direction, the tool support has been positioned further away from spent fuel racks. The probability ofoccurrence ofthe design basis accidents documented in the 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 by the consequences oftheFuel Handling Accident(FHA) documented in the NRC SE.

The consequences ofa drop ofthe tool support on top ofthe spent fuel racks are bounded by the consequences of the 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 fuel handlingtool and/or its support has no impact on the malfunction ofequipment important to safety. Therefore, the probabilityofoccurrence ofamalfunction ofequipment important to safety previously evaluated in the SAR remains unchanged.

The drop ofthe spent fuel handling tool and/or its support has no impact on the malfunction of equipment important to safety. Therefore, the consequences of a 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 in the spent fuel pool. These accidents are bounded by accidents documented in the NRC SE.

Theproposed modification does not introduce anew failure mode not analyzed in the SAR and the NRC SE. Therefore, the possibility of an 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 possibility ofamalfunction ofequipment important to safety ofadifFerent type than evaluated previously in the SARis not created.

The consequences ofa drop ofthe spent fuel handling tool on top ofthe spent fuel racks are bounded by the consequences ofthe Fuel Handling Accident(FHA) documented in the NRC SE. The consequences of a drop ofthe tool support on top of the racks are bounded by the consequences ofthe Tornado Missile Accident documented in theNRC SE. Therefore, the margin ofsafety as defined in the basis for any technical specification is not reduced.

SEV-1118 SEVERE ACCIDENT MANAGEMENTGUIDANCE SAMG IMPLEMENTATION The purpose ofthis change is to implement the Severe Accident Management Guidance(SAMG) at Ginna Station. The SAMGs are designed for use in extreme accident circumstances when the effective Plant EOPs are no longer 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 SAMG program, as well as the EOP transitions to the by SAMGs.

The SAMGs are not put into use until an 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 the 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 off exhausted, the consequences of a previously evaluated accident will not be increased.

The change addressed by this review simply establishes the administrative aspects ofthe SAMG affected 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 admiistrative aspects ofthe SAMGprogrambecause equipment operation or configuration is not addressed in staff these documents.

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

Ifa previously unevaluated accident should occur, the SAMGs should provide some guidance in dealing with the situation and thereby providing the plant wit a tool to perform their primary function of protecting the public.

The administrative aspects of the'SAMG program does not deal with equipment operation, configuration or functionality issues. Because the SAMGprogram does not result in any equipment design function changes, the possibility ofanunevaluated equipment failure is notincreased. The SAMG's do provide equipment lineups and operational suggestions but only after 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 ADJUSTABLE TRAVEL STOP SET POSITION FOR HCV-624 AND HCV-625 An adjustable valve travel stop willbe added to the actuators for RHR 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 or be 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 cavity fillinduring the 1999 outage. Following the setting ofeach valve actuator in position, the handwheel willbe chain locked in place. The modification willnot prevent the valves from being throttled in the closed direction. The valves'pen position willbe limited to a position less than full open as determined by analysis and be set during the flow test. No further adjustment of the valve is needed for any mode of operation.

The probability of occurrence ofan accident previously evaluated in the SAR is not increased, because the affected valves, HCV-624/625, do not change position following a postulated flow transient. They remain in their open position. The extent oftheir open position is being changed, and the new position willstill ensure the required low head safety injection flowfor the duration of the transient.

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

Providing alimiton system wil also ensure, under conditions resulting in maximuminjection flow, that RHRpump runout conditions do not exist. In the longer term, folio'wingswitchover to the sump recirculation phase, the modification provides a limitation on RHR flow rate, while assumingalossofinstrumentair,therebypreservingNPSHmargin. Therefore,corecoolingcan continue with no loss of function.

The probability of occurrence of a malfunction of equipment 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 following a 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 throttling is necessary using HCV-624/625, the travel stop willnot interfere with the throttling ofthe valves in the closed direction. There is currently no need to throttle the valves more open than the travel stop position will be set.

0 Administrative limits currently exist onRHR flo(1500 gpm) which limitthe low rate to a value less than the travel stop would allow.

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

Since the valves are normally maintained open while the plant is at power, there are no times when the fail open on loss ofinstrument air function would be called upon. There are no malfunctions that would cause the valves to fail closed since the spring in the actuator is a passive device not dependent on external controls, and the valves are routinely tested and calibrated. The travel stop cannot cause the valves to move in the closed direction, since its design only restrictsmotion in the upward direction. Therefore, LHSIflowrate willstill meet the COLR values and no increase in consequences can occur due to reduced core cooling assumed in the accident analysis.

The possibility ofan accident ofa different type than any evaluated previously in the SARis not created, because the travel stop does not interfere with the operation ofHCV-624/625 over the range of travel these valves are assumed to maintain.

The possibility ofamalfunction ofequipment important to safety ofa different type than evaluated previously in the SARis not created, because the travel stop is designed to provide suflicient flow to preserve LHSI capabilityunder thelimiting assumptions previously assumed, while limitingflow su6iciently to preserveRHRpump NPSHmarginduring the sump recirculation phase. The valve actuator handwheel willbelocked in place so that manually opening the valvemore thanits setpoint cannot be inadvertently performed..Operationofthevalve fiomthe control room and operators use of the valves will be unaffected.

The margin ofsafetyas define in the basis for any 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 of COLR flow requirements in the limiting case.

SEV-1120 REMOVALOF DEWPOINT MEASURING INSTRUMENTATIONFROM THE SEISMIC AND METEOROLOGICAL INSTRUMENTATIONSYSTEM The Ginna Station Seismic and Meteorological Instrumentation System (SMI) is made up of a 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 in which the dewpoint transmitter is required to operate (increased frequency of airborne dirt particles due to a fairly constant breeze) is not conducive to the sensitivity ofthe dewpoint transmitter. The dewpoint transmitter senses humidity via 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. In addition, the data gathered by the dewpoint monitoring system is not used in the decision process for mitigation of accidents 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 of an accident previously evaluated in chapter 15 of the 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 of all equipment used for the 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 that the proposed modification can not introduce a failure mechanism whichwould increase the consequences ofan accident previouslyevaluated in chapter 15 ofthe UFSAR. The modification does not impact or increase the calculated radiological dose to the general public for any event evaluated in theUFSAR. 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 functionally unrelated 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 in chapter 15 of the UFSAR. The modification will not degrade the performance of the 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 of all equipment used for the mitigation ofaccidents and transients. The independence ofthe dewpoint measuring system fiom any System, Structure or Component important to the mitigation of modification accidents and transients ensures the proposed wil not introduce a failuremechanism which would increase the consequences of a malfunction of equipment important to safety previously evaluated in chapter 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 for the 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 that the proposed modification will not introduce a failure mechanism which would increase the probability of an accident of a diFerent type than any will previously evaluated in chapter 15 of the UFSAR. There are no adverse affects upon other the systems, nor any new failure modes induced.

The dewpoint monitoring system is functionally unrelated and physically independent of all does equipment used for the mitigation ofaccidents and transients. The modification notintroduc independence ofthe dewpoint measuring system from any System, Structure or Component important to the mitigation of not notdegrade accidents and transients ensures the proposed a failure mechanism which would increase the consequences ofa malfunction ofequipment important to safety ofa different type than previouslyevaluated in the UFSAR. The power source for the Ginna Station will modification 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) itself willnot afFect equipment important to safety. The meteorolgical monitoring system.

The dewpoint monitoring system is functionally unrelated and physically independent of all modification equipmentused for the 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 intoduce a failure mechanism which would reduce any margin ofsafety as defined in the basis ofany Technical Specifications.

The required functions and characteristics of the Ginna Station Seismic and Meteorological .

Instrumentation System remain unchanged.

SEV-1121 PCN 8 98-4517 SAFETY EVALUATION CHANGES TO ATT-2.1 ATTACHMENTMIN SW TO ADDRESS ACTION REPORT 98-1042 CONCERNS ACTIONReport98-1042 identified a concern with guidance provided inRevision 4 ofprocedure ATT-2.1, "ATTACHMENTMNSW". This attachment is used to align the service water system for the recirculation phase of a LOCA with one operable SW pump. ATT-2.1 instructs the operators to fully open the service water globe valve on the discharge side of the CCW heat exchanger to be aligned (V-4619 or V-4620).

During a reconstitution ofthe service water system hydraulic model, an error was found in the hydraulic loss coefficient used to represent each CCWheat exchanger. The coefBcientused in the 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 originally modeled, the servicewater flowrate to the applicable CCW heat exchanger would be considerably higher than originally predicted if V-4619 or V-4620 were opened completely with a single service pump in service. As a result, the flow rate 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 of ATT-2.1 are proposed:

Delete step 3 which has operations request that the TSC evaluate isolation ofSW loads in containment. The step willbe replaced with explicit instructions 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 of the Bus 17-18 cross-tie and startup of a second service water pump. This change is considered inconsequential and willnot be addressed in this safety evaluation.

Breakout the step that isolates service waterto the SFP heat exchangers and place it prior 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 HXB on the FIA-2005 reading, which is used to set V-4619 or V-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 of this step has no implications on the timing ofthe transfer to recirculation. Therefore, relocation ofthis step is considered inconsequential and will not be addres'sed in this safety evaluation.

Modifystep 5 to throttle the SW outlet valve on the operating AuxiliaryBuilding service 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 flowto the CCW heat exchanger and to reduce load or refer to ER-D/G.2, ALTERNATECOOLING FOR EMERGENCY D/Gs, should an EDG temperature alarm occur. This change is considered inconsequential and willnot be addressed in this safety evaluation.

Add new step to notify TSC of all loads that were isolated. This change is considered inconsequential and will not be addressed in this safety evaluation.

Implementation ofthese steps willaddress the issue raised in ACTIONReport 98-1042.

Additionally, explicit isolation ofinoperable containmentbuilding loads willincrease the heat removal rates from containment, increase the margin to vapor locking in the CRFCs due to flashing in the downstream service water piping, and provide increased service water flow rates to the EDG coolers.

thatthe have TheproposedprocedurechangesapplyduringtherecirculationphaseofaLOCAonly. Theywill not increase the probability ofoccurrenceofan accident previously evaluate'd in the SARsince the changes accident will have already occurred prior to usage of the procedure.

The proposed procedure changes willnot increase the consequences ofan accident previously adequatejustification evaluated in the SAR. The analyses outlined in the functional impact section of this safety evaluation provide ony positive sects with respect withrespec to the capability to deal with and the consequences of a LOCA, which is the only impacted arebeneficial accident.

Theproposed changes to equipment reliability during the recirculation phase ofaLOCA and therefore it is reasonable to conclude that the probability ofoccurrence of a malfunction of equipment important to safety is reduced. Specifically:

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

The margin to vapor lockingin the CRFCs is increased due to the increase in service water flow through the CRFC.coolers. This is also beneficial with respect to cooling the

atmosphere in containment.

The margin to runout of the. 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 for post-accident conditions are met. Isolation ofinactive containment loads marginallybeneficialin increases overall containmentheatremoval so is term ofequipment reliability.

The consequences of a malfunction of equipment 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 failure willalready have occurred prior to usage ofATT-2.1, since this is required to get to one service water pump operation; therefore, any additional failure will be beyond the design basis of the plant.

The consequences of any failure that results in the loss of the sump heat sink willbe beneficiallyincreases reduced. Loss ofthis heat sink willresult in an increase in containment temperature; the increased flowto the CRFCs provided by the proposed changes 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 originally operating, 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 of active heat sinks.

The consequences ofthe loss ofthe operating service water pump are no more severe than they would beifthe proposed changes were not implemented sincethe end result in either case is a complete loss of active heat sinks.

The possibility ofan accident ofa different type than any evaluated previously in the 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 prior to 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 of the 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 prior to the transfer to recirculation. Further, it has been shown that the proposed changes do not challenge the profiles assumed for equipment qualification. Since no margins of safety have been challenged, the margin of safety as defined in the basis for any technical specification will not be challenged.

C

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

Based on sampling and testing, it has been determined that some leakage is occurring from the transfer canal.

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

A sudden increase in SFP Liner leakage would be the accident/event of concern which, is not presently addressed in the UFSAR accident analysis. Therefore, the probability ofoccurrence 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 of an accident previously evaluated in the SA is not increased.

The equipment in question would be the RHR and RCDT Pumps in the AuxiliaryBuilding sub Basement. The suspected leakage ofSFP water into the RHRroom is believed tobe originating through incomplete or defective seal welds ofthe liner to the embedment ofthe refueling canal.

Should there be a complete failure ofthesewelds, anunrestricted flow ofthe canal inventory into the RHR room is precluded by the concrete/bedrock interface. In addition, the increased frequency ofAuxiliaryBuilding sump Pump actuations would alert the Operators, providing an opportunity to take corrective actions. Therefore, there is no increase in the probability of occurrence of a malfunction of equipment previously evaluated in the SA.

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

The flowpath described above, does not lend itselfto a rapid outflow ofwater &omthe SFP. The leakage ofborated water into the RHRroom has beendetermined to be from the 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 restrictionofflowwould allow ample time for mitigating actions outflow such as installing the weir gate, closing ofthe transfer tube gate valve and/or draining the transfer canal. Even in the unlikely event ofa rapid ofwate from a failure in this area, the height of the weir gate path would preclude the uncovering of spent fuel in the pit. Therefore, the possibility of an accident of a different type than that evaluated in the UFSAR is not created.

Based on the above discussions of leak rates the operability of the RHR/RCDT pumps is not jeopardized by this condition. Therefore, the possibility ofa malfunction ofequipment important to safety ofa different type than previously evaluated in the SA is not created by this condition.

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

SEV-1124 VALVESTEM PACKING IMPROVEMENT PROGRAM SAFETY EVALUATION CHANGES TO DESIGN CRITERIA - EWR 4859 This SafetyEvaluationwasprepared to replaceRevision1 to the Safety Analysis (Revision 1 was never approved) for EWR 4859 to evaluate the addition of Expandable Valve Stem Packing (EVSP) of the "cup and cone" design as a packing system alternative.

This analysis covers the live-loading ofgland followers and/or replacement ofvalve stempacking of certain 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 in terms 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 with the valve steam packing leakage. These improvements are:

Replacement of traditional woven asbestos packing with die-formed square flexible graphite packing or Expandable Valve Stem Packing(EVSP) ofthe "cup and cone" design.

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

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

The proposed modification would not increase the probability of occurrence of an accident previously evaluated in the UFSAR since this change only allows replacing approved packing materials and methods with improved alternatives thatwillreducethe potential for packing leakage.

The proposed modification would not increase the consequences of an accident previously evaluated in theUFSAR since the expandable valve packing is an improvement in valve 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 criteria for thereplacement and upgrade ofpacking materials and methodologyin safety-related valves resulting in an increase in reliability for affected valve operation.

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

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

specification The proposed modification would not reduce any margin ofsafety as defined in the basis ofany technical 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 in ACTION 97-1110. The batteries have not degraded to the point where discharge testing indicates 'eport 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 for the 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.

orreplacent 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.

An initial evaluation determined no costbenefitbetweenmodification The added size and weight ofthenewbattery cellswould require modification ofthe existing racks.

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 of the Control Building.

The probability ofoccurrence ofan accident previously evaluated in the 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 in the 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 will not result in the occurrence of an accident. Other than the batteries, intercell connectors and racks, no new equipment is being added. No existing equipment needs tobe functionally modifie as a result of this proposed change.

The consequences of an accident previously evaluated in the SAR is not increased by this fi modification. The newbatteries have a greater capacity than the existing batteries therefore they have the ability to mitigate any design basis events which the 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 will not increase the consequences of any accident previously evaluated.

ornaqualifi The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SARis not increased by this modification. The newbatteries arebeing purchased as Class-1E equipment supplier to the design conditions oftheBattery Rooms. The new. battery racks are being purchased from the same supplier with seismic qualifications to the requirements of Ginna'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 float voltage willbe set at a new higher value in order to minimize the amount ofequalize charges which have to be performed on the batteries. There is no increase in the 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 in the SAR is not increased by this modification. There is no increase in the consequences of a If malfunction ofany equipment due to this modification. one ofthe batteries were to fail there would be no increase in the existing consequences resulting from the loss of a battery.

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

The possibility ofan accident ofa differen type than any evaluated previously in the SARis not created by this modificatio. No new equipment is being added to the DC distribution system due to this modification, therefore there is no potential of an accident of a different type than any previously evaluated. The batteries arebeing replaced with new cells with increased capacity and new seismic racks are being installed. The new batteries are functionally equivalent to the existing batteries and no new failure modes will be introduced due to this modification.

will The possibility ofamalfunction ofequipment important to safetyofadifferent type then evaluated installne modification previously in the SARis not created by this modificatio. This Class-1E battery cells and new seismic battery racks to replace existing equipment. The new batteries willbelead calcium, which is different than 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 for the batteries or racks.

Evaluations ofthe electrical system, HVAC system and fire loading demonstrate that the new batteries willbe capable ofoperation without impacting the operability'of any systems supporting the Battery Rooms or the DC distribution system.

The margin ofsafety as defined in the basis for any technical specification is not reduced by this modification. This modification willinstall new seismic battery racks designed for the 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 TIME DELAYRELAYS During safety injection, ifa single safeguards bushas anundervoltage actuation, either degraded or loss ofvoltage, its supplybreaker willtrip and the sister bus supplybreaker on that train willalso trip. Each bus willstart a 1.3 second timer that upon timing out closes the bus diesel generator (DG) supply breaker. The sister bus UV systemwill not actuate since the minimumtime required is the loss of voltage relay definite time delay of 2.4 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. It is also possible to load two service water motors onto the diesel generator. This would exceed design loading during SI.

modificationwillinstal This time delay relays in the 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 circuit shall actuate the respective bus UVsystemupon coincident opening ofboth the normal and diesel generator supply breakers. Logic shall allow for live bus transfers and bus restoration using bus tie breakers.

The new configuration will operate as follows:

Upon coincident open normal and diesel generator supply breakers to a bus, the new relays begin timing. After 0.5 seconds the first relay'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 UV system resets. Ifthe existing 1.3 second delay relay times out and the DG supply 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.

Allows operators 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 UV system actuation.)

The probability ofoccurrence ofan accident previously evaluated in the 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 will continue to have the independence" and separation which they are required to have to perform their safety related functions.

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

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

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

time a

configuration ofa loss ofa 480 volt bus delay The proposed control safeguards train during an SI. However, it increases the frequency per reactor year of an willinstall inoperable safeguards bus. The net change in frequency ofa safeguards bus loss is a decrease of two orders of magnitude, lE-4 decrease versus 1.91E-6 increase. Therefore the proposed configuration increases overall plant safety.

modification The consequences of an accident previously evaluated in the SAR is not increased by this modification. This elays in the control circuits ofthe diesel generator supply breakers. Any failure ofthe breakers to actuate due to the new relays willbe bounded by previously assumed failures of the 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 which willincrease the consequences ofany previously evaluated accidents.

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

The consequences ofa malfunction ofequipment important to safety previously evaluated in the 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 in the existing consequences.

The possibility ofan accident ofa different type than any evaluated previously in the SARis not created by this 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 within the control circuits ofthe diesel generatorbus supplybreakers, in parallel with existing time delay relays. The function performed by thebreakers is not being changed.

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

The margin ofsafety as defined in the basis for any technical specification is not reduced by this modification. This modification will install 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 of safety.

SEV-1128 SERVICE AIR SYSTEM UPGRADE PHASE B The scope ofthis modification is to replace the existing Service Air Compressor 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 of a cross-connect the between the instrument air system and the service air system downstream ofthe new dryer. A check valvefiomdegradin and an automatic isolation valve on the service air side ofthe cross-connect will function to prevent service air backflow into the instrument air systemand prevent a loss ofservice the in air pressure the instrument air system. Although no system or component name realisticallyreconfigures changes are involved, this proposed change instrumet air system into a four compressor system which takes advantage ofits increased compressor capacity to supply air to the ser vice air system. reliability the modificationwillimprove This fboth the service air and instrument air systems and fire thereby aid in meeting theNRC requirements oftheMaintenanceRule. The change installs avery dependable backup to the instrument air systemaswell providing a low maintenance source ofair capacity for use as service air.

This modification also includes reconfiguring door F28 such that itcanbe leftopen during periods turbinebuildin ofhigh ambient temperature in the turbine building. The door closer willbe redesigned so that a fusible linkwillallow automatic closure in the event ofa or all volatile effects treatment room.

The instrument and ser vice air systems have no failure modes or effects which are precursors to accidents evaluated in the SAR. The proposed change does not introduce any newfailure modes or t the 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 probability ofan accident previously evaluated in the SAR.

Pneumatically operated components required for accident 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 with the backup systems thus the change cannot cause any equipment failures which would reduce the availability ofequipment 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 of an accident.

As described the proposed change provides isolation between the service and instrument air systems. After the change the instrument air system willhave better capacity and equipment reliability then before the change. It can be concluded that the proposed change reduces the

probability of occurrence of a 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 of a loss of instrument 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 it can not create the possibility of an accident of a different type the previously reviewed.

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

.accident mitigation. Because the proposed change only interacts with non-safety equipment it can be concluded that the change does not increase the possibility of a malfunction of equipment 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 of safety addressed in the technical specifications.

SEV-1129 CONTROL ROOM HVAC UPGRADE PHASE 1 for The purpose ofthis evaluation is to determine ifan unreviewed safety question exist with the planned phase 1 modification to upgrade the reliability ofthe Control Room HVACsystem by filtrationtrain providing a redundant th ControlRoomEmergency AirTreatment Sub-system (CREATS). Phase 1 of the modification consist of:

1) Adding tie in connections to theexisting CRHVACductworkwithblindflanges to allow future connection of a second charcoal filter train.

2) Replacing MCC K spare breaker 1KK with an upgraded breaker

3) Swapping MCCKDC control power from train B to train A, removing 4KVtest cabinet load from Train AMainDC distribution panel and placing it on the Turbine Building DC distribution panel.

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

5) Replace degraded CRHVAC flex duct connectors

6) Connect a spare cable to one ofthe spare contacts on SI relays SI-16X and SI-26X for 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 in half and adding a support member on the Aux Bench Board in the ControlRoom to allow future on-line replacement for new trainB controls.

Theprobability ofoccurrence ofanaccidentpreviouslyevaluatedin the SARis not increased since the source terms during a LOCAwillnot be increased by this modification, it does not affect the RCS, Containment Filtration or the ECCS. The reliability ofthe AC and DC systems willnot be decreased since the electrical changes meet the original plant 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 will not be impacted.

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

proposed change since theperformance ofthe CREATS system, AC and 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 of occurrence of a malfunction of equipment important to safety previously evaluated in the SARis not increased since the new backdraft damper is similar in construction to existing active dampers in the system and therefore willhave the same type offailure mechanisms and probabilities. The new breaker, cables and flexible duct connectors are the same or equal to the original equipment theare replacing and therefore their probability ofoccurrence or malfunction willnot be increased over what was originally assumed. No new equipment is added to the SI system.

The consequences ofa malfunction ofequipment important to safety previously evaluated in the SARis not increased. Afailureofthe CREATS system filtering abilitywas already assumed in the original accident analysis therefore the addition of the backdraR damper will not 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 ability to cope with a loss ofall AC for four hours, 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.

fission product The possibility ofan accident ofa different type than any evaluated previously in the SARis not created. Theproposed change creates no new functional interactions with existing plant equipment nor does it introduce newfailure modes or mechanisms which could lead to reactor core damage configuration or releas. No new equipment is added to the SI system. The SI contact testing will not change any system 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. Train A train AMCCloads and there are no trainB loads onMCC DC power is already used willall otherrequirementsan fireprotection K. The loss of MCC K has already been considered and therefore, the use of train B control power is already bounded by the existing anaysis. The new spare SI cable will meet exist seperation criteria and is therefore bounded by existing analysis.

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

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

SEV-1130 NEW PROCEDURE AP-CVCS.3 LOSS OF ALLCHARGING FLOW New procedure AP-CVCS.3, Loss ofAllCharging Flow, has been developed to deal with the unique problems associated with that event. A loss of charging takes away the ability for RCS makeup atnormal system pressure. Even ifCVCS Letdown is isolated, the RCS continues to lose inventory through the RCP seals. The RCP seals are protected &omhigh temperature conditions by CCW flowthroughthe thermal barrier. However, ifaction is not taken, theRCS willcontinue

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

Attempt recovery of Charging 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 allow for RCS depressurization. This is ashutdownmargin issue for single 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 of an accident previously evaluated in the 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 of the CCW System ensures RCP seal integrity.

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

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The probability of occurrence of a malfunction of equipment 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 of charging). Other equipment is operated within previously established parameters, with the exception ofthepressure/temperature limitcurve for RCP NPSH. Although not recommended for routine 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 in the SARis not increased because the new procedure addresses this type ofevent without impacting the failure consequences of other plant equipment. which The possibility ofan accident ofa di6erent type than any evaluated previously in the SARis not created because the intended function and operation ofplant systems is not afFected. In addition, configuration the plant is not placed in a i not 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 defined operational limits. For example, inthebasi the operator is directed to observe starting duty ofSafetyInjection Pumps when cycling the pumps to stabilize RCS inventory.

The margin ofsafetyas for any technical specification is not reduced because, .

although a loss ofall charging may result in entry 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 for Cycle 28 consists of44 new fuel assemblies labeled as feed regions 30A and 30B.

This safety evaluation is valid for an End-of-Cycle 27 burnup of 15,509 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 in the core for Cycle 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 with Zircaloy-4, ZIRLO has no Chromium(Cr) and reduced contents of Tin(Sn) and Iron(Fe). In addition, ZIRLO has incorporated anominal amount ofNiobium(Ni). The purpose of changing the chemical composition ofZircaloy-4 to that of the ZIRLO alloy is to improve corrosion resistance and dimensional stability under irradiation.

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

The Cycle 28 reload willnot increase the probability of occurrence of an accident previously evaluated in the 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 probability ofan accident previously evaluated in the SAR.

The change to ZIRLO as the material for fuel rod cladding, guide tubes, and instrumentation tubes, is not directly related to the probability ofany accidentpreviously evaluated in the UFSAR. The use of ZIRLO as a material does not impact the mechanical integrity of the fuel rod, or the structural integrity ofthe fuel assembly or the coreunder normal or accident conditions. Alldesign criteria, applicable standards, and safety limits are met. Because of this, there are no new challenges to components and systems that would increase the probability of any 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 probability offailure 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 ZIRLO material 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 limits for the fuel rod cladding, guide tubes, and instrumentation tubes using ZIRLO material are met. Therefore, mechanical and structural integrity ofthe fuel assemblies willbe maintained under normal and accident conditions.

An analysis ofall fuel with gap reopening demonstrates compliancewiththecorrosion limit set forth in 10 CFR 50.46 for all of Cycle 28.

The radiological consequences ofaccidents with fuel assembliesusing ZIRLO material are the same as those documented in the UFSAR for fuel assemblies using.

Based on the above, the radiological consequences ofaccidents previously evaluated in the 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 in accident analyses, and (b) the design parameters are still within the assumed ranges.

The required SDMis met for all power levels and at anytime during the core lifewith the control rods above the insertion limits in the COLR.

The fuel assemblies ofthe Cycle 28 reload with ZIRLO material meet the same design criteria, applicable standards, and safety limits as those fuel assemblies with Zircaloy-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 limits are met for the fuel assemblies with fuel rod cladding, guide tubes, and instrumentation tubes fabricated with ZIRLO material. Meeting design criteria and applicable standardsprecludes new challenges to components and systems that could increase theprobability ofmalfunction. No new failure modes or limitingsingle failures have

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

Gap reopening is not expected to lead to fuel failure. Violating the gap reopening SAFDL criteria 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 with ZIRLO material. Meeting design criteria and applicable standards precludes new challenges to components and systems and/or a challenge to the integrity ofthe fuel rod cladding. No new failure modes or limitingsingle 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 of equipment malfunction.

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

Alldesign criteria, applicable standards, and safety limits are met for the fuel assemblies with fuel rod cladding, guide tubes, and instrumentation tubes fabricated with ZIRLO material. Meeting design criteria and applicable standards precludes new challenges to components and systems and/or a challenge to the integrity ofthe fuel rod cladding. No new failure modes or limitingsingle failures have been created.

Previous analyses assumed no gap reopening for simplicity. Analyses with gap reopening show acceptable consequences. This condition is acceptable provided that continued compliance with the 17% oxidation limitis maintained. This compliance has been confirme with a 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 in the basis for any technical specificationbecause itmeets all design criteria, applicable standards, and safety limits set forth in 44-

the licensing basis.

With respect 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 is to determine ifanunreviewed safety question exist with reducing the minimum temperature assumed in accident analysis for auxiliary feedwater from 50'F to 32'F.

The only components and systems affected by the proposed change are the Condensate Storage Tanks (CSTs) and AuxiliaryFeedwater System(~ whichprovide auxiliary feedwater to the steam generators(SGs). Normally, this is during plant startup/shutdown or followinga reactor trip.

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

Implementation of this change does not increase the probability of occurrence of an 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 for the 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 in that the consequences meet the required acceptance criteria; thus the consequences are acceptable.

Implementation ofthis change does not increase the probability ofoccurrence ofa malfunction of equipment important to safetyin that the temperature reduction does not cause equipment or piping 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 in that the change does not impact the capability to meet the accident analysis nor does it adversely impact the ability ofany 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 in that the accidents afFected still meet the required acceptance criteria. Since the acceptance criteria are met there is no reduction in the margin of safety.

SEV-1134 REACTOR 1NTERNALS BAFFLE BOLT REPLACEMENT The purpose ofthis safety evaluation is to examine the acceptability of 1999 baEe-former-bolt replacement strategy.

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

TheNuclear Regulatory Commission(NRC) has evaluated the acceptability ofperforming baffle-bolt replacements under the auspices of 10CFR50.59. (NRC Safety Evaluation ofTopicalReport WCAP-15029, "Westinghouse Methodology for Evaluating the Acceptability ofBafHe-Former-Barrel Bolting Distributions Under Faulted Load Conditions", TAC No: MA1152.)

In the review ofWCAP -15029 theNRC concluded that the general methodology presented was acceptable provided that: 1) the limitingbaftle bolt loading willbe determined by analysis for a class ofplants and a specific break; 2) the noding to be used in the 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 for the accounting oflimitations in the inspection methods used to detect flaws.

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

classified RG&E performed ultrasonictesting ofall theinspectablebolts. This examination identifie 59 bolts with flawindications. The replacement program changed 40 ofthesebolts along with 9 bolts that could notbe tested and 7 bolts a suspect. All 19 bolts with defect indications inFormer Plate levels 2 through 6 were replaced regardless oftheir location within the WOG pattern. All bolts that exhibited defects in Former levels 1 and 7 were left in service. Of the 6 bolts which exhibited flaws in positions required to conform to the WOG pattern in levels 1 and 7, sufficient adjacent bolts outside ofthe patternwere verified to be present to structurally compensate for the potential defects. The RG&E pattern of analysis thus incorporated both new and aged, but verified acceptable, bolts. The end result ofthis methodology yielded an output which provides reasonable assurance that all the 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 will function as required under faulted load conditions, as well as comprehensively addressing theNRCs criteria for making the changeunder 10CFR50.59, as delineated in their review of the 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 of occurrence of a accident previously evaluated in the SAR.

After completion 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 its function in accordance with the design requirements the consequences of an accident previously evaluated in the SAR willbe unchanged.

The baffl-formers physically and functionally interact with the reactor core barrel and provide lateral fuel support and reactor coolant flowdirection. Thebaffle-former-bolting replacement does not add any new functional interactions with configuration demonstratestha equipmentimportant to safety. The analysis ofthe as-leftbaffle-former-bolting 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 ability to cool the fuel following an accident.

Accordingly, it can 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 integrity to 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. After the 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 in the SAR is unchanged.

Thebaffle-former-bolting replacement program does not alter the design, fit, form, or function of 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 for an accident ofa different type then previously evaluated in the SAR.

Thebafile-former-bolting replacement program does not alter the design, fit, form, or function of the reactor internals. There are no new functional interactions created by bolt replacements. The analysis ofthe as-leftbafHe-former-bolting configurationdemonstrates that theunits have suf5cient integrity to resist all credible design basis loading conditions. AAer the 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 ability to cool the fuel 49-

followingan accident. Aftertheba61e-bolt replacements the reactor core internals willrespond to the stress of accidents and malfunctions the same way as before the change. This change does not introduce any new design or functional interactions therefore it cannotpossibly introduce the potential for a malfunction of a 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 it can be concluded that the margins ofsafety presumed in the bases for technical specifications will not be reduced.

10CFR50.59 SAFETY REVIEW FOR PCN 98-3013 TO PROVIDE GUIDANCE ON THE TWO OPERATOR CONCUIUMNCE RULE RG8cE responses to IEBulletinNo.79-06A, REVIEW OF OPERATIONALERRORS AND SYSTEM MISALIGNMENTS IDENTIFIED DURING THE THREE MILE ISLAND INCIDENT, state that administrative procedures were modified to require that two licensed modifies overriding action is executed on any safeguard operators shall agree on any overriding before thethatcommitmen system active component. The change to now require that the action be approved by the SRO that has the command function. This change still meets 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 will result in the procedural direction remaining in compliance with Technical Specification 5.1.2, which states that the Shift Supervisor or another SRO in his absence shall be responsible for the control room command function.

The change willnot alter the description in the UFSAR which states that the Shift Supervisor is responsible for the performance ofall personnel assigned to his shift who could afFect plant safety, regardless ofspecialty afhliation. The UFSAR also states that the operating shift crews 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 &om a license condition or a Technical Specification in an emergency when, as aminimum, it is approved by a licensed senior operator.