ML18017A264

Forwards Summary Rept of Safety Evaluations Approved During Period from 911231 to 920101 for Susquehanna Ses Units 1 & 2,per 10CFR50.59

ML18017A264
Person / Time
Site:	Susquehanna
Issue date:	06/04/1992
From:	Keiser H PENNSYLVANIA POWER & LIGHT CO.
To:	Martin T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
PLA-3757, NUDOCS 9206150082
Download: ML18017A264 (229)

Text

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"ACCELERATED STRIBUTION DEMONST TION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:.9206150082 DOC.DATE: 92/06/04 NOTARIZED: NO DOCKET FACIL:50-387 Susquehanna Steam Electric Station, Unit 1, Pennsylva 05000387 50-388 Susquehanna Steam Electric Station, Unit 2, Pennsylva 05000388 AUTH. NAME AUTHOR AFFILIATION KEISER,H.W. Pennsylvania Power & Light Co.

RECIP.NAME RECIPIENT AFFILIATION MARTIN,T.T. Region 1 (Post 820201)

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SUBJECT:

Forwards summary rept of safety evaluations approved during period from 911231 to 910101 for Susquehanna SES Units 1 &

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2,per 10CFR50.59.

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DISTRISOTION CODE: IE47D COPIES RECEIVED:LTR

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SIZE: (0 D

• 05000387 05000388 A

RECIPIENT COPIES RECIPIENT COPIES D ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-2 LA 1 0 PD1-2 PD 5 5 D RALEIGHPJ. 1 0 iNTERNAL: ACRS 6 6 AEOD/DOA 1 1 AEOD/DSP/TPAB 1 1 N DLP LHFB11 1 1 NRR/DOEA/OEABll 1 1 REG FILE 02 1 1 RGN1 FILE 01 1 1 EXTERNAL: NRC PDR 1 1 E NSIC 1 1 NOTES: 1 1 D

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NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASIE! CONTACT THE DOCVMENI'ONTROL DESK ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DIFRtlBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 22 ENCL 20

Pennsylvania Power 8 Light Company Two North Ninth Street ~Allentown, PA 18101-1179 ~ 215/774-5151 Harold W. Keiser Senior Vice President-Nuc/ear 215/7744194 Submitted pursuant to J

10CFR50.59 Juwa

>Os'r.

Thomas T. Martin Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 SUSQUEHANNA STEAM ELECTRIC STATION 10CFR50.59

SUMMARY

REPORT - 1991 PL A-3757 FILE R41-2A

Dear Mr. Martin:

Pursuant to 10CFR50.59(b), enclosed please find a summary report of the safety evaluations approved during the period from January 1, 1991 to December 31, 1991 for Susquehanna SES Units 1 and 2.

The report format is as follows:

SER No.- Unique number for'ach safety evaluation.

Cross Reference- Reference to the document for which the safety evaluation was prepared.

Description Change- A brief description of the change made to procedures, equipment or tests.

Summary- A summary of the three requirements for determining an unreviewed safety question as deftned in lOCFR50.59(a) P).

If you have any questions, please contact J. B. Wesner at (215) 774-7911.

Ve y yours, H'.W. iser Enclosure 920b150082 920b04 PDR ADOCK OS000387 R PDR

FILE R41-2A PLA-3757 Mr. Thomas T. Martin CC: (original)

Mr. J. J. Raleigh, NRC Project Manager Mr. G. S. Barber, NRC Sr. Resident Inspector

fgR gg.: 91-001

, R REFERE E: DCP 90-3002A, Rev. 0 F HA E r -9206150082 Install pipe taps to connect Reactor Building Outage Chiller.

R:

I. No. This modification does not affect the function of the safety-related components of the Reactor Building Chilled Water System. Effects on building design temperature limits are minimal since work was performed during the spring when the outside air temperatures were low. Partial cooling inside the Reactor Building was provided by running reactor building ventilation fans.

II. No. This modification installed piping taps and valves which do not alter the intended function of the Reactor Building Chilled Water System. It provides non-outage connection to a Reactor Building Outage Chiller.

IH. No; This modification neither reduces the operability of any equipment nor requires any additional surveillance to ensure plant safety; this modification improves drywell cooling capability during an outage.

$ ER~.: 91-002 R REFE E: DCP 90-3011, Rev. 0 DE RI F HA E:

Replacing the existing HPCI test valve internals with specially designed throttling internals, and resizing the downstream orifice.

R:

I. No. The modification does not affect the ability of the valve to close as required to support HPCI injection to the RPV. The ability of the HPCI system to operate as evaluated in the FSAR will not be affected.

II. No. This modification will enhance the ability of the HPCI full flow test valve to perform its testing function during the low pressure surveillance test. The new .

version of the friction element trim enhances the throttling capability of the valve and has no adverse effect on the valve's closing ability.

III. No. The safety related function of the HPCI valve is to remain in its normally closed position or actuate to the closed position if the valve was in the throttled open position during the required surveillance test. In either case, the margin of safety is unaffected since the new internals have no effect on the ability of the valve to close.

0

$ER l~.: 91-003 R RE RE: DCP 90-3049, Rev. 0 Install a highly accurate ultrasonic fiow monitoring system for. both of the Unit 2, 96 inch pipes supplying circulating water to the condenser.

I. No. This action does not materially affect the design featuies listed in the FSAR.

There is no quantifiable incrc'ise in the probability of occurrence or the consequences of an accident or malfunction of equipment related to safety.

II. No. No -features of the design or the location for installation have been identified which indicate the existence of any mechanism that would create the possibility of an accident or malfunction of a different type than previously analyzed in the FSAR.

III. No. This modification affects the circulating water system which has no safety-related function. Flow measurement of circulating water has no direct interface with any safety function in the plant.

$ ER~.: 91-004 R REFE N E: OP-124-001, Rev. 0 DE RlPTI F HA E:

RBCCW System Operating Temperature Reduction.

SDSKAilY:

I. No. RBCCW has no safety related function; it services heat loads vital for continued plant operation but services no equipment necessary to safely shut down the unit.

II. No. This evolution involves slowly lowering present RBCCW temperature by 10 degrees F (maximum) temporarily. Its effect will not be felt on vital plant operating equipment such as offgas and CIG which automatically regulates RBCCW flow to maintain system operating constant.

III. No. This evolution will not disrupt continued operation of any system supported by RBCCW.

0

~BR~.: 91-005

DCP 87-7032, Rev. 0 Convert the main turbine control valve operating mode from full arc to partial arc admission.

SIMLAAR ':

I. No. The modification does not change the performance of any safety system since the EHC system'is not a control system required for safety, and post-modification start-up testing will be done to verify that the control system responds as originally designed and tested.

II. No. There is no change in the design function of the control valves other than the stroke time. The overall effect is a less severe pressurization transient and the existing analysis conservatively bounds the change.

HI. No. The generator load rejection analysis is one of the bases for the "MCPR" limit.

As stated above, the generator load rejection transient analysis conservatively bounds the partial arc mode of operation.

$8R~.: 91-006

DCP 87-9032, Rev. 1 Convert Charcoal Filter Deluge Systems to Manual Systems and rewire flow switch alarm circuitry.

I. No. The deluge system configuration would be identical to that which exists today.

The only affect this modification has is that the control room will not be automatically notified if one of the outside stem and yoke valves was moved to the open position inadvertently.

Modifying the flow switch circuitry alarm priority will provide detailed information to the control room operator on fire protection system actuation.

II.

~ No. ~ The smoldering nature of a charcoal filter fire will allow adequate time for an operator to verify a fire condition in the filter and operate the deluge system manually without deleterious affects on the affected filter train or the redundant filter train.

III. No. Changing the deluge systems for the charcoal filters on the CREOASS system will not affect the acceptance limits for the system because this change will not affect the operation of the system.

Changing the flow switch circuitry to provide a "fire" alarm at the control room fire protection panel rather than a "trouble" alarm will not affect the operability of the fire suppression systems that provide suppression for safety related equipment nor affect the commitments regarding the fire protection panel in the control room.

/ERIN.: 91-007 R REFEREN E, DCP 88-3078, Rev. 0 DE RIPTI HAN E:

Replaces the currently installed Unit 2 Liquid Radwaste Collection system primary containment isolation valves with new Anchor Darling parallel disc gate valves and new air-operated actuators.

ZUdgBRF:

I. No. The replacement valves stroke times will meet the requirements of FSAR Section 6.2.4 and be seismically qualified consistent with Section 3.2.1. The modified valve has the same function as the existing Pacific valve and the operation of this valve will not be changed in anyway.

II. No. See I. above.

IH. No. This change replaces the original valves with valves which are designed and built to the same standards and design conditions. The new valves will not in anyway change the operation of the Liquid Radwaste Collection System.

~ER~.: 91-008 R REFEREN E: DCP 90-3032, Rev. 0 DE RII'TI N F HAN E:

Installation of interfaces to new Turbine and Reactor Buildings Sample Stations.

MARY:

I. No. This modification does not change the function of the Process Sampling Systems or the operator's interface with these systems or any safety system, thus does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment related to safety.

II. No. This modification to the filters exhaust system and the addition of the exhaust.

booster fan is considered to be non-safety related. The operation of the new sample station is not required for safe shutdown of the plant.

III. No. This modification does not reduce the margin of safety since the interfaces being installed for the TBSS and RBSS are not safety-related and does not affect any margins of safety described in the Tech Specs.

JERKIN>.: 91-009 R RE  : DCP 90-3033 A-G, Rev. 0 E:

Replace Condensate Demineralizer Vessel Underdrain System, Unit l.

I. No. 'he accident analysis addressed in FSAR Chapter 15 is not affected by this modification. The new underdrains will be designed to withstand maximum system delta-P and willbe adequately supported against reaction forces under full flow conditions.

II. No. This action improves flow distribution through the underdrains and reduces the likelihood of poor quality feedwater resulting from poor flow distribution through the underdrain laterals.

III. No. The conductivity monitoring equipment will not be affected by this action.

Reactor water conductivity should be improved by reducing/eliminating condensate demineralized resin heels.

$ ER~.: 91-010 R RE  : DCP 90-3034 A-G, Rev. 0 DE RIPTI F HA E:

Replace Condensate Demineralizer Vessel Underdrain System, Unit 2.

SUMMARY

I. No. The accident analysis addressed in FSAR Chapter 15 is not affected by this modification. The new underdrains will be designed to withstand maximum system delta-P and willbe adequately supported against reaction forces under full flow conditions.

II. No. This action improves flow distribution through the underdrains and reduces the likelihood of poor quality feedwater resulting from poor flow distribution through the underdrain laterals.

III. No. The conductivity monitoring equipment will not be affected by this action.

Reactor water conductivity should be improved by reducing/eliminating condensate demineralized resin heels.

DCP 90-9075 Eliminate Spray Pond Low Temperature Alarms.

I. No. This modification willnot have any adverse impact on plant safety function, since the low temperature alarms to be deleted do not have any safety functions. The spray pond has the safety function of supplying water for the Emergency Service Water System and the Residual Heat Removal Service Water System, as well as serving as the ultimate heat sink for the plant.

II. No. The low temperature alarm provides notification that the pond surface is getting close to freezing. However, the FSAR points our that a lower pond temperature will increase the quantity of sensible heat that can be absorbed by the water volume. Allowing the pond surface to freeze would increase the heat absorption capability of the pond even more.

III. No. As stated above; this modification will not have an adverse impact on the ability of ESW, RHRSW, or the ultimate heat sink to perform their design functions.

+F~R~.: 91-014 R REFEREN: NL-90-042, Rev. 0; OP-139 (230) - 001, 002 & 003 DE RI HA E:

Revise procedures to support operation of seven condensate demineralizers at 100% power.

I. No. The system will maintain the required condensate purity as already defined in the existing chemistry programs while in the seven demineralizer configuration. The intent of the change is to improve the effluent purity.

II. No. A pressure boundary failure of a condensate demineralizer while in the seven demineralizer configuration would be no different that a failure in the six demineralizer configuration. Negative effluent quality changes could occur for many reasons other than this action and the effects on other equipment would be the same regardless of the reason.

III.

~ No.~ Tech Spec 3/4.4.4 sets Chemistry limits for the reactor coolant system. The basis of the specification is to prevent damage to the reactor materials in contact with the coolant. The intent of this action'is to improve reactor coolant chemistry.

$ ER~.: 91-015 R R  : DCP 88-3076A, Rev. 1 E:

Replace the Unit 2 Building Chilled Water Loop "A" primary containment outboard 3" isolation valves with new Anchor Darling parallel disc gate valves and new air-operated actuators.

R:

I. No. The replacement valves'troke times willmeet the requirements of FSAR Section 6.2.4 and will be seismically qualified consistent with FSAR Section 3.2.1. The modified valves have the same function as the existing Pacific valves. This modification will not change the operation of this valve in anyway.

II. No. See I above; the modification does not impact any safe shutdown paths as evaluated in Appendix R.

r III. No. This change replaces the original valves with valves which are designed and built to the same standards and design conditions. The new valves will not in anyway change the operation of the Reactor Building Chilled Water System.

/gal l~.: 91-016 Upgrade Auxiliary Boiler Conductivity Monitoring and Control Systems.

F A E: DCP 89-3031, Rev. 0 MMARY:

I. No. Since this modification replaces existing auxiliary boiler instrumentation with more accurate and reliable devices to enhance the operability and availability of the auxiliary boiler in support of producing a more reliable steam supply to end users, the modification will in fact decrease the probability of occurrence of any conductivity related failures that could affect auxiliary boiler operation.

II. No. The new conductivity system performs the same functions (control interlocks and alarms) as the system that is replaces. All failure modes of the new conductivity instruments are the same as the existing devices.

III. No. There is no discussion of the auxiliary boiler conductivity monitoring system, the auxiliary boiler itself, or the use of auxiliary steam by any plant component in the basis for any Technical Specification.

~Eg l~.: 91-017 R REFBRE: DCP 89-9127, Rev. 0 DE RI Install a check valve for each of the two Diesel Generator "E" starting air systems.

KNNIARY:

I. No. Failure of the check valve or malfunction of the air compressor that prevents the restoring of receiver pressure during normal plant operation willinitiate an alarm and appropriate operator actions in accordance with the cuirent Tech Specs. This action malines the ",E" diesel air start lineup similar to that of diesels A through D.

II. No. Failure of the check valve, of any kind, does not cause any different malfunctions than those of the existing air compressor or the associated components.

IH. No. In an event that the receivers cannot be recharged due to failure of this check valve, the low receiver pressure alarm will be actuated at 240 psig, and the operator is required to take the proper action. Recharging of the receiver occurs during normal operation rather than during emergency diesel starting.

~ER~V~.: 91-018 R REFBRB: DCP 89-9203, Rev. 0 DE RlPTI N F HA B.

Install additional public address system components in the Unit 2 condenser bay.

SUMh/A R Y:

I. No. The equipment to be installed by this modification will not interface with any plant component described as nuclear safety qualified or necessary for safety plant operation. The plant PA system is used to provide normal intra-plant communication; or in the event of a condition which threatens personal safety, to broadcast an alarm and provide instructions to plant personnel concerning the emergency condition.

II. No. The addition of the PA system equipment to the condenser bay will enhance the ability of the system to perform its primary design requirements of providing normal and emergency intra-plant communications. These components will be installed in the Turbine Building using established specifications and procedures.

III.

~ No.~ Tech Spec Section 3/4.9.5 deals with the requirements for communication capability during fuel movements in the reactor vessel. This communication capability is not affected by the changes to be incorporated by this modification.

NL-91-001, Rev. 0; OP-261-002 Cation resin in carboxylic acid form (normal cation resin is in sulfunic acid form).

I. No. The new resin is expected to improve vessel chemistry and reduce the probability of chemistry induced failures of all equipment in contact with coolant. This action reduces the potential effects of a RWCU resin intrusion since the impurities from the resin are reduced.

II. No. This action is expected to raise the purity of the coolant, by reducing baseline sulfate levels slightly.

III. No. This action should improve Tech Spec chemistry parameters.

$gR~.: 91-020 R RE E: DCP 89-3021C, Rev. 0 Improve heat trace control of containment atmosphere control isolation valves.

MARY:

I. No. This action assures that the heat applied to the H,/0, analyzer and CRM panel piping sensing lines and valves maintain the sample temperature above the dew point and does not exceed the environmentally qualified temperature of the target rock solenoid containment isolation valve.

II. No. The action to reduce the watts per foot applied by the heat trace to improve the heat trace control system on the H,/0, analyzer and CRM piping and install insulation blankets on the target rock solenoid valves does not create the possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.

III. No. The Reactor Coolant Leakage Detection System as identified in the Tech Specs is required to monitor and detect leakage from the reactor coolant pressure boundary. This action does not affect the detection system's ability to perform its function.

/ERIN.: 91-021 R REFERE E: DCP 89-9142, Rev. 0 DE RlPTl F H K Expand fire protection capabilities of fire protection subsystem WPS-131.

SgdMAAR':

I. No. The additional fire protection provided by this modification will decrease the probability of a major fire in the turbine area and thus decrease the probability of a turbine trip which may chanenge safety systems. Implementation of this modification will increase the availability of fire protection for the turbine and components supporting turbine operation.

II. No. The modification expands the protective function of WPS-131, and is designed in accordance with the original design criteria and codes imposed on the original equipment.

III. No. Since we are adding to the fire protection coverage of the turbine and auxiliaries, the margin of safety is increased.

$ ER~.: 91-022 R REFEREN K DCP 89-9169, Rev. 0 DE RlPTl F HAN fire protection capabilities of fire protection subsystem WPS-231.

K'xpand

+I~~RY:

I. No. The additional fire protection provided by this modification will decrease the probability of a major fire in the turbine area and thus decrease the probability of a turbine trip which may challenge safety systems. Implementation of this modification will increase the availability of fire protection for the turbine and components supporting turbine operation.

H. No. The modification expands the protective function of WPS-231, and is designed in accordance with the original design criteria and codes imposed on the original equipment.

III. No. Since we are adding to the fire protection coverage of the turbine and auxiliaries, the margin of safety is increased.

PMR 89-9174, Rev. 0 Install emergency lighting for Technical Support Center.

I. No. Emergency lighting units are being added to provide iHumination at the TSC during a station blackout and willnot interfere with the logic control or operation of any safety-related plant system or components.

IL No. By assuring that the electrical coordination, separation requirements, diesel generator loading, voltage drop, circuitry ampacity and seismic design requirements and limits are met, the possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR, is not created.

III. No. The circuit breaker for the power feed to the individual emergency lighting units is coordinated with upstream devices. Voltage drop and load ampacity are within design limits as analyzed. Therefore, this modification does not reduce the margin of safety as defined in the basis for any Technical Specification.

$ER~.: 91-024 R REFE SCP J891005, Rev. 0 Increase the HPCI turbine high exhaust pressure trip setpoint from 75 psig to 110 psig for Unit 2.

I. No. The setpoint change will decrease the consequences of an accident (LOCA),

because the higher pressure setpoint will decrease the probability that the HPCI turbine will be tripped as a result of high suppression chamber pressure during accident conditions.

II. No. The malfunction (failure) of the HPCI system is considered under the single failure criteria; therefore, a malfunction that results in a premature turbine trip or a failure to trip the turbine on high exhaust pressure has been evaluated.

III. No. Since this setpoint change only affects an equipment protection function and does not change the operation of the HPCI system, the change will not reduce the margin of safety of any system as defined in the basis of the Technical Specifications.

ggR~7.: 91-025

SCP J882032, Rev. 0 Increase the HPCI turbine high exhaust pressure trip setpoint,from 75 psig to 110 psig for Unit 1.

I. No. The setpoint change will decrease the consequences of an accident (LOCA),

because the higher pressure setpoint will decrease the probability that the HPCI turbine will be tripped as a result of high suppression chamber pressure during accident conditions.

II. No. The malfunction (failure) of the HPCI system is considered under the single failure criteria; therefore, a malfunction that results in a premature turbine trip or a failure to trip the turbine on high exhaust pressure has been evaluated.

III. No. Since this setpoint change only affects an equipment protection function and does not change the operation of the HPCI system, the change will not reduce the margin of safety of any system as defined in the basis of the Technical Specifications.

~EE~).: 91-026

NL-91-002, Rev. 0; TP-249-040 This procedure installs test equipment to monitor the RHR system piping pressure and relief valve actuations.

M R:

I. No. The pressure monitoring instrumentation will remain isolated from the main process piping instrumentation during normal plant operation through the use of normally closed isolation valves. The PSV actuation monitoring instrumentation is 100% passive in that it monitors flow out of the relief valve discharge piping through the use of water flow switches and change state to indicate relief valve actuation.

II. No. A leak different from those already analyzed in the FSAR is only possible when .

this procedure is actually being performed at which time the test director will be located at the instrument racks and any problem could be isolated immediately and reported directly to the control room; shutdown cooling could remain in service without any effect.

III.

~ No.~ All containment isolation features and the shutdown cooling isolations remain operable throughout the entire testing. This instrumentation installed will be separate from all other permanent plant instrumentation.

DCP 89-9102, Rev. 0 Replace the existing Unit 1 Automatic Priority Interrupt (APQ detected circuit of the display control system in the plant computer with an improved, more reliable design. Only the hardware design of the API detect circuit will change.

\

I. No. The rephcement of the API detect circuit in the plant computer system does not increase the probability or consequences of an accident since there are no safety-related systems in the plant computer room. The plant computer system is not essential for safe shutdown of the plant and serves no active emergency function during operations.

G. No. Improving the design of the switching detection circuit willenhance the operation of the display control system. This modification does not change the design criteria for any safety-related system or function as described in the FSAR.

HI. No. The intent of the modification is to improve the operability of the display control system and, thereby, increase operator confidence and margin of safety.

1

$jggj~.: 91-028 DCP 89-9103, Rev. 0 DE RI HA E:

Replace the existing Unit 2 Automatic Priority Interrupt (APQ detected circuit of the display control system in the plant computer with an improved, more reliable design. Only the hardware design of the API detect circuit will change.

RY:

I. No. The replacement of the API detect circuit in the plant computer system does not increase the probability or con~uences of an accident since there are no safety-related systems in the plant computer room. The plant computer system is not essential for safe shutdown of the plant and serves no active emergency function during operations.

II. No. Improving the design of the switching detection circuit willenhance the operation of the display control system. This modification does not change the design criteria for any safety-related system or function as described in the FSAR.

III.

~ No.~ The intent of the modification is to improve the operability of the display control system and, thereby, increase operator confidence and margin of safety.

~ER l~.: 91-029 R REFERE E: DCP 90- 3010, Rev. 0 DE RIPTI F H E:

Replace the existing RCIC test valve (HV-249F022) internals with specically designed thiottling internals. This will permit the valve.to pass rated flow during both the high and low pressure surveillance test with no excessive vibration.

I. No. This modification does not affect the ability of the valve to close,'r stay closed, as required to support RCIC injection to the Reactor Piessure Vessel (RPV).

II. No. This modification will enhance the ability of the RCIC full flow test valve to perform its testing function at the 150 psi surveillance test.. This valve is required to go closed on a RCIC injection signal iri order to inject to the RPV during an accident. The new remachined throttling trim installed by this change enhances the throttling capability of the valve and has no adverse effect on the valve's closing ability. The valves full open to full closed cycle time has not changed.

HI. No. The safety related function of this RCIC valve is to remain in its normally closed position or actuate to the closed position if the valve is in the throttled open position during the Tech Spec required surveillance test. In either case the margin of safety is unaffected since the new internals have no effect on the ability of the valve to close.

JERKIN).: 91-030 R REFERE E: SCP E911001, Rev. 0 HA Change the setpoints for existing Class 1E 250V DC system overvoltage and undervoltage relays for Unit 2.

KMhQRY:

I. No. The 250V DC system provides reliable and continuous power to safety related pumps, valves, and power supplies. A failure of these relays to alarm does not cause a decrease in core coolant temperature or any other potential accident listed in the FSAR.

H. No. A failure of these relays does not prevent the Class 1E 250V DC system from meeting its minimum performance requirements. A voltage higher than the alarm setpoint of 288V DC is indicative of a battery charger problem or failure. The effects of a battery charger failure have been evaluated as part of the FSAR single failure evaluation.

III.

~ No. ~ This setpoint change package does not affect the operation of the Class 1E 250V DC system. Further, this setpoint change package does not cause a failure of the Class 1E 250V DC system.

~ER~N.: 91-031 R REFE  : SCP E911004, Rev. 0 Eo Change the setpoints for existing Class 1E 125V DC system overvoltage and undervoltage relays for Unit 2.

SIMEARY:

I. No. The 125V DC system provides reliable and continuous'control and indication power to isolation valves, circuit breakers, inverters, annunciators, and other loads supplied from the 125V DC Class 1E and non-Class 1E distribution panels.

A failure of these relays to'alarm does not'cause a decrease in core coolant temperature or any other potential accident.

H. No. A failure of these relays does not prevent the Class 1E 125V DC system from meeting its minimum performance requirements. A voltage higher than the alarm setpoint of 144V DC is indicative of a battery charger problem or failure. The effects of a battery charger failure have previously been evaluated as part of the FSAR single failure evaluation.

IH. No. This setpoint change package does'not affect the operation of the Class 1E 125V DC system. Further, this setpoint change package does not cause a failure of the Class 1E 125V DC system.

~E~.: 91-032 R REFERE E: SCP E911009, Rev. 0 DE RI F HA E:

Change the setpoints for existing Class 1E 250V DC system overvoltage and undervoltage relays for Unit l.

SUbLhfARY:

I. No. The 250V Dc system provides reliable and continuous power to safety related pumps, valves, and power supplies. A failure of these relays to alarm does not cause a decrease in core coolant temperature or any other potential accident listed in the FSAR.

H. No. A failure of these relays does not prevent the Class 1E 250V DC system from meeting its minimum performance requirements. A voltage higher than the alarm setpoint of 288V DC is indicative of a battery charge problem or failure. The effects of a battery charger failure have been evaluated as part of the FSAR single failure evaluation.

IH. No.

~ ~ This setpoint change package does not affect the operation of the Class lE 250V DC system. Further, this setpoint change package does not cause a failure of the Class 1E 250V DC system.

QlR gg.: 91-033

SCP E911010, Rev. 0 Change the setpoints for existing Class 1E 125V DC system overvoltage and undervoltage relays for Unit 1.

I. No. The 125V DC system provides reliable and continuous control and indication power to isolation valves, circuit breakers, inverters, annunciators, and other loads s'upplied from the 125V DC Class 1E and non-Class.lE distribution panels.

~

A failure of these relays to'alarm does'ot cause a'e'ciease in coie coolant temperature or any other potential accident.

II. No. A failure of these relays does not prevent the Class 1E 125V DC system from meeting its minimum performance requirements. A voltage higher than the alarm setpoint of 144V DC is indicative of a battery charger problem or failure. The effects of a battery charger failure have previously been evaluated as part of the FSAR single failure evaluation.

HI. No. This setpoint change package does not affect the operation of the Class 1E 125V DC system. Further, this setpoint change package does not cause a failure of the Class 1E 125V DC system.

JgR Pg).: 91-034 R REFERE: DCP 90-3072, Rev. 0 E RlPTI F HA Add one 2" ball valve to each ESW line returning from the "A" and "B" Dx Coolers. Remove currently instaHed orifice plates.

JVMhfARY:

I. No. The modification does not create any functional changes to the operation of the ESW system nor does it affect any active system components. This modification essentially substitutes a valve for a short section of pipe having the same pressure boundary classification. The valve is prevented from changing position due to vibration by a positive locking mechanism.

H. No. The modification does not affect the function or the operation of the ESW system or any system relying on ESW. Specifically the 2" ball valve willbe used instead of the existing flow orifice to limit the maximum flow to the Dx coolers while ensuring that the minimum flow requirement is also met.

III.

~ No. ~ The addition of the 2" ball valve does not impair the ability of the system to provide minimum cooling capacity because the valve may be positioned to permit the same pressure drop as the existing flow orifice.

/gal.gQ.: 91-035

DCP 90-3075, Rev. 0 Install a highly accurate ultrasonic flow monitoring system for both the Unit 1, 96 inch pipes supplying circulating water to the condenser.

I. No. This installation does not materially affect the design features listed in the FSAR.

It is judged that any increase in probability due to the simple addition of components within the Circulatin'g Water piping is entirely bounded by the uncertainty present in the calculation of the probability of occurrence.

II. No. The installation of an ultrasonic flow monitoring system on the interior walls of the circulating water pipes and cable penetration assemblies through the pipes is slightly different than existing Susquehanna flow measurement design. However, as long as the tranducers, pull boxes (and shields), and piping are welded onto the pipes per PP&L Specification and ASME Section IX, there should be no concern with the ultrasonic flow monitoring system affecting another component or system.

III.

~ No. ~ This modification affects the circulating water system which has no safety-related function. In particular, flow measurement of circulating water has no direct interface with any safety function in the plant.

JEER~.: 91-036 FERE: DCP 90-9080 E RIPTI F A E:

To reinstate the feedwater heater high level alarm which was disconnected by a bypass which was initially installed to alleviate a nuisance alarm on a local panel.

A Y:

I. No. The reconnection of a local. alarm function does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated.

II. No. This modification enhances the feedwater heater level controls by providing the operator an alarm on rising water level. This modification adds no new function to the feedwater heater vents and drains nor does it remove any function of the system.

IH. No.

~ ~ This modification only reinstates a local alarm function.

/gal.~.: 91-037 R REFERE E: DCP 91-9001 A thru 0, Rev. 0 Replace Unit 2 Class 1E Oversized 125V DC Circuit Breakers.

I. No. The replacement breakers are identical replacements for original equipment except that the trip ratings match the derated ampacities of the connected cables in order to minimize to the extent practicable the effects of a sustained limited magnitude fault.

II. No. See I above.; installation of properly sized and dedication tested replacement breakers will not modify the safety functions of the system loads.

HI. No. This action does not modify the function or capacity of any connected safety system and will increase the probability of safety systems remaining functional in the event of a sustained limited magnitude fault above the cable ampacity by o pe ning the affected circuit.

$ ER~.: 91-038 EFERE: NL 90-029, Rev. 0 DE Rl F HAN E:

Evaluate use of the Temporary Laundry Facility to sort, store, ship and receive protective clothing.

JQMMARY:

I. No. The Temporary Laundry Facility is a stand alone, temporary facility. The facility has no safety-related functions, and the facility does not interface with any safety related systems. Since the laundry trailers do not play a critical role in plant operations, their operation does not increase the probability or consequences of an accident or malfunction.

II. No. Accident analyses in the FSAR are not performed for scenarios involving the processing of contaminated protective clothing.

III. No. See I. above. The facility does not interfere with any other system or component governed by Technical Specifications.

NL-91-003, Rev. 0 Eo Establish a control zone break area for the Unit 2 generator rewind project.

KMLY:

I. No. The break area for the turbine rewind project will be a stand alone structure located on e1..729'f the unit 2 turbine deck. This non-safety related structure contains no safety-related equipment or radwaste equipment whose failure was postulated in the accident analysis of the FSAR.

H. No. The design and location, of the bre'ak area provides an area within the'ontrolled Zone which is maintained free of contamination. Extensive administrative controls have been invoked to ensure personnel and materials entering this facility are not contaminated. This temporary eating, drinking and smoking area will, therefore, be maintained free of contamination as required by the FSAR.

III. No. No requirements applicable for temporary break areas are contained within Technical Specifications.

~E>~Q.: 91-040 RE E: DCP 89-3021C, Rev. 1 DE RI Improve Heat Trace Control of the Unit 2 Containment Atmosphere Control Valves (CAC).

KMILF:

I. No. This action assures that the heat applied to the H,O, analyzer and CRM panel piping sensing lines and valves maintains the sample temperature above dew point and does not exceed the environmentally qualified temperature of the target rock solenoid containment isolation valves. This permits the analyzers, panels and isolation valves to perform their intended safety functions.

II. No. The action to reduce the watts per foot applied by the heat trace, to improve the heat trace control system on the H,O, analyzer and CRM piping, to rework or remove the heat trace from the CAC target rock solenoid isolation valves and .

install insulation blankets on the solenoid valves does not create the possibility of a different type previously evaluated in the FSAR.

III. No. This action does not affect the detection system's ability to perform its function; and does assure operability of the valves to meet their isolation function by maintaining the valve temperatures below their environmentally qualified limit.

$ ER~.'.91-041 R REFE  : DCP 89-9152 A, B & C, Rev. 0 Replace the existing Unit 2 RHR Pump Seal with a new "Hot" seal which does not require any external cooling medium.

I. No. The performance of the new "hot" seal will not increase the consequences of an accident. The new seal is. expected.to leak no more than the existing design under normal and accident conditions. The new seal design has no impact on the ability of the RHR pump to provide LPCI or containm'ent cooling flow.

II. No. The new seal will not increase the potential for a seal failure incident. The seal is designed to withstand the maximum RHR temperature conditions, without cooling, and designed generally in compliance with the original pump specifications. The new seal shall eliminate the RHR/ESW interface thus tending to simplify the operation of the systems and reduce the potential for operational error or radioactive releases to the environment.

III.

~ No.~

~ ~ ~ ~ ~ ~

This action will not compromise the assumptions used in the accident analysis,

~ ~ ~ ~

nor invalidate the existing integrity of the containment system to control release

~ ~

~ ~ ~

of radioactive materials to the environment. ~

$ ~~.: 91-042

DCP 89-9177, Rev. 0 Repair and restore to the original design condition a platform located in the Unit 2 Turbine Building.

SIMfSRR I. No. This platform is classified as "other" safety class (FSAR Table 3.2-1) and no equipment important to safety is located in the vicinity of this platform.

II. No. Since the repair of this platform effectively replaces the structure which was partially dismantled during a previous modification, this action does not create the possibility of an accident not previously evaluated.

III.

~ No.~ See I. and II. above.

~ ~ ~

$ ER~.: 91-043 DCP 90-3076, Rev. 0 Install Flanges for Unit 2 Containment Instrument Gas System.

I. No. This modification only installs flanges in existing piping in order to facilitate easier maintenance of safety related check valves. The piping including the new flanges will remain in conformance with all requirements as'originally designed.

II. No. This modification does not alter the function, change the operation or introduce new failure modes for the Containment Instrument Gas System, the MSIVs or any other safety related components.

III. No. Since this modification does not alter the function or the operation of any safety related systems, the margin of safety defined in the basis of the Technical Specification is not reduced.

$ gR ggP.: 91-044

DCP 90-3077 A & B, Rev. 0 A E:

Standby Gas Treatment Circuits Voltage Improvement Modification Unit 1.

SUMMARY

I. No. The purpose of this modification is to provide a resolution to the existing voltage drop condition. Paralleling existing cables with the spare cables will result in improved voltage which will increase reliability of the standby gas treatment circuits.

II. No. This modification does not affect the operation of the standby gas treatment system or any related equipment. It increases the reliability of two circuits as a result of the voltage drop decrease.

III. No. See I. and II. above.

$ ER~.: 91-045 DCP 90-9052, Rev. 0 Install a desktop type of public address station in the Health Physics access area on the south end of the refuel floor - Unit 2.

8 I. No. Components and conduit instaHed by this modification are located on Unit 2 refuel floor at the south end. The majority of equipment will be installed within the Health Physics access area and the remaining conduit and terminal box at approximately'Col. Q-37.4. Neither area includes safety-related components that will be impacted by this modification.

II. No. The addition of the desk top PA to the Health Physics access area will enhance the ability of the system to perform its primary design requirements of providing normal and emergency intra-plant communications. No new unanalyzed type of malfunction is made probable by the modification and relocation of these PA components.

IH. No. The communication capability, as described in Tech Spec Bases, is not affected by the changes to be incorporated by this modification.

/BRAN.: 91-046

DCP 91-3001, Rev. 0 DE RI F N E:

Provides the requirements for the storage of spent control blades in the Unit 2 Spent Fuel Pool.

SfEKILY I. No. For the long-term storage of control blades addressed by this modification (seismic/hydrodynamic loads apply), a maximum of two control blades (attached to sling assemblies) per bracket is required. This assures that the structural integrity of the existing brackets is maintained for long-term storage of control blades.

H. No. Control blade banging will not cause malfunction or failure of any safety-related components. Therefore, the banging of control blades does not represent a new unanalyzed type of malfunction or accident.

III. No. This modification does not impact the acceptance limit (1100 pounds) since the weight of a control blade (215 pounds) plus the sling assembly (10 pounds J) is much less than the 1100 pound limit.91-047

DCP 91-9010 A & B and 91-9011 A & B, Rev. 0 Make physical changes to the Reactor Building HVAC damper actuator assemblies so that they may be locked in their normal design operating position.

I. No. This modification removes the "Refuel Mode" operation of these dampers.

Operating under this mode during plant refueling has resulted in undesirable contamination'pioblems due to unfainiliar and unpredictable air flow patterns on the refueling floor. As long as we meet the secondary containment breakdown and inleakage requirements, the damper position has no impact on any design b'asis accident.

II. No. The changes and/or failure modes associated with the changes, have no impact on the function and performance of the safety related systems (R.B. Recirculation Fan and SGTS). The changes do not modify the plant response to any design basis accident.

III. No. These modifications assure that the secondary containment integrity can be maintained by locking the above dampers in their draw down test position. These modifications neither reduce the operability of any equipment nor require additional surveillance to ensure plant safety.

DER.M-'1-048

NL-91-004, Rev. 0; TP-016-006 D I Pumpdown of B Loop RHRSW and ESW piping.

SlggfkkLY A Safety Impact Assessment was completed to analyze the affects of removing the B Loop ESW and RHRSW system for 7 days and was found to be acceptable.

Draining and refilling of the B Loop ESW/RHRSW piping will be accomplished under this procedure within the 7 day LCO window and this does not impact the Safety Impact Assessment.

II. No. The TBCCW/RBCCW heat exchangers are non-essential heat exchangers and removing a backup cooling source will not affect safe shutdown of the plant. No required ECCS systems willbe rendered inoperable by this procedure. Secondary containment integrity will be adequately controlled. Operability of the RHRSW loop will be maintained.

III.

~ No.~ Operability of the A RHRSW system, ECCS systems, secondary containment integrity will not be affected by this procedure; thus margin of safety will not be reduced.

SCP-J9-1015 A & B, Rev. 0 Increase the setpoint of the "instrument failure" function of the flow switches installed in the HVAC ducts in the Control Structure.

I. No. Implementation of this setpoint change has no affect on the functions of the equipment other than to improve the reliability of the low-flow switches. The increased reliability is achieved by preventing erroneous trips of the low-flow switches during normal flow conditions. This results in a corresponding increase in the reliability of the Control Room and control Structure HVAC system.

II. No. This setpoint change does not affect the design or operation of the Control Room and Control Structure low-flow switches, HVAC fans or the heater/humidifier.

This change increases'system reliability by eliminating the possibility of erroneous trips caused by faulty operation of the low-flow switches.

III.

~ No.~

~ ~

This change increases the reliability of the Control Structure HVAC system, and therefore, the Control Structure Chillers; it does not reduce the margin of safety.

5~V~.: 91-050 R RE R E PMR 86-7036, Rev. 0 Rework the source of Unit 2 Fuel Pool Makeup from Demineralized Water to Condensate.

+1~ARY:

I. No. The normal fuel pool makeup source has no direct safety related function and is designed only to provide one source of makeup water for the fuel pools evaporative losses. This modification will not impact any safety-related equipment as defined in the'FSAR.

II. No. See I. above, the normal fuel pool makeup operation is backed up by two independent safety-related ESW makeup sources.

III. No. This modification does not effect the safety function of any components associated with the fuel pools or fuel pool cooling system.

0

'I 91-06, Rev. 0; TP-234-040 Unit 2 Reactor Building HVAC, ECCS and RCIC Pump Room Cooler Replacement Coil Acid Flush and Fin cleaning.

I. No. The acid fiush and fin cleaning does not alter the function or the performance of these safety-related systems. It will restore the integrity and design cooling capability of the replacement coils.

II. No. The integrity and heat removal capability of the replacement coo1ing coil is restored by the acid flush and fin cleaning. This does not alter the function or the performance of any safety-related system.

III. No. This cleaning does not reduce the operability of any equipment and it does not require any additional surveillance to ensure plant safety. This does not change the function and performance of any safety-related system.

ggg.ggl.: 91-052

NL-91-008, Rev. 0; TP-283-013 DE RlP17 F Electrically operate and backseat the 1.0 inch Drip Leg Drain Bypass Valve on the Unit 2 Main Steam Bypass Header.

I. No. While this procedure may result in a small increased risk of an operational transient due to a valve mechanical failure, it does not affect equipment important to safety. This valve is open and remains open under all normal plant evolutions and binding of this valve in the open position causes no operational concerns.

II. No. Failure of the valve bonnet is evaluated as a small potential risk during the backseating evolution. Ifa rupture were to occur, sufficient indication of or physical conditions is available to enable prompt operator action adverse'adiological to mitigate the consequences of the rupture.

III.

~ No. ~

~ ~

This valve is not a safety-related component and is not addressed in Tech Specs.

/BR hQ.: 91453 R  : NL-91-06, Rev, 1; TP-234-040 Unit 2 Reactor Building HVAC, ECCS and RCIC Pump Room Cooler Replacement Coil Acid Flush and Fin Cleaning.

I. No.. The acid flush and fin cleaning does not alter the function'or the performance of these safety-related systems. It will restore the integrity and design cooling capability of the replacement coils.

II. No. The integrity and 'heat removal capability of the replacement cooling coil is restored by the acid flush and fin cleaning. This does not alter the function or the performance of any safety-related system.

III. No. This cleaning does not reduce the operability of any equipment and it does not require any additional surveillance to ensure plant safety. This does not change the function and performance of any safety-related system.

/gal gg7.: 91-054 R RE RE E: DCP 90-9032A, Rev. 0 DE RlPTI A E:

To delete the alarm, "Condenser A, B, C HI Conductivity", in Control Room Panel 2C651.

I. No. Conductivity monitoring will continue to provide tube leak or rupture detection through the installed conductivity cells input to recorders, local alarms and demineralizer inlet monitors. The conductivity monitoring is not required to aid in the mitigation of a design basis accident.

l II. No. Condenser tube leaks are monitored by the existing conductivity elements and data is input to recorders and local panel alarms. Therefore, removal of the remote alarm will not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR.

III.

~ No. ~ The Technical Specification bases are not affected by this modification and thus the margin of safety is not reduced.

$gR~Q.: 91-055 R  : DCP 90-9077, Rev. 0 Separate the Non-Class 1E Limit Switches from the Class 1E RBCCW Valve Control and Position Indication Circuitry.

I. No. The operation of the limitswitches is not requirod for a safety function. A failure of the switches could only result in erroneous valve position indication. The failure does not compromise a safety related system 'or component. Also, the present fail-safe operation of the ESW valves is not changed. The affiliated cables and internal wiring of the MDR isolation relay circuitry continue to be subjected to the same separation requirement as Class 1E cables and wiring in accordance with PP&L Specification E-1012.

II. No. The action to separate the non-Class 1E Service Water Valve Limit Switches from RBCCW-ESW valve control and position indication circuitry does not adversely affect the operation of the ESW system valves and does not impact the operation of the Class 1E electrical system. The present fail safety operation of the ESW valves is not changed.

III. No. The Emergency Service Water System ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. This action does not affect the operability or capacity of the ESW system nor introduce any new mode of operation.

~SR No.! 91456

DCP 90-9078, Rev. 0 Upgrade the Unit 2 TBCCW-ESW Valve Control and Position Indication Circuitry from affiliated to Class lE and to separate the non-Class 1E limit switches from the Control and Position Indication circuitry.

j I. No. The limit switches on the SWS valves only provide position'indication through the MDR isolation relays to the TBCCW-ESW valve control and remote position indication circuitry. The operation of the unit switches is not required for a safety function. A Mure of the switches could only result in erroneous valve position indication. The failure does not compromise a safety related system or.

component.

II. No. Since the pilot solenoid valves, limit switches on the ESW valves, hand switches, auxiliary relays, lights, resistors, terminal blocks and fuses are continuing to be Class 1E devices, the present fail safe operation of the ESW valves is not changed.

III. No. The Emergency Service Water System ensures that sufficient cooling capacity is available for continued operation of safety related equipment during normal and accident conditions. This action does not affect the operability or capacity of the ESW system nor introduce any new mode of operation.

~ER~.: 91-057 R REFE  : NL-91-007, Rev. 0; OP-205-004 & OP-205-005 DE RIPTl A E Maintain Unit 2 Reactor Building Zone IUIIIVentilation operable by temporarily powering selected safety related Zone II/IHventilation damper schemes.

R:

I. No. This temporary power is provided from the same divisions, is reliable Class 1E power, and will be available, during the bus outages. There are no failure mode differences between the permanent and temporary power sources.

II. No. This temporary configuration is substituting the normal Class 1E power source with another source from the same division. This application of temporary power allows Unit 1 to operate by keeping the Unit 2 Zone III ventilation system operable during the 4Kv bus outages.

III.

~ No. ~ The installation of temporary Class 1E power supply does not prevent any design safety function of safety related equipment. The power supply has been analyzed and shows that utilization of temporary Class 1E power sources does not degrade the Zone II & III Reactor Building HVAC system below an acceptable level.

'

~ERjg7.: 91-058 NL-91-009, Rev. 0; TP-241-007 Unit 2 Main Condenser Descaling.

I. No. The Mure mode associated with the. cleaning process is degradation of

'irculating/service water system components. The cleaning chemicals will not cause degradation to system materials or components. The failure mode associated with this procedure will not be an initiating event for any design basis accident.

II. No. The tube cleaning process will not have any adverse effects on any materials in the system. The chemical product used for cleaning the condenser tubes will not adversely impact or compromise the life of any materials the products will contact. There is no change to the design basis and there is not impact on the operation of any systems in the plant as a result of the cleaning operation.

III. No. See I. Ec II. above.

$ ~~.: 91-059 DCP 90-3071, Rev. 0 DE RI Install a Replacement Actuator for RHR Suction Isolation Valve HV-251F004A.

$Ql~fARY:

I. No. The modification will increase the reliability of this valve by insuring full closure when called upon to do so. The replacement actuator will deliver more thrust to the valve stem. Operational and functional requirements will remain unchanged.

II. No. The modification is intended to improve the isolation integrity function of the RHR F004A valve by providing the required actuator thrust necessary to overcome valve internal resistance due to its disk/seat angle mismatch. The valve is normally open and no change of state is required for accident conditions.

III. No. The modification will not reduce the valve's ability to remain open, but will increase the valve's ability to function as an isolation point. The margin of safety will not be reduced in any of the required functional requirements of the RHR F004A valve. The new actuator will insure the reliability to act as an isolation boundary when required.

$ ~R~.: 91-060 DCP 89-9205, Rev. 0 Increase Diesel Rooms A, B, C & D general room illumination and control panel lighting to at least the recommended foot-candles.

I. No. The lighting system is designed to furnish illumination levels required for safe performance of plant operation, security, shutdown, and maintenance'duties. The operational integrity and safety of the diesel generators and subsystems will not be compromised by this modification.

II. No. The lighting is designed to meet the same standards, thus the ability of the diesel generators and subsystems to satisfy the design bases is not compromised. No change to the lighting system function is occurring. System interfaces are remaining unchanged and no new accident or malfunctions are introduced.

III. No. This change does not alter the ability of the diesel system to perform its intended function and satisfy the Technical Specification Basis.

Q~Q.: 91-061

DCP 90-3070, Rev. 0 Add components that provide auxiliary control to the damper circuit of the Control Structure Heating, Ventilating, and Air conditioning (CSHVAC) System.

I. No. This modification is being performed to ensure compliance with 10CFR50, Appendix R. The criteria of Appendix R requires that equipment and systems comprising the means to achieve and maintain cold shutdown condition shall not be damaged by fire.

II. No. This modification does not create any additional scenarios which could prevent the CSHVAC system from fulfillingits design requirements, either from normal or emergency operation. The addition of auxiliary damper control logic maintains the redundant HVAC systems as described in the FSAR.

III. No. This modification does not effect the Control Room Emergency Outside Air Supply System (CREOASS) as described in the FSAR and thus does not reduce the margin of safety.

DCP 91-9012M thru 91-9012ME, Rev. 1 Replace Unit 2 Non-1E Oversized 125V DC Circuit Bre dms.

I. No. The replacement breakers are identical replacements for original equipment except that the trip ratings match the derated ampacities of the connected c'ables in order to minimize to the extent practicable the effects of a sustained limited magnitude fault.

II. No. Installation of properly sized and dedication tested replacement breakers will not modify the non-safety related functions of the system loads. This modification willnot introduce any failure modes different than already analyzed in the FSAR.

III. No. This action does not modify the function or capacity of any connected non-safety related systems and wiH increase the probability of systems remaining functional in the event of a sustained limited magnitude fault above cable ampacity by opening the affected circuit. This limits, to the extend practical, damage to adjacent cables of other non-safety systems in the same raceway.

ging.~.: 91-063

TP-259-013, Rev. 0 DE 1PTI F H N Perform a 45 psig LLRT on the MSIVs in the accident direction.

$ Qg~AgF:

I. No. The safety function of the MSIVs and the Leakage Control System to isolate with their appropriate containment isolation signal will not be degraded with the relatively small vacuum being applied inside the test volume, and not required to be operable during this test.

II. No. The Main Steam Line (MSL) Plugs will be pressurized below the evaluated pressure. All requirements for use of the MSL plugs are incorporated into the prerequisites and flowpath of this procedure.

III. No. This procedure will increase the test pressure to 45 psig but this change will highlight actual valve seating flaws and decrease maintenance on minimal defects, which can introduce valve problems. The overall intent is to test primary containment integrity by leak testing penetrations at accident pressure.

0

~8~.: 91-064

TIP-234-036, Rev. 0 Tie-In of Temporary Outage Chiller for Drywell Cooling during Unit 2 outages.

I. No. Installation of this temporary system will not impact operation of primary containment isolation, the only safety function of RBCW system. Normal operation of the reactor building ventilation system- will'not'be impacted by installation of the temporary chiller. Additionally, this equipment will be installed during a period when secondary containment is not required.

II. No. The installation of this temporary system does not adversely impact the operation or safety function of any existing installed equipment.

III. No. See I. & II. above.

$ /)~).: 91-065

DCP91-0071, Rev. 0 Short-term administrative elimination of a bypass that replaced the non-pressure retaining pipe cap (1/4" dia. weep hole in it) and the reactor recirculation loop 'B'rain line with a pressure-retaining cap (no weep hole) to contain leakage through the Reactor Coolant pressure boundary (RCPB) valves 243F051B and 243F052B.

$g ARY:

I. No. The addition of the pipe cap acts to contain leakage past the RCPB isolation valves only. The change does not alter the system contents, functions, or boundaries in any manner which would increase the radiological consequences of the Loss-Of-Coolant-Accident described in FSAR 15.6.5.

H. No. The postulated failure of the HBD piping does not present a new mode of failure for the loss-of-coolant accident discussed above. Since the stress levels in the HBD piping have been determined to meet the break exemption criteria, the probability of occurrence of the design basis accident discussed above is considered to be unchanged.

III. No. This change provides a temporary 'secondary'ressure retaining boundary downstream of the RCPB/ASME boundary which is intended to contain leakage due to normal wear of the boundary isolation valves until the valves are replaced under the valve upgrade program during the Unit 2 5RIO (under DCP 91-9007B).

The temporary configuration created by this DCP does not affect the existing boundaries or the bases for the Tech. Spec. limits in Section 3/4.4.3 (3.4.3.2)

"Reactor Coolant System Leakage".

e

/HARV~.: 91-066 R REFB K DCP89-9159A, Rev. 0 ll The proposed change will enable the HPCI Stop Valve balance chamber pressure to be monitored which will subsequently provide for optimum stop valve adjustment and operation.

This modification shall permanently upgrade a pressure transmitter loop now existing on a temporary basis. The output from the pressure transmitter-will be used to provide a readout on GETARS.

R:

I. No. The proposed modification does not impact any station design features that are used in radioactive release or loss of coolant analyses for postulated instrument tubing failures. Failure analyses for the proposed changes are enveloped by the accident analyses "Instrument Line Break" (FSAR Section 15.6.2) and Mscellaneous Small Releases Outside Containment (FSAR Section 15.7.2. 1).

II. No. Worst case scenarios for an accident involving an instrument tubing rupture have been conservatively analyzed in the FSAR. The proposed change does not involve a change in system operation or add a more severe type of failure mode.

No new unanalyzed type of malfunction is made probable by this modification.

HI. No. Technical Specifications 3/4.3 Instrumentation and 3/4.5 Emergency Core Cooling Systems provide the bases and limiting conditions of operation for associated plant instrumentation and the HPCI system. All instrument tubing installation will follow accepted standards and codes with the potential for a HPCI line break scenario being extremely low and offset by the increased probability of correct stop valve operation due to the improved monitoring capability during adjustment; therefore, the margin of safety will not be reduced.

~.: 91-067 R REFE E: DCP 90-9048, 90-9049 V

Several instances of blown control fuses have occurred resulting in loss of ARI control power.

Blown fuse events occurred as a result of indicating lamp failures in, 1CB224A,B and The present ARI control logic utilizes a full voltage indication lamp design. -'CB224A,B.

This change willreplace the full voltage lamp with a lower voltage'lamp with a resistor in series to reduce the short circuit current and prevent the fuses from blowing.

I. No. As listed in FSAR Section 7.2.3, the ARI system is required to function during all anticipated operating occurrences except those associated with a LOCA, HELB or SSE. Since these modifications only replace indicating lamps, the trip capability of ARI is not changed. Electrical separation, 10CFR50 Appendix R requirements, seismic mounting, power supply loading effects and other design considerations are incorporated in the design output documents.

II. No. These modifications add no new function to the ARI system nor do they remove any function of the system.

III. No. Technical Specification section 3/4.1.3 lists operability requirements for the Control Rod Drive mechanisms. These modifications reduce control loss associated with indicating lamp failures, thus increasing ARI system reliability.

8E/t No.! 91M8 I

DCPs 91-9001A thru 0 Certain circuit breakers in Unit 2 125V DC Class 1E distribution panels 2D614, 2D624; 2D634, 2D644 have fixed trip points which are oversized in relation to derated ampacities of the

'nd connected cables. 15 of the 16 breakers identified require replacement with breakers having trip ratings/characteristics which are compatible with cable derated ampacities, greater than connected load currents,. arid coordiiuMl with downstream subfuses.

I. No. The functions of the connected load systems are not modified by this action and failure mode and effects analysis for the replacement breakers is bounded by existing analysis provided in FSAR Table 8.3-21 for 1D614, 1D624, 1D634 and 1D 644.

II. No. Installation of properly sized and dedication tested replacement breakers will not modify the safety functions of the system loads. The proposed modification will not introduce any failure modes different than already analyzed in FSAR Table 8.3-21.

The proposed action does not modify the function or capacity of any connected safety systems and will increase the probability of safety systems remaining functional in the event of a sustained limited magnitude fault above the cable ampacity by opening the affected circuit. This limits, to the extent practical, damage to adjacent cables of other safety systems in the same raceway. The proposed action does not affect any Technical Specification/Bases and does not reduce the -margin of safety as defined in the basis for any Technical Specification.

sam mo.: 91-069 R NL 91-010 This safety evaluation analyzes the use of Nuclear Energy Services (NES) LPRM Bender. The LPRM bender is designed to produce a 180 degree bend at the midpoint of a discharged LPRM string while maintaining SSES's minimum shielding requirement of 5 feet above an irradiated detector. The resulting LPRM string will be "hair-pin" shaped; each-leg measuring approximately 22 feet in length. An LPRM of this shape allows for easy storage/accountability while placed over control rod brackets in the fuel pool.

BlMBEZ I. No. A Fuel Handling Accident as evaluated in FSAR 15.7.4 and 15A (event 41) is the only postulated accident which is applicable to this procedural change. The radiological consequences of dropping a fuel assembly are well within the guidelines of 1OCFR100. The effects of a dropped LPRM bender/LPRM are within the scope of a dropped bundle.

II. No. The LPRM bender has been extensively tested to ensure the LPRM integrity is maintained during the bending process. Should the LPRM break during the bending process, half of the LPRM would fall into the cavity/RPV area (the other half would remain in the LPRM benders clamp). An examination of the break area would then be performed, and retrieval of the fallen pieces would commence. In addition, the use of the LPRM bender will not violate PP&L's 5 foot shielding requirement for LPRM "hot ends". With the bending elevator in the full up position and the overhead crane aux hoist in the max up position, the shielding provided over the top guide ring will be 5'l l" when the LPRM is fully bent.

III. No. From Tech Spec 3/4.9.7, Crane Travel-Spent Fuel Pool, LCO, loads in excess of 1100 pounds shall be prohibited from travel over fuel assemblies in the spent fuel storage racks. Since the combined weight of an LPRM bender and LPRM is only 550 pounds, this procedural addition does not impact the acceptance limit.

~ER h+: 91-070

DCP 90-9008, Rev. 0 The change involves circuit modifications and installation of qualified isolation devices to prevent interaction between the'class 1E and non-Class 1E portions of the circuits for Main Steam Drain, Reactor Water Cleanup, Residual Heat Removal, Main Steam Radiation Monitors.

I. No. Although the proposed modification installs additional components to existing control circuits and instrument loops, there is no effect on the design basis of the various systems covered by this change.

'he proposed action protects the safety functions of class 1E components of the various systems under consideration. It does not change the operation of the instrumentation loops or the plant operator's interface with these loops or any other safety related equipment.

II. No. This modification provides acceptable isolation between Class 1E and non-Class 1E interface wiring as described in FSAR Section 8.1.6.1q and does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.

III. No. The proposed action involves changes to interfaces between Class 1E and non-Class 1E circuits and does not affect the margin of safety as defined in the basis for the Technical Specification of any safety related system. It has no direct impact on the operation of the containment isolation valves as indicated in Section 3/4.6.3 and Table 3.6.3-1 of the Technical Specifications. The isolation actuation instrumentation requirements for RWCU differential flow and RHR flow as indicated in Table 3.3.2-1 of the Technical Specification are not affected by the proposed modification. Similarly, it has no impact on the reactor protection system instrumentation for main steam line radiation as required by Table 3.3.1-1 of the Technical Specification. By isolating non-Class 1E circuit components and rewiring valve limit switches, the proposed action ensures that Class 1E circuits, instrument loops and isolation valves can perform their safety function in the event the non-class 1E interfacing circuits are subjected to impressed voltage faults.

DCP 91-9012, Rev. 2 This change replaces certain non-1E breakers in Unit 2 125V DC non-1E distribution panels 2D615, 2D625, 2D635, and 2D645 having fixed trip points which are oversized in relation to derated ampacities of the connected cables.

1 I. No. The replacement breakers are identical replacements for original equipment except that the trip ratings match the derated ampacities of the connected cables. The non-safety related functions of the connected load systems are not modified by this action.

II. No. Installation of properly sized and dedication tested replacement breakers will not modify the non-safety related functions of the system loads. The changes willnot introduce any failure modes different than already analyzed in FSAR Table 8.3-21.

III. No. This change will increase the probability of systems remaining functional in the event of sustained limited magnitude fault above the cable ampacity by opening the affected circuit.

$ ER~.: 91-071 R R  : 88-3045 A/B, Rev. 0 DE Rl KA E Replace cooling coils which are located in the RCIC pump room coolers as a solution to the corrosion damage in the cooling coils.

R:

I. No. The replacement cooling coils do not alter the function and performance of the original cooling coils. None of the design basis accidents discussed in FSAR Section 15 are affected. Failure modes associated with the replacement coils are the same as the original coils and do not become an initiating event for any design basis accident.

II. No. FSAR 9.2.5.1 and 9.4.2.2.'1 discuss the design basis for RCIC pump room coolers. Replacement of the coils will neither alter the function nor the performance of the RCIC pump room coolers and they meet the design basis discussed in these FSAR sections.

III. No. Tech Spec Section 3/4.7.3, relating to RCIC pump operability were reviewed for applicability. These modifications neither reduce the operability of any equipment nor require additional surveillance to ensure plant safety.

$ ER Pgl.: 91-073 R E: DCP 89-3021C, Rev. 2 F H of the heat trace for the containment atmosphere control isolation valves to assure E'odification the valve environmental qualification will not be invalidated by heat trace miscalibration or failure and to assure that credible failure(s) of the nonqualified heat trace system does not have the potential to affect redundant isolation valves. Sample, lines for the H2/02 analyzers and containment radiation monitoring panels aie also heat traced.

I. No. The proposed action assures that the heat applied to the H2/02 Analyzer and CRM Panel Piping sensing lines and valves maintains the same temperature above the dew point and does.not exceed the environmentally qualified temperature of the Target Rock solenoid containment isolation valves. This permits the analyzers, panels and isolation valves to perform their intended safety functions.

H. No. The proposed action does not alter the function of the heat trace on the H2/02 Analyzer and CRM sampling lines. Temperatures on the piping will continue to be maintained above the dew point of the sampled air during normal operation with reduced heat trace cycling and thermal stress to the piping and valves.

III. No. The Reactor Coolant Leakage Detection System identified in Section 3/4.4.3 of the Technical Speci6cation is required to monitor and detect leakage from the reactor coolant pressure boundary. The proposed action does not affect the detection system's ability to perform its function.

The operability of the primary containment isolation valves identified in Section 3/4.6.3 of the Technical Specification ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment. The proposed action assures operability of the valves to meet their isolation function by maintaining the valve temperatures below their environmentally qualified limit.

~ER~.: 91474 R RE  : DCP 88-3018S, Rev. 1 Installation of valves, piping, and a new refrigerant receiver tank to eliminate refrigerant leaks and maintenance problems in the Unit 2 direct expansion (DX) refrigerant system.

I. No. FSAR Chapter 15 "Accident Analysis" does not specifically discuss any accident related to the DX System. FSAR Section 9.4 (FSAR Table 9.4-5) presents a Failure Mode and Effect Analysis for the RB HVAC system'and provides a functional description of the ESWGR Room Cooling System. This modification does not affect the failure mode and effect analysis as it relates to the ESWGR Room Cooling system.

II. No. The addition of upgraded materials and flexible metallic'hoses is intended to eliminate refrigerant leaks and improve overall system reliability. The emergency cooling function will not be changed in any way by this modification. No additional failure modes, as described in FSAR Table 9.4-5, need to be analyzed for this system as a result of this modification.

III. No. There are no formal Technical Specifications related to the ESWGR Room Cooling System. The SSES TEAM Manual identifies an administrative Tech Spec for this system, but the proposed action does not reduce the margin of safety associated with the basis for this requirement. This modification does not alter the emergency cooling function of this system in any way, and is intended to improve overall system reliability and equipment accessibility.

$ ~~.: 91-075

Procedure ES-161(261)401 DE RI F HA E:

To describe the procedural steps necessary to utilize reactor water cleanup (RWCU) as an alternate mode of decay heat removal when other heat removal systems are unavailable.

KMbLARY:

I. No. The blowdown mode of RWCU is used as a mitigating action during an already existing accident condition, and because of the RWCU equipment willbe operated within its design limits (pressure, temperature, and flow), there is no increased, probability of accident occurrence.

II. No. RWCU is used after the accident occurs. Use of RWCU requires two bypasses to be performed. The first is to bypass the RWCU Non-Regen Heat Exchanger High Output Temperature logic described in FSAR 7.7.1.8.3.2. The second is the bypass of logic which closes the valve in RBCCW piping which supplies RWCU Non-Regen. Heat Exchanger described in FSAR 9.2.2.2. The intent of both of these logics is satisfied.

III. No. The action does not involve any Tech Spec margin of safety.

TP-104-010 Replacement of the unqualified relay socket with a qualified relay socket for the.K11 BX3 relay and associated test procedure.

I. No. This change only replaces unqualified sockets with qualified sockets. However, the performance of this work disables the Unit 1, Division 1 BOP LOCA load shed logic scheme. This disabling does not affect the operation of the 13.8Kv loads, it simply disables the automatic load shedding feature associated with non-1E equipment after a LOCA.

Therefore, the action of disabling the BOP LOCA Load Shed Scheme for the Unit 1B 8c 1D Buses, for approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, during the performance of the K11BX3 relay socket replacement is considered both safe and prudent.

II. No. The BOP LOCA Load Shedding Scheme is not specifically described in the FSAR.

IH. No. Surveillance requirements associated with the operability of the K11BX3 relay are satisfied as defined in Unit 1 Tech Spec 4.3.3.2, Function SC.

$ g~.: 91-077 R REFER E: TP-104-009 E RI Replacement of the unqualified relay socket with a qualified relay socket for the K11 AX3 relay and associated test procedure.

I. No. This change only replaces unqualified sockets with qualified sockets. However, the performance of this work disables the Unit 1, Division I BOP LOCA load shed logic scheme. This disabling does not affect the operation of the 13.8Kv loads, it simply disables the automatic load shedding feature associated with non-1E equipment after a LOCA.

Therefore, the action of disabling the BOP LOCA Load Shed Scheme for the Unit 1A and 1C Buses, for approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, during the performance of the Kl1 AX3 relay socket replacement is considered both safe and prudent.

II. No. The BOP LOCA Shedding Scheme is not specifically described in the FSAR.

III. No. Surveillance requirements associated with the operability of the Kl1 AX3 relay are satisfied as defined in Unit 1 Tech Spec 4.3.3.2, Function SC.

ging.~.: 91-078 DCP 88-3018T, Rev. 1 Modifications to the direct expansion (DX) refrigeration system to repair refrigerant leaks and improve maintainability.

I. No. FSAR Chapter 15 "Accident Analysis" does not specifically discuss any accident related to the DX System. FSAR Section 9.4 (FSAR Table 9.4-5) presents a Failure Mode and Effect Analysis for the-RB =HVAC.-system and provides a functional description of the ESWGR ROOM Cooling Sy'tem'. This modification does not affect the failure mode and effect analysis as it relates to the ESWGR Room Cooling System.

II. No. The emergency cooling function will not be changed in any way by this modification. No additional failure modes, as described in FSAR Table 9.4-5, need to be analyzed for this system as a result of this modification.

III. No. There are no formal Technical Specifications related to- the ESWGR Room Cooling System. The SSES TEAM Manual identifies an administrative Tech Spec for this system, but the proposed action does not reduce the margin of safety associated with the basis for this requirement.

fgR~Nl.: 91-079 TP-234436, R . 1 To allow tie-in of a temporary outage chiller to the Reactor Building Chilled Water system during a Unit 2 outage.

R:

I. No. Installation of this temporary system will not impact operation of primary containment isolation, the only safety function of RBCW system. Normal operation of the reactor building ventilation system will not be impacted by installation of the temporary chiller. Additionally, this equipment will be installed during a period when secondary containment is noi required.

II. No. The installation of this temporary system does not adversely impact the operation or safety function of any existing installed equipment.

III. No. See I. and II. above.

DCP 90-9101, Rev. 0 Replace the one hour rated Kaowool fire wrap with one hour rated TSI Thermo-Lag fire wrap on the Unit 1 safety related 4.16Kv bus connected to the diesel generators to meet ampacity requirements.

I. No. 'ince the derated cable ampacity of the cables is greater than the ampacity required for mitigating a design basis accident, the installation of the TSI Thermo-lag fire wrap does not reduce the reliability of the cables to perform their safety function.

II. No. The action to replace the Kaowool fire wrap with TSI Thermo-Lag fire wrap on

'conduits C1FO33 and CIF034 does not interfere with the logic, control or operation of any plant systems or components. There is no impact on the periodic maintenance, calibration or surveillance activities of any plant systems or components.

III.

~ No.~ The action permits the required surveillance testing per Section 3/4.8.1 to be performed without the potential future degradation of the cables. This action assures the present margin of safety is not reduced. The fire rating of the TSI Thermo-Lag fire wrap is equivalent to the Kaowool fire wrap so that the margin of safety required by Technical Specification 3/4.7.7, Fire Rated Assemblies, is not affected.

'I 91-012 Hot taps into Reactor Building Closed Cooling Water (RBCCW) and Turbine Building Closed Cooling Water (TBCCW) as part of the In Line Process Instrumentation Upgrade Project.

I. No. The review of FSAR Section 9.2.3 indicates that the TBCCW system performs no function that is important to safety. The review of FSAR Section 9.2.2 indicates that the RBCCW system containment penetrations and isolation valves perform containment isolation functions which are important to safety. The taps are located on the side or bottom of the pipes to reduce the chance of chips falling into the process pips. Additionally, the RBCCW and TBCCW head tanks assure positive system pressure to flush cutting chips out of the system and into the hot tap apparatus.

II. No. In accordance with FSAR Sections 9.2.2 and 9.3.2, the RBCCW and TBCCW lines are not required to function after an accident. As previously discussed, the TBCCW and RBCCW systems willbe protected during the hot tapping procedure to minimize the introduction of system crud or cutting chips into the process plplllg.

III. No. Technical Specification Sections 3/4.6.1 and 3/4.6.3 contain requirements for the RBCCW Primary Containment Isolation Valve leakage integrity and operability.

The RBCCW hot taps which are being performed do not make any direct changes to these valves. The Technical Specifications do not address the TBCCW system since it has no function which is important to safety.

0

$ ER~.: 91482 DCP 90-3055C, Rev. 0 Replace existing control switch HS-28723B, GE Type CR2940UB203B, located in Non-Class 1E Panel 2C279 with control switch type CR2940UB203 as part of the Chiller Upgrade Project.

I. No. The replacement control switch is specified to the same criteria as the existing control switch. However, the new control switch has a Cam operator 03 with contact arrangements which allow the circuit to remain energized when the switch position is changed from "auto" to "start". The switching logic willbe improved, eliminating unnecessary trip of pumps 2P235B and 2P217B when control switch HS-28723B is placed from "auto" to "start" position.

II. No. This modification adds no new function to the Reactor Building Chilled Water System as described in FSAR Sections 9.2.2 and 9.2. 12.3.

III. No. The operation of the Reactor Building Chilled Water System has no safety related function. Although isolated portions of chilled water piping of the containment are safety related, these portions are not affected by this modification.

$ER Pgk.: 91-083 gg DD333-3D33, 3 . D Replacement of Unit 1 Reactor Building Sample Station (RBBS)(1C210) and Turbine Building Sample Station (TBSS)(1C132).

I. No. The FSAR sections 9.3.2.3, 9.2.2.1, 9.4.4.1, 9.3.1.1;1 and 9.2.9.1 contain the functions and requirements of the interfacing and Process Sampling Systems.

They state that the systems have no safety-related functions. These sections were reviewed and it was determined that this change does not change the function of the systems or the operator's interface with these systems or any safety system.

II. No. Malfunction of the sample stations or modulating damper is not a malfunction of a different type than previously evaluated because it does not have an adverse affect on the safety function of the reactor building recirculation mode. Based on the review of FSAR Section 9.4.4 the Turbine Building HVAC system is a non-safety related system which is not required to operate to mitigate the consequences of an accident.

III.

~ No.~ This modification does not reduce the margin of safety since the interfaces being installed for the TBSS and RBSS are not safety-related and do not affect any margins of safety described in the Technical Specifications. The acceptance limits for the RCS chemistry as noted in the Technical Specification Section 3/4.4.4 will not be affected as a result of modification performed by this change.

Ega,MQ.: 91484 DCP 90-3029, Rev. 1 The existing Unit 2 Reactor Building Sample Station (RBSS) (2C210) and Turbine Building Sample Station (TBSS) (2C132) are being entirely replaced and also relocated so that their installation may be completed non-outage.

I. No. The modification replaces process sample lines and adds cooling water piping from the RBCCW system. In accordance with FSAR Sections 9.2.2.3 the sample lines and RBCCW lines are-not required to function after an and'.3.2.3 accident, and the installation does not impact the safety isolation function of containment isolation valves HV21314, HV21346, HV21313, and HV21345.

II. No. The sample lines are used only for chemistry monitoring and provide no direct control functions. Loss of sample line flow, due to blockage or breakage, results in loss of chemistry monitoring until corrected but does not create any possibility for unevaluated accidents or malfunctions. Breakage of lines of this size is addressed in FSAR Section 15.3.2 and the results found acceptable.

The Technical Specifications were reviewed to determine ifany margins of safety or limiting conditions for operation are defined for the Process Sampling and the RBCCW Systems. No Technical Specification requirements, including safety margins, are established for the Process Sampling system. Technical Specification Sections 3/4-6.1 and 3/4-6.3 contain requirements for the RBCCW Primary containment Isolation Valve leakage integrity and operability. This change does not make any direct changes to these valves.

Egg~;91-085 RE-2TY-008, Rev. 1 This procedure is used to investigate and repair the persistent leak through control cell 38-31.

L No. PSAR Sections 15.0 "Accident Analysis", 6.5.3.2 "Secoiidary Containment", and 9.4.2 "Reactor Building Ventilation" were reviewed. Zone I and Zone III Secondary Containment will not be impacted by this procedure and therefore will function as designed. None of the above referenced sections discuss accidents requiring Secondary Containment on a unit that has been defueled with the fuel pool gates installed.

II. No. See I. above. Also, there are no accident scenarios that could be postulated when the reactor. vessel is defueled and the fuel pool gates are installed. The worst case scenario could be an unisolable leak through the bottom of the vessel. Zone III Secondary Containment would be operable to isolate the Refueling Floor should an airborne situation occur.

III.

~ No.~ Technical Specification Section 3.6.5.1 "Secondary Containment Integrity" will not be affected for Unit I and is not applicable for Unit 2. Tech Spec Sections 3.9.10.1 and 3.9.10.2 address the requirements for CRD removal. Since the RPV is void of fuel bundles, the possibility of an inadvertent criticality is removed, therefore the margin of safety as defined in the basis is not reduced.

SEKZQ-'1486

Procedure OP-257405, Rev. 1 This procedure addresses prerequisites, precautions, and both operational and safety impacts for taking UPS panel 2Y629 out of service while cleaning cabinets 3 and 4 of the vital UPS 2D666.

I. No., .The only, accident identified in Chapter 15 that would be applicable in- this condition is a fuel move accident. A prerequisite exists in GP-257-065 that no fuel movement is *allowed. The loads powered 'during this'rocedure are not required for the safe shut down of the plant, effect the containment barriers or capabilities of heat removal systems.

H. No. This procedure OP-257-005, has been performed in the past. The revision one enhancements (temporary power to a number of circuits and jumpering of the refueling bridge motion interlocks) are being implemented since a long term loss of vital UPS would have a negative effect on outage activities. The individual loads have been reviewed and the proposed actions do not create a possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR.

IH. No. Panel 2Y629 is a non-class 1E load panel and is not required or discussed in the basis for Technical Specification 3/4.8.3, onsite power distribution systems.

SELM-'1487

DCP 90-9079, Rev. 0 Reinstate the Feedwater Heater High Level Alarm which was disconnected by a bypass.

I. No. FSAR Sec. 10.4.10.1 states "The ... Feedwater Heater Drain and Vent System has no safety related function." The accident analysis of FSAR Chapter 15 is not affected by this modification. The reconnection of a local alarm function does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the.

FSAR.

II. No. The change enhances the Feedwater Heater level controls by providing the operator an alarm on rising water level. This change adds no new function to the Feedwater Heater Vents & Drains nor does it remove any function of the system.

III. No. Technical Specification does not address Feedwater Heater water level. This modification only reinstates a local alarm function.

SggjK.: 91-088

DCP 86-7026, Rev. 4 Replace Static 0-ring (SOR) pressure switch used in RHR "B" loop with a Barton Model 288A with a lower setable range. The Barton Model 288A in the "A" loop willbe placed with another Model 288A, with a lower range to match the "B" loop.

I. No. RHR system is discus!uzi in FSAR 7.1.2a1.25, 7.1.la.2, 7.4.1, 7.4.1.3, and 7.6.1.a.3. The switches are not specifically addressed in any Chapter 6 or 15 safety analysis. The valves are containment isolation valves and are assumed to close or be closed during a LOCA. This function is not affected by the replacement of the pressure switches. Replacement of the existing pressure switches does not change the system design intent. The function of the switches remains the same in that the safety isolation setpoint will be set to close the valves at a value less than or equal to that listed in the Technical Specifications.

The failure modes of the replacement switches are no different than that of the existing switches. A failure resulting in switch actuation, when pressure is low, will cause a loss of Shutdown Cooling. Loss of Shutdown Cooling is an analyzed event in Chapter 15 of the FSAR (Event 18).

III. No. The replacement of the existing Barton and SOR switches does not affect Technical Specifications Bases 3/4.3.2: Isolation Actuation Instrumentation or 3/4.4.9: "Residual Heat Removal". The switches are addressed in Technical Specification Table 3.3-2-1: "Isolation Actuation Instrumentation", which states that the switches must be operational in Conditions 1, 2 and 3 or else the system isolation valves must be declared inoperable. This modification willenhance the operation of the valves and RHR system.

/~Rgb.: 91-089 R RE  : SCP J91-1013, Rev. 0 Raise the MSIV low pressure setpoints for pressure switches PSL-B21-2N015A, B, C and D from their present values back to their original values.

~AR: I I. No. Because-a previous:circuit.modification eliminated spurious actuations without changing system performance, restoring the setpoints'to their original value does not inciease the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the SAR.

H. No. The setpoints were set abnormally low to eliminate spurious false isolations. The setpoints were lowered without affecting Technical Specification setpoints.

Because the reason for lowering the setpoints has been eliminated without affecting system performance, restoring the original setpoints does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the SAR.

III.

~ No.~

~

This setpoint change raises a low pressure setpoint. This moves the setpoint away from the Allowable Value which increases, not decreases, the Margin of Safety.

DCP 90-9083, 90-9084, Rev. 1 Remove the solid state auxiliary relay and its Class lE 125V DC power supply from HPCI lube oil tank level switches LSHL 15660 and 25660."

I. No. This change does.not affect the operation of the HPCI or any related equipment or decrease the reliability of, the level alarm.

II. No. Since only non-safety related HPCI functions are affected, no new modes of

. operation or system failure modes are cre'ated. Additionally,'the Class 1E 125V DC system is simplified by the elimination of unnecessary non-1E loads.

III. No. Referring to Tech Spec 3/4.3.3 and 3/4.5.1 for the HPCI system and 3/4.8.2 for the 125V DC system, this change does not negatively affect the instrumentation, control, performance or availability of the HPCI system, the 125V DC system, or any related equipment.

~8~.: 91-091 DCP 90-3055D, Rev. 0 Reduce the collection of service water silt and debris in the root valves for the condenser pump low fiow switch on the Unit 2 Reactor Building Chiller Service Water Sensing lines by replacing the existing root valves (1" globe valves) with new full port 1" ball valves.

$ggM4,RY:

I. No. As discussed in FSAR Section 9.2.12.3, the only safety related functions of the RBCW System are performed by the primary containment piping penetrations and containment isolation valves. The proposed action does not affect safety related portions of the RBCW System. The installation of the new root valves willaffect only a portion of the SW System associated with the chiller condenser water circulating pump.

II. No. The modification will not alter the intended function of the RBCW System described in FSAR Sections 9.2.12.3.2.3 and 9.2.12.3.3. The proposed action does not affect operability of the RBCW system nor introduce any new mode of operation. The present operation of the chiller is not changed and no new failures or malfunctions are created by installation of the new root valves.

III. No. The RBCW System and the SW System operability does not contribute to any margin of safety as defined in the basis of any Technical Specification. SSES Technical Specifications (Unit 2) Section 3.6.1.7 and Table 3.6.3-1 and their bases were reviewed for applicability. Section 3.6.1.7 contains the requirements for average drywell temperature during reactor operating conditions 1, 2 and 3.

Table 3.6.3-1 includes RBCW primary containment isolation valves, which will not be affected by the proposed modification.

~E~.: 91-092 R R  : DCP 90-3055F, Rev. 0 DE PTI Eo 1

Existing temperature modulating butterfly valves on the Unit 2 Reactor Building Chiller Condenser are replaced with new control valves having high turndown ratio and incre ised anticavitation protection to improve chiller recirculation loop flow control.

I. No., As discussed in FSAR Section 9.2.12.3, the only safety related functions of the RBCW System are performed by the primary containment piping penetrations and containment isolation valves. The proposed action does not affect safety related portions of the RBCW System. The installation of the new temperature control valves and their temperature controller will affect only a portion of the SW System associated with the chiller condenser water circulating pump.

H. No. The proposed modification will install new temperature control valves and their temperature controller which do not alter the intended function of the SW system and the RBCW System as described in FSAR Section 9.2.1.1 and 9.2.12.3. The existing method of the service water temperature control by recirculation of the service water as described in FSAR Section 9.2.1.2 will be unchanged.

III. No. The RBCW System and the SW System operability does not contribute to any margin of safety as defined in the basis of any Technical Specification. SSES Technical Specifications (Unit 2) Section 3.6.1.7 and Table 3.6.3-1 and their bases were reviewed for applicability. Section 3.6.1.7 contains the requirements for average drywell temperatures during reactor operating conditions 1, 2 and 3.

Table 3.6.3-1 includes the RBCW primary containment isolation valves, which will not be affected by the proposed modification.

5'.: 91-093

'CP 90-3083, Rev. 0 This modification will replace the existing Diesel Generator IAC66B4A single phase induction disk overcurrent relay, located in PNLs OCS19A, B, C & D with a Class lE, single phase, ASEA Brown Boveri (ABB) type ITS-50D definite time overcurrent relay.

I. No. The AC Power Systems and DC Power Systems design basis and function, as previously evaluated in sections 8.3.1 and 8.3.2 respectively of the FSAR, is not altered by this modification..

II. No. This modification replaces the existing electromechanical IAC66 relay with a solid state definite time overcurrent relay. This modification does not modify the function or the design philosophy of the existing annunciation system. The replacement relay will allow more accurate pick-up setting, therefore reliability and operation of the annunciation will be improved.

IH. No. This modification does not affect the instrumentation, control, performance or availability of Diesel Generators A, B, C & D. The "Near Full Load" alarm is not discussed and does not affect Technical Specification Section 4.8.1.1.2 which provides the Diesel Generators operability surveillance requirements.

$ gR ~.: 91-094 FERE: SCP J-91-1005, 1006, 1007 & 1008 RJPTI The action is to provide engineering review of switch setpoints for the Backup Water Fire Protection System (BWFPS) at the Well Water Pump House.

$ g~ARY:

I. No. The BWFPS is not discussed in the FSAR. There are no design basis accidents which address operation of the BWFPS. The design basis accidents contained in FSAR Chapter 15 were reviewed.

II. No. The action does not affect the design or operation of the BWFPS. The task adds no new functions to the BWFPS. Nor does it remove any existing functions of the system. Rather, it provides a documented basis and calculation for the BWFPS setpoints.

III.

~ No. ~ There is no discussion of the BWFPS in the basis of any Technical Specification.

Egg~;91-095

DCP 90-3026 Existing Unit 1 Reactor Building Sample Station (RBSS)(1C201) and Turbine Building Sampling Station (TBSS)(1C132) are being replaced to update instrumentation an monitoring systems.

I. No. FSAR Section 9.2.2.1, 9.3.1.1.1, 9.2.9.1, 9.4.4.1 and 9.3.2:3 were reviewed.

They state that all systems, except Reactor Building ventilation and RBCCW, have no safety-related functions. The reference sections were reviewed and it was determined that this change does not change the function of the systems or the operator's interface with these systems or any safety system. The review of FSAR Sections 6.2.3.2.3.1, 6.2.4.3.3.5 and 9.2.23 indicate that the RBCCW System containment penetrations and isolation. valves perform containment isolation functions which are important to safety. The installation of new piping under this DCP does not change the probability of occurrence or the consequences of an accident or malfunction of these components over that which was previously evaluated.

II.

~ No.~ Malfunction of the sample stations or the modulating damper does not have an adverse affect on the safety function of the reactor building recirculation mode.

Based on the review of FSAR Section 9.4.4 the Turbine Building HVAC system is a non-safety related system which is not required to operate to mitigate the consequences of an accident. The additional electrical load added by this change does not exceed the load margin provided in FSAR tables 8.3.1 through 8.3.5.

III. No. This modification does not reduce the margin of safety since the interfaces being installed for the TBSS and RBSS are not safety-related and do not affect any margins of safety described in the Technical Specifications.

0 ggR ~.: 91-096

DCP 90-9031A, Rev. 0 This change will eliminate the Unit 1 alarm TAHL-11653, Caustic DBution Water Heater Temperature High/Low, for the Condensate Demineralizer.

I. No. The change deletes an alarm which monitors an equipment that is no longer in service. FSAR Section 10.4.6 - Condensate Cleanup System and FSAR Chapter 15.0 - Accident Analysis were reviewed to determine the impact of this modification. The Chapter 15 analyses do not explicitly address the Condensate Cleanup System. FSAR Section 10.4.6.1 states the design bases for the system.

None of the design bases will be affected by this modification.

II. No. This change does not impact the logic, control or operation of any plant systems.

The Condensate Cleanup system performance requirements are met with the deletion of the subject alarm.

III.

~ No.~ The Unit I Technical Specifications have been reviewed for impact of the proposed modification, specifically Sections 3.0 and 4.0- Limiting Conditions for

-

Operation and Surveillance Requirements and Section B3/4 - Bases for Sections 3.0 and 4.0. The proposed modiGcation deletes an alarm which monitors equipment that is no longer in service.

~8R~N ~ !91-097 NL-91-020, Rev. 0 Documents are being revised to reflect the organizational changes associated with transferring the Shift Technical Advisor and Reactor Engineering Groups from the Technical Section to the Operations Section.

I. No. FSAR Section 13.1.2.2.7 describes the Technical Organization with the Plant Engineering Supervisor, Reactor Engineering (R.E.) Supervisor, Lead Shift Technical Advisor and Senior Compliance Engineer reporting directly to the Technical Supervisor. Although the R.E. Supervisor and Lead STA will now report to the Operations Supervisor, the Reactor Engineering Group and STA Group functions, as described in this FSAR section will not be changed.

II. No. The proposed changes are only organizational in nature and will not affect the possibility for accidents and malfunctions of a different type than previously evaluated in the FSAR.

IH. No.

~ ~ Technical Spec. Section 6.0, Administrative controls, requires an STA on shift and designates the Reactor Engineering Supervisor or the Unit Reactor Engineer as a member of the PORC. These requirements will not be affected by this change, and T.S. Section 6.0 does not have a Tech Spec Bases Section which could be affected.

Tech Spec. Section 6.2 describes the organizational characteristics the NRC has determined to be important to safety (ref: Generic Letter 88-06). The proposed changes do not affect Tech Spec Section 6.2.

Egg~.: 91-098

Nl 91-018, Rev. 0 The proposed action is to use a cation resin in the Reactor Water Cleanup (RWCU) precoat material with a functional group that does not degrade into sulfate or any other ionic species known to be detrimental to BWR chemistry.

I. No. Section 15.7 of the FSAR analyzes Radioactive Release from subsystems and components. Since the amount of activity within the system is not expected to change, this section is not affected. The new resin is expected to improve vessel chemistry and reduce the probability of chemistry induced failures of all equipment in contact with coolant.

II. No. As stated above, this action is expected to raise the purity of the coolant, by reducing baseline sulfate levels slightly. This reduction of sulfate and conductivity levels and the relationship between conductivity, sulfate and other vessel chemistry parameters on IGSCC are discussed in Section 5.2.3.2.2 of the FSAR, BWR Chemistry of Reactor Coolant.

The Tech Spec bases for requiring conductivity measurements is to be able to determine abnormal conditions. From the Tech Spec Bases 3/4.4.4 Chemistry

- "When the conductivity is within limits," (K I umho/cm), "the pH, chlorides, and other impurities affecting conductivity must also be within acceptable limits."

This action, as stated previously, is expected to improve vessel coolant quality.91-099

'L-91-019, Rev. 0 To determine the acceptability of three bladder assembly designs as installed in the accumulators on the discharge line of each Standby Liquid Control (SLC) pump.

I. No. SLC system remains functionally identical to the original system. The SLC system operation is discussed in the FSAR chapter 9.3.5.3. As a special system, SLC is not required to'perform its intended function during design basis events

"(i.e. LOCA, LOOP, etc); it is however, required to be reliable to a degree consistent with its role as a special safety system". As such, the pumps are powered from an emergency diesel generator and the components of the system are seismically qualified. (Note: SLC electrical components and piping have not been affected by the change.)

II. No. The SLC system is used in three-special plant capability demonstration events cited in Appendix A of Chapter 15. Specifically events 51, 52 and 53 which are extremely low probability non design basis postulated incidents. The analyses provided therein demonstrate additional plant safety consideration far beyond reasonable and conservative assumptions (FSAR Section 9.3.5.4). The changes to the accumulator charging connections do not incorporate any new design base or alter the existing operational scheme of the SLC system.

III. No. Technical Specification Sections 4.1.5-c and 3/4.1.5 require each SLC pump to demonstrate a minimum flow of 41.2 GPM at a minimum discharge pressure of 1,190 psig. Additionally delineated are requirements for injection time, boron concentrations, boron volume, etc. The changes to the accumulators do not affect any of the Technical Specification parameters.

~:

$E~~.: 91-100 R BCPL-3a!4,90 3023 &9MO37 These changes instaH Units 1 8c 2 turbine and reactor building sample station local acquisition modules and the central data handler and engineering analysis processor.

i I. No. Per FSAR 9.3.2.3 the process sampling system:is not required to function during an accident nor is it required to either prevent or mitigate the consequences of an accident. Also, per FSAR 7.7.1.7.1.2, the plant computer system is non-'safety related and therefore is not essential for the safety of the plant.

II. No. These changes install non-safety related raceway and cables to make this equipment operational. There are no failure modes of equipment important to safety that could cause new unanalyzed type of malfunction as the results of these modifications. The only interface with equipment important to safety is with the diesel generators. Failure of the diesel generators could render portions of the Plater Chemistry Data Acquisition System inoperative but could not cause a new unanalyzed type of plant malfunction.

III.

~ No.~ Technical Specifications do not give any requirements for the Process Sampling System or any of the systems interfacing with them.

0

DCP 90-3070, Rev. 1 DE RlPTI H E Add components that provide auxiliary control to the damper circuit of the Control Structure HVAC (CSHVAC).

~igARY:

I. No. This modification is being performed to ensure compliance with 10 CFR 50, Appendix R. As stated in FSAR 9A.1.3, equipment such as fans and dampers are redundant. In this change the components added are redundant, that is they are installed in each of the redundant fan/dampers circuits. As analyzed in Tables 9.4-16, 17, 18, and 20 the failure of any fan or damper will have no effect on the system because the standby fan willautomatically start, the same statement holds true for the transformers/relays added by this change.

II. No. This modification does not create any additional scenarios which could prevent the CSHVAC system from fulfillingits design requirements, either from normal or emergency operation.

The addition of auxiliary damper control logic maintains the redundant HVAC systems as described in FSAR Sections 9.4.1.3. The design of the modification conforms to the design evaluation of the Control Structure HVAC as stated in FSAR Section 6.4.4.

III. No. The Control Room Emergency Outside Air Supply System (CREOASS) (3/4.7.2) is addressed by the Technical Specifications for Unit 1 and Unit 2. This modification does not affect the CREOASS system as described in FSAR 9.4.1.2.4.

R RE RE: DCP 90-3084M, Rev. 0 Replace the existing 250V DC fuses in the battery chargers, load centers and motor control centers within the Unit 1 250V DC system.

I. No. The replacement fuses and fuse blocks are specified to the same criteria as the existing fuses and fuse blocks with the exception that they are qualified to a minimum of 300V DC and 20,000 amps interrupting capacity in order to assure interruption of faulted circuits and prevent further propagation of the failures.

The failure mode and effect analysis provided in FSAR Table 8.3-22 for the 250V DC System is not impacted by this change.

H. No. Installation of properly rated and sized fuses and fuse blocks will not modify the safety functions of the system loads. The proposed modification will not introduce any failure modes different than already analyzed in FSAR Table 8.3-22.

III. No. The proposed action does not affect any Technical Specification/Bases. The proposed action does not modify the function or capability of any connected safety systems and will increase the probability of safety systems remaining functional in the event of a fault by isolating the affected circuit.

~ER j~.: 91-103

DCP 90-3085M, Rev. 0 DE RIPTI F H E:

Replace the existing 250V DC fuses in the battery chargers, load centers and motor control centers within the Unit 2 250V DC system.

fARY:

I. No. The replacement fuses and fuse blocks are specified to the same criteria as the existing fuses and fuse blocks with the exception that they are qualified to a minimum of 300V DC and 20,000 amps interrupting capacity in order to assure interruption of faulted circuits and prevent further propagation of, the failures.

The failure mode and effect analysis provided in FSAR Table 8.3-22 for the 250V DC System is not impacted by this change.

II. No. Installation of properly rated and sized fuses and fuse blocks will not modify the safety functions of the system loads. The proposed modification will not introduce any failure modes different than already analyzed in FSAR Table 8.3-22.

III.

~ No.~ The proposed action does not affect any Technical Specification/Bases. The proposed action does not modify the function or capability of any connected safety systems and will increase the probability of safety systems remaining functional in the event of a fault by isolating the affected circuit.

DCP 90-3088, Rev. 0 Provide overpressure protection for the Emergency Service Water (ESW) system by the installation of a 1-inch safety relief valve.

I. No. The safety relief valve is designed and constructed to meet the requirements of ASME Section III, Class 3. This code classification is applicable to valves, piping and other components in the ESW system as specified in the FSAR Table 3.2-1, and ensures the integrity of the system's pressure retaining boundary.

II. No. Because the valve is located on a short length of piping, in close proximity to the existing ESW piping, a failure of the new valve or inlet piping to the valve does not create a greater potential for equipment damage due to water spray or flooding than previously considered in FSAR Sections 3.6.1 and 3.4, respectively.

III. No. The margin of safety established in Technical Specification sections B 3/4.7.1 and B 3/4.8.1 is based on a minimum level of operability of the Diesel Generator and Emergency Service Water systems which considers that a single failure of one of the aligned diesels or the loss of one flow path or one pump does not prevent the systems from performing their safety functions. The potential failure modes of the safety relief valve and the resulting water losses does not prevent sufficient cooling water from being supplied to one ESW loop.

R R  : DCP 90-9032B, Rev. 0 Deletion of Condensate Demineralizer nuisance alarms: LAL-21631, Acid Storage Tank Low Level, LAL-21677, Caustic Storage Tank Low Level, and TAHL-21653, Caustic Dilution Water Heater Temperature High/Low.

$ QgM4.RY:

I. No. FSAR Section 10.4.6 - Condensate Cleanup System and FSAR Chapter 15.0-Accident Analysis were reviewed to determine the impact of this modification.

The Chapter 15 analyses do not explicitly address the Condensate Cleanup System or take credit for the Condensate Cleanup System in the mitigation of a Design Basis Accident (DBA). FSAR Section 10.4.6.1 states the design bases for the system. None of the design bases will be affected by this modification.

II. No. This change does not impact the logic, control or operation of any plant systems that perform a safety function. FSAR Section 10.4.6 - Condensate Cleanup System and Chapter 15 - Accident Analysis were reviewed for the impact of the proposed change. The Condensate Cleanup System performance requirements are met with the deletion of the subject alarms. The alarms monitor equipment which is no longer in service.

III. No. Technical Specifications have been reviewed for impact of the proposed modification, specifically Sections 3.0 and 4.0 - Limiting Conditions for Operation and Surveillance Requirements and Section B3/4 - Bases for Sections 3.0 and 4.0. The proposed modification will not affect the operation of any system as described in the Technical Specifications.

$ ER~.: 91-106 R E  : DCP 90-9032D, Rev. 0 The change will eliminate Circulating Water nuisance alarms: AAH-23233, Unit 2 Circulating Water ph-High, AALL-23233, Unit 2 Circulating Water ph-Low, AALI 23233, Unit 2 Circulating Water ph-Low-Low, and CAHL-23236, Unit 2 Circulating Water Conductivity.

SUMhBILY:

I. No. FSAR Section 10.4.5 - Circulating Water System and FSAR Chapter 15.0-Accident Analysis were reviewed to determine the impact of this modification.

The Chapter 15 analyses do not explicitly address the Circulating Water System or take credit for the Circulating Water System in the mitigation of a Design Basis Accident (DBA). FSAR Section 10.4.5.1 states the design bases for the system. None of the design bases will be affected by this modification.

II. No. FSAR Section 10.4.5 - Circulating Water System was reviewed for the impact of the proposed change. The Circulating Water System performance requirements are met with the deletion of the subject alarms. The proposed modification will not impact the operation of any plant systems that perform a safety function.

The Unit 2 Technical Specifications have been reviewed for impact of the proposed modification, specifically Sections 3.0 and 4.0- Limiting Conditions for Operation and Surveillance Requirements and Section B3/4 - Bases for Sections 3.0 and 4.0. The proposed modification will not affect the operation of any system as described in the Technical Specifications.

~ER~N.: 91-107 g  : NL-91-021, Rev. 0 Replacement of an existing resistor and capacitor on the Rod Sequence Control System (RSCS)

Test 2B card with a resistor, capacitor and diode to increase reliability of the circuit to restrict control rod insertion to one notch at a time.

I. No. The FSAR, including Sections 15.4.1, Rod Withdrawal Error-Low Power and

15. 14.2, Rod Withdrawal Error-At Power, has been reviewed.

Other than operator errors, the above sections deal with continuous rod withdrawal and movement of control rods beyond banking limits. The RSCS signals which prevent continuous rod withdrawals under specified conditions and which enforce banking limits are generated in circuitry independent of the generation of the operator insert requested signals.

II. No. Changing the insert circuit as described above will have no effect on the possibility of a multiple notch withdrawal. Therefore, the proposed circuit change does not create a possibility for an accident or malfunction that has not been previously analyzed.

III. No. Basis 4.1.4, Control Rod Program Controls, which deals with continuous rod withdrawal and rod patterns which minimize the consequences of a rod drop accident, has been examined. The change as analyzed does not affect the Technical Specification bases.

E: DCP 90-3049 D RIPTI H E:

Installation of a highly accurate ultrasonic flow monitoring system for both of the Unit 2, 96-inch pipes supplying circulating water to the condenser.

R:

I. No. As discussed in FSAR 10.4.5, the Circulation Water System has no safety related function. Section 15.2.5 of the FSAR analysis a loss of condenser vacuum whereby there is a loss of one or more of the Circulating Water Pumps. This change will not increase the probability of a loss of condenser vacuum.

H. No. Section 10.4.1.3.2 of the FSAR analysis condenser leakage. An analysis was conducted to evaluate the effects of failure of the mounting plates on condenser leahge. The loss of the stainless steel pipes, transducers, and/or pull boxes will not increase the probability of an accident.

III. No. This modification affects the circulating water system which has no safety-related function. In particular, flow measurement of circulating water has no direct interface with any safety function in the plant.

SEEM.: 91-109 DCP 88-3047 A/3 DE RIPTI F HAM E:

Replacement of corroded cooling coils which are located in the Unit 1 HPCI pump room coolers.

MARY:

I. No. FSAR Section 9.4.2.2, 9.2.5 and Table 9.4-1, 9.4-4, 9.2-2 and 9.2-3 provide details of RCIC pump room cooling. These sections and tables were reviewed for applicability. The proposed modifications to replace the existing coils neither alter the function nor the performance of the HPCI pump room unit coolers.

II. No. FSAR Section 9,2.5.1 and 9.4.2.2.1 discuss the design basis for HPCI pump room unit coolers. Replacement of the coils will neither alter the function nor the performance of the HPCI pump room coolers and they meet the design basis discussed in these FSAR sections. These modifications will enhance the reliability of the unit coolers.

III. No. Although SSES Technical Specification does not include specific requirements for HPCI pump room coolers, they are required as attendant cooling for the HPCI pumps. Therefore, Technical Specification Sections 3/4.3.3 and 3/4.5.1 relating to HPCI pump operability were reviewed for applicability. These modifications neither reduce the operability of any equipment nor require additional surveillance to ensure plant safety.

SCP E911021 Establish the relay alarm setpoints for existing Unit 1 Class 1E 24V DC battery charger failure alarm relays and battery monitors.

SVMMkkR':

I. No. The 24 VDC System provides a reliable and continuous source of uninterrupted DC power to the process radiation monitors, the source range monitors, and the intermediate range monitors. A failure of these relays to alarm does not cause a decrease in core coolant temperature or any other potential accident listed in Chapter 15 of the FSAR.

II. No. The effects of a battery or battery charger failure have previously been evaluated as part of FSAR Table 6.3-5 Single Failure Evaluation. There are no other failures in the 24 VDC System which can be postulated as a result of this setpoint change package.

III. No. This setpoint change package does not affect the operation of the Class 1E 24 VDC System. Further, the setpoints established by this change package do not cause a failure of the Class 1E 24 VDC System. Therefore, margin in the 24 VDC System is not affected by this setpoint change package.

~ER~N.: 91-111 R REFE E: SCP E911022 DE RIPTI F HAN E:

Establish the setpoints for existing Unit 1 Class 1E 125V DC battery charger failure alarm relays, battery monitors, and ground detection alarm relays.

SQMh/AR Y.'.

No, The 125 VDC System provides reliable and continuous control and indication power to isolation valves, circuit breakers, inverters, annunciators, and other loads supplied from the 125 VDC Class 1E and Non-Class 1E distribution panels.

A failure of these relays to alarm does not cause a decrease in core coolant temperature or any other potential accident listed in Chapter 15 of the FSAR.

II. No. The effects of a battery or battery charger failure have previously been evaluated ~

as part of FSAR Table 6.3-5 Single Failure Evaluation. There are no other failures in the 125 VDC System which can be postulated as a result of this setpoint change package.

III. No. This setpoint change package does not affect the operation of the Class 1E 125 VDC System. Further, the setpoints established by this change package do not cause a failure of the Class 1E 125 VDC System. Therefore, margin in the 125 VDC System is not affected by this setpoint change package.

~B~.: 91-112 R REFERE K E9 11023 DE RlPTI F H K Establish the setpoints for existing Unit 1 Class 1E 250V DC battery charger failure alarm relays, battery monitors, and ground detection alarm relays.

~MAII:

I. No. The 250 VDC System provides reliable and continuous power to safety related pumps, valves, and power supplies. A failure of these relays to alarm does not cause a decrease in core coolant temperature or any other potential accident listed in Chapter 15 of the FSAR.

H. No. The effects of a battery or battery charger failure have previously been evaluated as part of FSAR Table 6.3-5 Single Failure Evaluation. There are no other failures in the 250 VDC System which can be postulated as a result of this setpoint change package.

III. No. This setpoint change package does not affect the operation of the Class 1E 250 VDC System. Further, the setpoints established by this change package do not cause a failure of the Class 1E 250 VDC System. Therefore, margin in the 250 VDC System is not affected by this setpoint change package.

~ER~.: 91-113 R REFEREN E: SCP E911024 DE RlPTI F HAN E:

Establish the relay alarm setpoints for existing Unit 2 Class IE 24V DC battery charger failure alarm relays and battery monitors.

RMMMARI:

I. No. The 24 VDC System provides a reliable and continuous source of uninterrupted DC power to the process radiation monitors, the source range monitors, and the intermediate range monitors. A failure of these relays to alarm does not cause a decrease in core coolant temperature or any other potential accident listed in Chapter 15 of the FSAR.

II. No. The effects of a battery or battery charger failure have previously been evaluated as part of FSAR Table 6.3-5 Single Failure Evaluation. There are no other failures in the 24 VDC System which can be postulated as a result of this setpoint change package.

III.

~ No.~ This setpoint change package does not affect the operation of the Class 1E 24 VDC System. Further, the setpoints established by this change package do not cause a failure of the Class 1E 24 VDC System. Therefore, margin in the 24 VDC System is not affected by this setpoint change package.

R REFBREN E: SCP E911025 DE RIPTI F HA E:

Establish the setpoints for existing Unit 2 Class 1E 125 VDC battery charger failure alarm relays, battery monitors, and ground detection alarm relays.

RQMMARY:

I. No. The 125 VDC System provides reliable and continuous control and indication power to isolation valves, circuit breakers, inverters, annunciators, and other loads supplied from the 125 VDC Class lE and Non-Class 1E distribution panels.

A failure of these relays to alarm does not cause a decrease in core coolant temperature or any other potential accident listed in Chapter 15 of the FSAR.

II. No. The effects of a battery or battery charger failure have previously been evaluated as part of FSAR Table 6.3-5 Single Failure Evaluation. There are no other failures in the 125 VDC System which can be postulated as a result of this setpoint change package.

III. No. This setpoint change package does not affect the operation of the Class 1E 125 VDC System. Further, the setpoints established by this change package do not cause a failure of the Class lE 125 VDC System. Therefore, margin in the 125 VDC System is not affected by this setpoint change package.

~ER~N7.: 91-115 R REFFRE E: SCP E911026 DE RIPTI N F HAN E:

Establish the setpoints for existing Unit 2 Class 1E 250 VDC battery charger failure alarm relays, battery monitors, and ground detection alarm relays.

RMMMAR Y I. No. The 250 VDC System provides reliable and continuous power to safety related pumps, valves, and power supplies. A failure of these relays to alarm does not cause a decrease in core coolant temperature or any other potential accident listed in Chapter 15 of the FSAR.

II. No. The effects of a battery or battery charger failure have previously been evaluated as part of FSAR Table 6.3-5 Single Failure Evaluation. There are no other failures in the 250 VDC System which can be postulated as a result of this setpoint change package.

III. No. This setpoint change package does not affect the operation of the Class 1E 250 VDC System. Further, the setpoints established by this change package do not cause a failure of the Class 1E 250 VDC System. Therefore, margin in the 250 VDC System is not affected by this setpoint change package.

R REFERE: DCP 91-9013, Rev. 0 Restores the fuel oil supply header vent line and reroutes the flowing vent header drain line for diesel generator 'D'o address slow diesel starting times.

I. No. FSAR Chapters 6 and 15 and Sections 8.3.1.4 and 9.5.4 were reviewed for those accidents or equipment malfunctions for which this auxiliary fuel oil system modification could have an impact. The probability of occurrence of an accident has not been increased because the modification does not change the design basis and operation of the fuel oil system but only serves to ensure that normal diesel generator starting times are achieved.

H. No. The only credible accident which this modification could affect is the failure of a diesel generator. A new failure mode of the fuel oil system piping,,leading to a diesel failure, has not been created.

IH. No.

~ ~ In Technical Specification Section 4.8.1.1.2, each diesel generator is required to reach 600 RPM within 10 seconds of a initiation signal. This modification does not require a revision to this starting time requirement, and only enhances the ability of the diesel to meet this requirement.

R RE  : DCP 89-3018C, Rev. 2 DE RlPTI F Eo Remove existing heat trace and then install new heat trace with a single pass on Containment Atmosphere Control (CAC) solenoid valves where required. Improve the heat trace controls and the CAC solenoid valve temperature monitoring.

I. No. The reconfiguration of heat trace on the CAC solenoid valves is an improvement and does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.

H. No. The action does not alter the function of the heat trace. Temperatures on the piping will continue to be maintained above the dew point of the sampled air during normal operation with reduced heat trace cycling and thermal stress to the piping and valves.

III. No. The Reactor Coolant Leakage Detection System identified in Section 3/4.4.3 of the Technical Specification is required to monitor and detect leakage from the reactor coolant pressure boundary. The proposed action does not affect the detection system's ability to perform it's function.

The operability of the primary containment isolation valves identified in Section 3/4.6.3 of the Technical Specification ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment,

SER M-'1-118 R E  : DCP 89-9158A, Rev. 0 DE RT F ANE Installation of a new pressure transmitter to monitor HPCI Stop Valve balance chamber pressure which will subsequently provide for optimum stop valve adjustment and operation.

R:

I. No. Failure analyses for the proposed changes are enveloped by the accident analyses "Instrument Line Break" (FSAR Section 15.6.2) and Miscellaneous Small Releases-Outside Containment (FSAR Section 15.7.2. 1).

II. No. Worst case scenarios for an accident involving an instrument tubing rupture have been conservatively analyzed in the FSAR. The proposed change does not involve a change in system operation or add a more severe type of failure mode.

No new unanalyzed type of malfunction is made probable by this modification.

IH. No. Technical Specifications 3/4.3 Instrumentation and 3/4.5 Emergency Core Cooling Systems provide the bases and limiting conditions of operation for associated plant instrumentation and the HPCI system. All instrument tubing installation willfollow accepted standards and codes with the potential for a HPCI line break scenario being extremely low and offset by the increased probability of correct stop valve operation due to the improved monitoring capability during adjustment.

~ERMQ.: 91-119 R REFERE: NL-91-011 DE RIPTI F HAN E:

Fuel Pool Cleanout Project to remove all irradiated reactor vessel components (excluding spent fuel) from the Unit 1 and Unit 2 spent fuel pools.

KSL445 I. No. Accident analyses contained in the FSAR for a Fuel Handling Accident (FSAR 15.7.4) address accidents far more severe than similar accidents postulated for this project. Section 15.7.5 in the FSAR addresses a Spent Fuel Cask Drop Accident and concludes that this event is not credible due to the use of redundant rigging and a single failure proof crane. The Fuel Pool Cleanout Project activities are therefore bounded by the FSAR analyses noted.

H. No. Two types of accidents are postulated: a loaded cask drop scenario and draining of the cask storage pit with a significant inventory of irradiated hardware becoming uncovered. Cask drop accidents are not considered a credible accident per FSAR Section 15.7.5 due to redundant rigging and a single failure proof crane, preventing a single failure.

Loss of water shielding in the cask storage pit is a credible scenario for evaluation. The airborne radiological consequences of this event would be bounded by the Fuel Handling Accident analyzed in FSAR Section 15.7.4.

III. No. All Fuel Pool Cleanout Project activities will be conducted in accordance with plant Technical Specifications.

$ER lpga.: 91-120

DCP 88-3046 A/B, Rev.0 Replace cooling coils in the Unit 2 RCIC pump room coolers.

~~A~Y:

I. No. FSAR Section 9.4.2.2, 9.2.5 and Table 9.4-1, 9.4-4, 9.2-2, and 9.2-3 provide details of RCIC pump room cooling. These sections and tables were reviewed for applicability. The proposed modifications to replace the existing coils neither alter the function nor the performance of the RCIC pump room unit coolers.

II. No. The performance, function and the heat removal capacity of the replacement coils are the same as that of the original coils. The failure modes associated with the replacement coils are the same as those for the original coils. This change does not modify the plant response to a design basis accident.

III. No. Technical Specification Sections 3/4.7.3, relating to RCIC pump operability were reviewed for applicability. These modifications neither reduce the operability of any equipment nor require additional surveillance to ensure plant safety. These modifications do not change the function or performance of RCIC pump room unit coolers.

~ER P(g.: 91-121 R RE  : DCP 88-3048 A/B, Rev. 0 DE RIPTI F H E:

Replace cooling coils in the Unit 2 HPCI pump room coolers.

+1~MA~Y:

I. No. FSAR Section 9.4.2.2, 9.2.5 and Table 9.4-1, 9.4-4, 9.2-2 and 9.2-3 provide details of HPCI pump room cooling. These sections and tables were reviewed for applicability. The proposed modifications to replace the existing coils neither alter the function nor the performance of the HPCI pump room unit coolers.

II. No. FSAR Section 9.2.5.1 and 9.4.2.2.1 discuss the design basis for HPCI pump room unit coolers. Replacement of the coils will neither alter the function nor the performance of the HPCI pump room coolers and they meet the design basis discussed in these FSAR sections. These modifications will enhance the reliability of the unit coolers.

IH. No. Technical Specification Sections 3/4.3.3, and 3/4.5.1, relating to HPCI pump operability were reviewed for applicability. These modifications neither reduce the operability of any equipment nor require additional surveillance to ensure plant safety. These modifications do not change the function or performance of HPCI pump room unit coolers.

SEXED-'1-122

DCP 89-9153 A & B, Rev. 0 DE I Removal of existing seals and installation of new RHR pump "Hot" seals and seal piping. Also, isolation of existing seal coolers from the RHR and ESW fluid systems. Removal of ESW piping, seal coolers and associated hardware made obsolete by this change.

I. No. The design has been reviewed against Section 5.4.7 (RHR), Section 6.2 (containment cooling) Section 6.3 (ECCS-RHR/LPCI) and Section 9.2.5 (ESW) to confirm that the design changes will not increase the possibility of an accident nor affect the subject systems or any other systems relying on their function/operation.

II. No. Installing the new "HOT" seal and deleting ESW cooling to the new seal will not affect the function of the RHR or ESW systems or any component or system relying on the subject systems. The new seal shall eliminate the RHR/ESW interface thus tending to simplify the operation of the systems and reduce the potential for operational error or radioactive releases to the environment.

Technical Bases 3/4.4.9 and 3/4.9.11 (RHR) 3/4.5.1 and 3/4.5.2 (ECCS-Operating and Shutdown), 3/4.7.1 Service Water Systems and 3/4.6.1 Primary Containment were reviewed. This change will not alter RHR pump functional capability for LPCI operation. Also, the elimination of ESW cooling water to the RHR pump seal cooler reduces the cooling capacity load on the system, thus increasing availability of cooling water for operation of the safety-related equipment.

DCP 90-3050, Rev. 0 Provide fire protection service to the Temporary Laundry Facility and access to this Facility from the Unit 2 Turbine Building through door ¹32.

R:

I. No. FSAR Section 9.5.1 describes the requirements of the fire protection program and the types of fire protection systems in operation at SSES. Based on the flow and pressure capabilities of the fire protection water supply, implementation of this modification will not degrade the existing fire protection as described in the FSAR, Section 9.5-1.

II. No. The proposed modification expands the protective functions and is designed in accordance with original design criteria and codes imposed on the original equipment.

III. No. The Unit 2 Technical Specification and "Bases" have been reviewed, specifically section B3/4.7.6, Fire Suppression Systems. The Turbine Building Sprinkler System is not specifically addressed in the Technical Specifications. Also, Section B3/4.8 - Electrical Power Systems of the Unit 2 Technical Specifications were reviewed. The proposed modification will eliminate a magnetic switch and local alarming horn from door ¹32. The deletion will have no effect on the margin of safety for electrical power systems as described in the technical specifications.

~ER~.: 91-124 R REFER  : DCP 90-3046, Rev. 0 DE RIPTI F HA E:

Replacement of tube bundles used in the Jacket Water Cooler Heat Exchangers for Diesel Generators A-D with tube bundles using AL-6XN material for both the tubes and tubesheets.

QgM ARY:

I. No. The proposed modification does not increase the probability of occurrence of the accidents described in FSAR Sections 6.2, 6.3 and Chapter 15 because the replacement tube bundles will be designed, constructed and tested to standards:

equivalent to that used for the original tube bundles.

II. No. The replacement tube bundles have been designed to be functionally identical to the originals except for the possibility of a greater pressure drop through the tube side of a bundle. Any additional pressure drop will however, be compensated for by an increased opening of the downstream flow balance valve.

III. No. The margin of safety implicitly described in Technical Specifications Sections 3/4.7.1 and 3/4.8.1 is not reduced because the replacement tube bundles will be designed, constructed and tested to an equivalent standard as compared to the original tube bundles.

$ gR~).: 91-125 RE  : SCP 791-1019 Ec 1020, Rev. 0 Eo Decrease the setpoint of Cooling Tower Screen Differential Level Switches LDIS-11506, LDIS-11507, LDISH-21506 and LDISH-21507 to prevent too great a level/pressure differential across the Cooling Tower Screens before actuating warning alarms.

L No. Neither the switches, the screens or the Circulating Water System are important to safety as evaluated in the FSAR Section 10.4.5.1. The setpoint change does not affect the design, function, accuracy, reliability, response time or actuation logic of any other component or system. Implementation of the setpoint change also does not require any physical modifications.

II. No. The effect of the change is to reduce the amount of screen blockage required to p'roduce an alarm. The enhanced sensitivity allows more time for corrective action before the onset of pump cavitation. The decrease in differential level setpoint increases sensitivity to screen blockage which enhances, not degrades,

~

the switches protective function. ~

~ ~ ~

IG. No.

~ ~ The switches are not safety-related and their setpoint is not governed by Technical

~ ~ ~ ~

Specifications. The screens and Circulating Water system also have no safety-related function.

0

DCP 90-3056E, Rev. 0 D I Eo Provide protection for sensor wiring inside the turbine building and reactor building chiller compressor transmission compartments by routing the wiring inside copper tubing inside the transmission compartment of Unit 1.

A I. No. FSAR Section 9.2. 12.3 - Reactor Building Chilled Water System, FSAR Section 9.2.12.2 - Turbine Building Chilled Water System, FSAR Chapter 15.0-Accident Analyses, SER (NUREG-0776) Section 9.2 and the Fire Protection Review Report were reviewed.. The Chapter 15 Accident Analyses do not explicitly address the RBCW or TBCW Systems, or take credit for the operation of the RBCW or TBCW systems in the mitigation of a Design Basis Accident (DBA). The proposed modification does not adversely affect the performance of the RBCW and TBCW Systems as described in the above FSAR sections.

II. No. The RBCW and TBCW systems performance requirements will be maintained with the installation of protective copper tubing on the interior of the chiller compressor transmission compartments. Sensor wiring has been subject to damage and breakage due to the impact of high velocity oil spray inside the chiller compressor transmission compartments. Since breakage of these wires causes the chillers to trip, the installation of the copper tubing to protect the wires will prevent nuisance trips and therefore enhance system reliability.

III. No. Technical Specification B3/4 - Bases for Sections 3.0 and 4.0, and 3/4.3-Instrumentation, Section 3/4.6.1, Containment Systems, Sections 3/4.7.1, Service Water Systems and 3/4.8.4, Electrical Equipment Protective Devices were reviewed for impact of the proposed modification. The proposed modification wiH not affect the operation of any system as described in, the Technical Specifications.

~BR~N.: 91-127 RBFE  : DCP 90-3055E, Rev. 0 Provide protection for sensor wiring inside the turbine building and reactor building chiller compressor transmission compartments by routing the wiring inside copper tubing inside the transmission compartment of Unit 2.

I. No. FSAR Section 9.2. 12.3 - Reactor Building Chilled Water System, FSAR Section 9.2.12.2 - Turbine Building Chilled Water System, FSAR Chapter 15.0-Accident Analyses, SER (NUREG-0776) Section 9.2 and the Fire Protection Review Report were reviewed. The Chapter 15 Accident Analyses do not explicitly address the RBCW or TBCW Systems, or take credit for the operation of the RBCW or TBCW systems in the mitigation of a Design Basis Accident (DBA). The proposed modification does not adversely affect the performance of the RBCW and TBCW Systems as described in the above FSAR sections.

II. No. The RBCW and TBCW systems performance requirements will be maintained with the installation of protective copper tubing on the interior of the chiller compressor transmission compartments. Sensor wiring has been subject to damage and breakage due to the impact of high velocity oil spray inside the chiller compressor transmission compartments. Since breakage of these wires causes the chillers to trip, the installation of the copper tubing to protect the wires wiH prevent nuisance trips and therefore enhance system reliability.

III. No. Technical Specification B3/4 - Bases for Sections 3.0 and 4.0, and 3/4.3-Instrumentation, Section 3/4.6.1, Containment Systems, Sections 3/4.7.1, Service Water Systems and 3/4.8.4, Electrical Equipment Protective Devices were reviewed for impact of the proposed modification. The proposed modification will not affect the operation of any system as described in the Technical Specifications.

fER~N.: 91-128 R E  : DCP 90-3055B, Rev. 0 This modification will install the sensors required for the new Reactor Building Chiller "B" diagnostic system.

I. No. As discussed in FSAR Section 9.2.12.3, the only safety related functions of the RBCW System are performed by the primary containment piping penetrations and containment isolation valves. The proposed action does not affect safety related portions of the RBCW System. The addition of the sensors will only affect portions of the SW and RBCW systems associated with chiller 2K206B.

II. No. The proposed modification will add sensors required for a diagnostic system which do not alter the intended function of the SW system and the RBCW System as described in FSAR Section 9.2.1.1 and 9.2.12.3. The sensors willbe designed and installed in accordance with the original codes of construction listed in FSAR Section 3.

The proposed action does not affect operability of the RBCW System nor introduce any new mode of operation. The present operation of the chiller is not changed and no new failures or malfunctions are created by addition of the sensors.

III. No. RBCW System and the SW System operation does not contribute to the margin of safety as defined in the basis of any Technical Specification. SSES Technical Specifications (Unit 2) Section 3.6.1.7 and Table 3.6,3-1 and their bases were reviewed for applicability. Section 3.6.1.7 contains the requirements for average drywell temperatures during reactor operating conditions 1, 2 and 3. These requirements are not affected by this modification. Table 3.6.3-1 includes the RBCW primary containment isolation valves, which will not be affected by the proposed modification.

~ER~N.: 91-129 R R  : DCP 90-3055A, Rev. 0 DE RTPTI F This modification willremove the existing Reactor Building Chiller "B" vent and drain lines and replace them with drain lines designed to aUeviate clogging problems.

SIIMhfARF.'.

No. As discussed in FSAR Section 9.2. 12.3, the only safety related functions of the RBCW System are performed by the primary containment piping penetrations and containment isolation valves.

The proposed action does not affect safety related portions of the RBCW System.

As stated in FSAR Section 9.2.1, the SW System is designed to operate during normal plant operation and plant shutdown with offsite power available. The SW System will not operate on loss of offsite power concurrent with a LOCA.

The SW System operation has no safety related function, and failure of the system will not compromise any safety related system or component or prevent a safe nuclear shutdown.

The proposed modification adds no new function to the RBCW System as described in FSAR Section 9.2.12.3.

II. No. The proposed modification willreplace existing vent and drain lines which do not alter the intended function of the SW system and the RBCW System as described in FSAR Section 9.2.1.1 and 9.2.12.3.

III. No. RBCW System and the SW System operation does not contribute to the margin of safety as defined in the basis of any Technical Specification. SSES Technical Specifications (Unit 2) Section 3.6.1.7 and Table 3.6-3-1 and their bases were reviewed for applicability. Section 3.6.1.7 contains the requirements for average drywell temperatures during reactor operating conditions 1, 2 and 3. These requirements are not affected by this modification. Table 3.6.3-1 includes the RBCW primary containment isolation valves, which will not be affected by the proposed modification.

~ER~.: 91-130

DCP 90-3025, Rev. 0 Provides for the installation of the Unit 2 Turbine and Reactor Building sample stations and the connection of the process sample streams and interface systems, including water chiller units to the sample stations.

$ Q~ARF I. No. The review of FSAR Sections 6.2.3.2.3.1, 6.2.4.3.3.5 and 9.2.23 indicate that the RBCCW System containment penetrations and isolation valves perform containment isolation functions which are important to safety. The installation of new piping does not affect these functions.

A review of the FSAR Section 9.4.2.1.3 indicates that Reactor Building Zone I filtered exhaust and Zone I unfiltered exhaust interface with the recirculation system via secondary containment isolation dampers. These dampers which are important to safety for the Reactor Building, and the recirculation system and the Standby Gas Treatment System are not affected by this modification.

II.

~ No.~ The process sample system described in FSAR section 9.3.2.3 does not perform any safety functions. Sample lines made operational by this change are only for chemistry monitoring and provide no direct control functions. Malfunction of the sample stations or the modulating damper is not a malfunction of a different type than previously evaluated because it does not have an adverse affect on the safety function of the reactor building recirculation system operation.

IH. No. This modification does not reduce the margin of safety since the interfaces being installed for the TBSS and RBSS are not safety-related and do not affect any margins of safety described in the Technical Specifications. The acceptance limits for the RCS chemistry as noted in the Technical Specification Section 3/4.4.4 will not be affected as a result of this modification.

DCP 89-9151, Rev. 0 Location of two "A" and two "B" loop temperature elements one each into the north side and south side return air ducts of the Control Structure HVAC system prior to the mixing of return air, outside air, or the CREOASS air fiows.

SMALL':

I. No. FSAR sections 6.4, 7.3 and 9.4 provide descriptions of the HVAC systems. The references were reviewed, and it was deterinined this modification maintains the function or operation of the existing Control Structure HVAC System Electrical separation, equipment qualification, 10CFR Appendix R requirements, seismic mounting and other design considerations are incorporated in the design output documents.

II. No. The new temperature monitoring circuits'ailure mode is the same. They can fail high or low. If the temperature elements and temperature transmitter fail, this condition is now alarmed. This is an enhancement over the original design. If the temperature selector were to fail, it can only fail high or low, which is exactly the same failure mode as the existing system.

The relocation of the temperature elements and high temperature selector does not introduce a different failure mode, but improves on the old system by providing an alarm on temperature element failure.

III. No. Requirements for the Control Structure HVAC system are not specifically addressed by the Technical Specifications, except that Section 3/4.7.2 of the Technical Specifications, Control Room Emergency Outside Air Supply System (CREOASS), requires two independent and operable subsystems of emergency outside supply air for the Control Room. The Control Structure HVAC system is included by inference in Section 3/4.7.2 in that it has to be operational when CREOASS is operational. This modification does not affect the ability of the instrument loops to meet the requirements of the Control Structure HVAC system, but enhances operability by providing a more representative return air temperature for control.

/ERIN.": 91-132

DCP 88-3049 A & B, Rev. 1 PTI Replacement of cooling coils 1E230 A/B/C/D which are located in the RHR pump room coolers 1V210 A/B/C/D with new coils due to corrosion and localized pitting.

R:

I. No. None of the design basis accidents discussed in FSAR Section 15 are affected by these modifications. Further, the function and performance of the cooling coils is not altered by these modifications. Failure modes associated with the replacement coils are same as the original coils and the failure modes of the original coils do not become an initiating event for any design basis accident.

II. No. FSAR Section 9.2.5.1 and 9.4.2.2.1 discuss the design basis for RHR pump room unit coolers. Replacement of the coils will neither alter the function nor the performance of the RHR pump room coolers and they meet the design basis

. discussed in these FSAR sections. These modifications will enhance the reliability of the unit coolers.

III.

~ No.~ Although SSES Technical Specification does not include specific requirements for RHR pump room coolers, they are required as attendant cooling for the RHR pumps. Therefore, Technical Specification Sections 3/4.4.9, 3/4.5.1, 3/4.5.2, 3/4.6.2.2, 3/4.6.2.3, 3/4.9.11.1 and 3/4.9.11.2, relating to RHR pump operability were reviewed for applicability. These modifications neither reduce the operability of any equipment nor require additional surveillance to ensure plant safety. These modifications do not change the function or performance of RHR pump room unit coolers.

/BRAN.: 91-133 R REFE E: DCP 90-9007M, Rev. 0 DE RlPTI F HAN E:

Circuit modifications and installation of qualified isolation devices to prevent interaction between the class 1E and non-class 1E portion of circuits in Main Steam drain, RWCU, RHR and Main Steam Rad monitors.

~il ARY:

I. No. The proposed action protects the safety functions of class 1E components of the various systems under consideration. It does not change the operation of the instrumentation loops or the plant operator's interface with these loops or any other safety related equipment. Hence, the proposed action does not increase the probability of occurrence or the consequences of an accident or malfunction of safety equipment, as previously evaluated in the FSAR.

H. No. This modification provides acceptable isolation between Class 1E and non-Class 1E interface wiring as described in FSAR Section 8.1.6.lq and does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.

III. No. This change has no direct impact on the operation of the containment isolation valves described in Section 3/4.6.3 and Table 3.6.3-1 of the Technical Specifications. The isolation actuation instrumentation requirements for RWCU differential flow and main steam line radiation monitoring as indicated in Table 3.3.2-1 of the Technical Specification are not affected by the proposed modification. Similarly, it has no impact on the reactor protection system instrumentation or response times for main steam line radiation as required by Tables 3.3.1-1 and 3.3.1-2 of the Technical Specification.

~ER~.: 91-134

SCP-J-90-1040 Change from 85'F to 123'F for temperature switches TSH-27702A2 and TSH-27702B2 associated with drywell unit coolers 2V411A and 2V411B to eliminate numerous false alarms.

I. No. This change does not affect any of the accidents discussed in FSAR Section 15.

Further, this setpoint change does not alter the function and performance of the drywell unit coolers 2V411A & 2V411B. This change does not alter the failure modes associated with either TSH-27002A2 and TSH-27002B2 or the drywell unit coolers 2V411A and 2V411B. Further, FSAR Sections 9.45, 7.3.1.1b.8.5.6 and tables 9.4-8, 9.4-9, 9.4-10 provide details of Drywell Cooling System. These sections and tables were reviewed for applicability. The proposed change has no impact on these sections and tables.

II. No. FSAR Section 9.4.5.1 discusses the design basis for drywell cooling system including unit coolers 2V411A&B. This change does not alter the function and the performance of any of the above drywell unit coolers and it does not alter the design basis discussed in the FSAR. This change will eliminate false alarms.

III. No. SSES Technical Specification sections 3/4.6.1.7 was reviewed for applicability.

This change neither reduces the operability of any equipment nor requires additional surveillance to ensure plant safety. This change does not change the function or performance of drywell unit coolers 2V411A&B.

~ER~N.: 91-135 R REFERE: DCP 90-9074, Rev. 1 DE RI F H E:

Removal of existing Topaz Static Inverters B21B-K801A and B21B-K801B in panels 2C614C and 2C614D and replacement with equivalent Topaz Inverters to preclude imminent failure due to capacitor aging.

RRMMARY:

I. No. The RCIC inverters and the temperature switches and isolation valves are redundant. Failure of a single inverter or switch to detect a leak does not preclude detection. The other methods of leak detection provide backup for the temperature instruments. (Ref. FSAR 7.6.1a.4.3.3) Additionally, the replacement inverters are functionally equivalent in all respects of standards and specification and performance as the old inverters.

H. No. The proposed change does not affect any system other than leak detection and batteries 2D630 and 2D640. This change does not affect the ability to detect and isolate leakage. Additionally, the replacement inverters represent an upgraded version of equipment equivalent in performance standards.

The proposed change satisfies the bases for Section 3/4.3.2 and 3/4.8 by ensuring that the replacement inverters perform the same function and meet all design requirements for providing power to the RCIC Steam Leak Detection System from the 125VDC batteries 2D630 and 2D640 and does not reduce any margin of safety defined for any technical specification.

te

~ER~N.: 91-136 R RE  : DCP 90-9051, Rev. 0 DE RIPTI H E:

This modification will install gravity drains in the following manholes: MH2510, MH2511, MH2512, MH2513, MH2031 and MH2032 to decrease the frequency of weather induced flooding.

R:

I. No. The FSAR has been reviewed, specifically Sections 9.2.5 - ESW System and 9.2.6 - RHRSW System. The addition of gravity drains in the instrumentation pits and manholes willonly remove unwanted ground water from these areas thus increase the reliability of instrumentation and circuits involved with these systems. The FSAR does not specifically address Instrumentation Pits and/or Conduit Manholes. The FSAR, Section 3.2.1, does state, however, that plant structures, systems, and components including their foundations and supports designed to remain functional in the event of a Safe Shutdown Earthquake, are designated as Seismic Category I. Since these manholes house safety related equipment, these structures are analyzed for seismic conditions. Adding drains in these manholes will not degrade or have any structural affect on the manholes ability to resist an earthquake or any other type of failure the structures were designed to withstand.

H. No. The addition of drains within the six conduit and instrumentation manholes south of the ESSW Pumphouse does not change the function and/or operation of any plant component or system. This modification does not affect the safety function of any equipment, therefore, no new unanalyzed type of malfunction important to safety is created.

IH. No. Technical Specifications and bases have been reviewed, specifically Sections 3.3.7.4- Remote Shutdown Monitoring Instrumentation, 3.7.1.1 RHRSW System and 3.7.1.2 - ESW System. The drains will not alter any system or system function as it relates to supporting safety related components as required by the Technical Specifications.

DCP 90-9034B, Rev. 0 This modification will eliminate alarm PAL03341, Generator Hydrogen Supply Bulk Storage Low Pressure, and pressure switch PSL03341 the originating source of the alarm signal.

KSBILY:

I. No. FSAR Section 3.2 - Classification of Structures, Components and Systems, Section 10.2.2.2 - Generator and Exciter, and Chapter 15 - Accident Analysis were reviewed to determine the impact of this modification. Chapter 15 of FSAR does not explicitly address the bulk hydrogen storage system or take credit for the bulk hydrogen storage system in the mitigation of a Design Basis Accident (DBA).

II. No. The proposed action does not change the basic function of the existing hydrogen storage system, it only deletes a nuisance alarm. The design and implementation of this modification will be consistent with PP&L's existing standards and technical engineering guidelines.

III.

~ No.~ The Technical Specifications, especially sections: 3/4.7 - Plant Systems, 3/4.8-Electrical Power Systems, and B3/4 - Bases for sections 3 and 4 were reviewed.

No margin of safety or limiting conditions related to the bulk hydrogen storage system were found.

~ER~N.: 91-138 R REFERE E: DCP 90-3064F, Rev. 0 DE RI In the Unit 1 Reactor Building Chiller 1K206B, the existing air-operated butterfly valves will be replaced with new control valves having an equal percentage control characteristics, high turndown ratio, and increased anticavitation protection.

R:

I. No. As discussed in FSAR Section 9.2.12.3, the only safety related functions of the RBCW System are performed by the primary containment piping penetrations and the containment isolation valves. The proposed action does not affect safety related portions of the RBCW System. The installation of the new temperature control valves and their temperature controller will affect only a portion of the SW System associated with the chiller condenser water circulating pump.

II. No. The proposed modification will install new temperature control valves and their temperature controller which do not alter the intended function of the SW system and the RBCW System as described in FSAR Section 9.2.1.1 and 9.2.12.3. The existing method of the service water temperature control by recirculation of the Service Water as described in FSAR Section 9.2.1.2 will be unchanged.

III. No. The RBCW System and the SW System operability does not contribute to any margin of safety as defined in the basis of any Technical Specification. SSES Technical specifications (Unit 1) Section 3.6.1.7 and Table 3.6.3-1 and their bases were reviewed for applicability. Section 3.6.1.7 contains the requirements for average drywell temperatures during reactor operating conditions 1, 2, and 3.

Table 3.6.3-1 includes the RBCW primary containment isolation valves, which will not be affected by the proposed modification.

R REFE E: DCP 90-3060, Rev. 0 DE RIPTI N E:

In the Unit 2 Reactor Building Chiller 2K205A, the existing air-operated butterfly valves will be replaced with new control valves having an equal percentage control characteristics, high turndown ratio, and increased anticavitation protection.

SUMMARY

I. No. As discussed in FSAR Section 9.2.12.3, the only safety related functions of the RBCW System are performed by the primary containment piping penetrations and the containment isolation valves. The proposed action does not affect safety related portions of the RBCW System. The installation of the new temperature control valves and their temperature controller will affect only a portion of the SW System associated with the chiller condenser water circulating pump.

II. No. The proposed modification will install new temperature control valves and their temperature controller which do not alter the intended function of the SW system and the RBCW System as described'in FSAR Section 9.2.1.1 and 9.2.12.3. The existing method of the service water temperature control by recirculation of the Service Water as described in FSAR Section 9.2.1.2 will be unchanged.

HI. No. The RBCW system and the SW system operability does not contribute to any margin of safety as defined in the basis of any Technical Specification. SSES Technical Specifications (Unit 2) Section 3.6.1.7 and Table 3.6.3-1 and their bases were reviewed for applicability. Section 3.6.1.7 contains the requirements for average drywell temperatures during reactor operating conditions 1, 2, and 3.

Table 3.6.3-1 includes the RBCW primary containment isolation valves, which will not be affected by the proposed modification.

~ER~.: 91-140 R  : DCP 90-3058F, Rev. 0 In the Unit 1 Reactor Building Chiller 1K206A, the existing air-operated butterfly valves will be replaced with new control valves having an equal percentage control characteristics, high turndown ratio, and increased anticavitation protection.

KMILY:

I. No. As discussed in FSAR Section 9.2. 12.3, the only safety related functions of the RBCW System are performed by the primary containment piping penetrations and the containment isolation valves. The proposed action does not affect safety related portions of the RBCW System. The installation of the new temperature control valves and their temperature controller will affect only a portion of the SW System associated with the chiller condenser water circulating pump.

H. No. The proposed modification will install new temperature control valves and their temperature controller which do not alter the intended function of the SW system and the RBCW System as described in FSAR Section 9.2.1.1 and 9.2.12.3. The existing method of the service water temperature control by recirculation of the Service Water as described in FSAR Section 9.2.1.2 will be unchanged.

III. No. The RBCW system and the SW system operability does not contribute to any margin of safety as defined in the basis of any Technical Specification. SSES Technical specifications (Unit 1) Section 3.6.1.7 and Table 3.6.3-1 and their bases were reviewed for applicability. Section 3.6.1.7 contains the requirements for average drywell temperatures during reactor operating conditions 1, 2, and 3.

Table 3.6.3-1 includes the RBCW primary containment isolation valves, which will not be affected by the proposed modification.

DCP 89-9173, Rev. 0 Removal of the existing automatic Air Relief Valves from the vent lines on the service water side of the Main Generator Hydrogen Coolers and Aterex Air Coolers in order that venting of the coolers can be done manually and positively verified by visual inspection for Unit 1.

/gal(~RF:

I. No. The Generator, Aterex, Generator Hydrogen Coolers and Aterex Air Coolers are equipment required for power generation only per FSAR Sections 10.1 and 10.2.

There is no safety function associated with this equipment, neither is any equipment important to safety as previously evaluated in the FSAR connected to or located near this equipment.

II. No. The vent piping willbe installed and supported per all original construction codes and standards. Any potential failure mode of the proposed action such. as pipe failure, valve leakage or operator error will not impose any malfunction of a different type than that of the existing piping associated with the subject coolers, and will not have any effect to the plant safety function.

The Unit I Technical Specifications have been reviewed. The Generator Hydrogen Coolers and the Aterex Air Coolers do not form any design bases or limiting operating conditions as specified in the Technical Specification. The proposed action does not alter the current design bases of the Generator Hydrogen Coolers and the Aterex Air Coolers.

~ER~.: 91-142

DCP 88-3051 A & B, Rev. 1 Replacement of cooling coils 1E231 A/B/C/D which are located in the core spray (C.S.) pump room coolers 1V211 A/B/C/D because of corrosion and localized pitting in Unit 1.

$Qg~gY:

I. No. These modifications do not affect the accidents discussed in FSAR section 15.

Further, the function and performance of the cooling coils is not altered by these modifications. Failure modes associated with the replacement coils are the same as the original coils and the failure modes of the original coils do not become an initiating event for any design basis accident.

II. No. FSAR Sections 9.2.5.1 and 9.4.2.2.1 discuss the design basis for C.S. pump room unit coolers. Replacement of the coils will neither alter the function nor the performance of the C.S. pump room coolers and they meet the design basis discussed in these FSAR sections. These modifications will enhance the reliability of the unit coolers.

III.

~ No.~ Although SSES Technical Specification does not include specific requirements for Core Spray pump room coolers, they are required as attendant cooling for the Core Spray pumps. Therefore, Technical Specification Sections 3/4.5.1 and 3/4.5.2 relating to C.S. pump operability were reviewed for applicability. These modifications neither reduce the operability of any equipment nor require additional surveillance to ensure plant safety. These modifications do not change the function or performance of Core Spray pump room unit coolers.

~ER~N.: 91-143 E: ES-158/258-001 DE RlP77 N E:

This procedure directs removal of eight (8) fuses from the RPS electrical circuit leading to SCRAM pilot valve solenoids in order to de-energize the solenoids and cause a reactor SCRAM.

+I~M'ARY:

It I. No. Intent of the proposed procedure is to provide an alternate method of opening scram valves when RPS does not function per design. Removing fuses in electrical circuits powering SCRAM pilot valve solenoids results in de-energizing SCRAM pilot valve solenoids similar to design which de-energizes the circuits.

Design basis accident analyses in the SAR are not affected since the proposed procedure is only implemented per emergency operating procedures when a transient has progressed well beyond single failure criteria established in FSAR 15.0.3.2.1.1.

II. No. The plant modification created by the proposed procedure does not create a configuration which could result in a new type of accident or increase the probability of an accident to the point where it should be considered within the design basis. Intent of proposed procedure is to open scram valves when RPS does not function per design when required to initiate a reactor SCRAM.

Removing fuses in electrical circuits powering SCRAM pilot valve solenoids results in de-energizing SCRAM pilot valve solenoids similar to design which de-energizes the circuits.

III. No. Technical Specification 3/4.3.1 provides the limiting conditions for operation necessary to preserve the ability of RPS to perform its intended function even during periods when instrument channels may be out of service for maintenance.

The purpose of ES-158/258-001 is the same as the bases described in Technical Specifications bases. Pulling RPS fuses to initiate a reactor SCRAM as proposed in ES-158/258-001 is only performed after RPS fails to automatically perform its design function, that is, a transient has progressed well beyond single failure criteria established in FSAR 15.0.3.2.1.1.

/ERIN.: 91-144 R REF RE E: DCP91-9013 A, B & C, Rev. 0 DE RIPTI F HAN E:

These modifications restore the fuel oil supply header vent line and reroutes the flowing vent header drain line for Diesel Generators 'A', 'B'nd 'C'.

ZllMMARY:

I. No. FSAR Chapters 6 and 15 and Sections 8.3.1.4 and 9.5.4 were reviewed for those accidents or equipment malfunctions for which these auxiliary fuel oil system modifications could have an impact. The probability of occurrence of an accident has not been increased because each modification does not change the design basis and operation of the fuel oil system but only serves to ensure that normal diesel generator starting times are achieved.

H. No. The only credible accident which the modifications could affect is the failure of a diesel generator. A new failure mode of the fuel oil system piping, leading to a diesel failure, has not been created since this change restores a tubing line that was originally supplied by the diesel generator manufacturer.

III.

~ No.~ The only applicable margin of safety affected by this modification is contained in Technical Specification Section 3/4.8 and concerns the starting time operability requirements for the diesel generators. In Technical Specification Section 4.8.1.1.2, each diesel generator is required to reach 600 RPM within 10 seconds of a initiation signal. These modifications do not require a revision to this starting time requirement, and only enhances the ability of the diesel to meet this requirement.

~ER~N.: 91-145 R REFEREN E: DCP 90-3108 A thru D, G & H, Rev. 0 DE RlPTI N F AN E:

Modifications to existing pipe support/restraint configurations that are required in order to reduce the total number of mechanical snubbers on SSES piping systems for Diesel Generator A, B, C & D piping, CREOASS and ESW common valve vault area piping.

~~i~RY:

I. No. Applicable code design limits have been met to ensure piping integrity and system function; FSAR Section 15.0.3.1. The accident events/causes described in FSAR Section 15.1 through 15.8 do not involve systems or components included in this action.

II. No. The action only reduces the number of seismic and hydrodynamic restraints (snubbers) on selected piping systems by reanalyzing each line using optimum restraint configurations. The original allowable equipment interface loads are still being met. Removal of snubbers from thermally active lines will actually minimize the challenges to the systems by eliminating potential snubber failure locations; e.g. a snubber lock-up can create large reaction loads and/or damaging pipe stresses.

III. No. Technical Specifications 3/4.8.1, 3/4.3.7, 3/4.7.2, 3/4.7.1 and 3/4.7.4 were reviewed. In each case, the review margin of safety is not reduced and the capability of the system to perform its intended safety function is not affected.

SCP-J91-1034 Setpoint change for SGTS charcoal bed inlet and outlet pre-ignition temperature alarm instrumentation to avoid iodine desorption and to prevent spurious actuations during normal system operation.

I. No. The SGTS system design as described in the FSAR will not be affected by this proposed change. The ability for SGTS to respond to design basis events as assumed in the FSAR transient and accident analysis will not be affected. This change does not affect any of the instrumentation required for SGTS initiation under postulated accident conditions. The ability for SGTS to purge the containment after postulated transients which result in SRV discharge will not be affected. FSAR Sections 3.13, 6.2.3, 6.5, 7.3.1.1b.4, 9.4.2, 11.5.2 and Ch. 15 were reviewed.

II.

~ No.

~ The basis for this change is to assure proper operation of the SGTS cooling mode by establishing appropriate setpoints. This change is a change to setpoints only and does not involve any new equipment or new failure modes.

III. No. The Unit 1 and Unit 2 Technical Specifications have been reviewed, specifically Sections 3/4.6.5.1, 3/4.6.5.2, and 3/4.6.5.3. The proposed setpoint change as described above does not reduce the margin of safety as set forth in the Technical Specification Bases. This change will enhance SGTS cooling mode operability as required by Technical Specification 4.6.5.3.d.3.

/BR'.: 91-147 R REF8  : NL-91-024, Rev. 0 Increase the setpoint for the RWCU Steam Leak Detection System - Penetration Room High Temperature Isolation Setpoint. This action is required in order to provide sufficient margin above the expected room temperature to preclude inadvertent isolation of the RWCU System.

~il ARY:

I. No. The design basis established for the new temperature setpoints is timely detection of a 25 gpm primary coolant leak from the RWCU system. Existing accidents which result in coolant leakage outside containment are analyzed in FSAR 15.6.2 and 15.6.4. These analyses bound the plants'bility to respond to a 25 gpm RWCU steam leak.

II. No. Implementing this setpoint change does not create any new accidents or malfunctions of a different type. It implements the Steam Leak Detection design basis that was believed to have already existed. That is, to provide timely detection and isolation for small primary coolant leaks while allowing sufficient margin above the expected room temperatures to prevent inadvertent isolations.

III.

~ No.~ Technical Specification 3.3.2 basis section does not address the setpoint basis with regards to leakage rate. The bases does however state "the setpoints ... are established at a level away from the normal operating range to prevent inadvertent actuation of the system involved." Temperature measurement is not discussed in the basis for Tech Spec Section 3.4.3 "Reactor Coolant System Leakage". The proposed setpoint change satisfies the basis stated in Section 3.3.2 by allowing sufficient margin above the maximum operating temperatures while still performing its intended safety function.

SCP J-91-1018, Rev. 0 E:

Increase the setpoint of Condensate Pump, Room Temperature Switch TSH-27442 to reduce the potential for false actuations.

r

~g4RY:

I. No. Neither the switch, the Condensate Pump Room Cooling System (CPRCS) or the Condensate System are important to safety as evaluated in FSAR Sections 9.4.4.2, 9.4.4.3, 10.4.7.1. Furthermore, the equipment in the Condensate Pump Room is designed to function within the operating limits specified in the FSAR.

Although the setpoint is higher than before, it is low enough to insure that the standby cooling unit and trouble alarms are actuated before the temperature exceeds the FSAR limit.

II. No. The setpoint change does not affect the design or function of the switch or any other component or system. Nor, does it cause any component or system to operate outside the performance boundaries described in the FSAR or Technical

~

Specifications. ~

III.~ No.~

~ ~

Neither the switch, the CPRCS or the Condensate System are safety-related.

Furthermore, the Condensate Pump Room Temperature High Setpoint is not governed by Technical Specifications.

~ER~N.: 91-149 R REFE E. NL-91-023, Rev. 0 DE RIPTI F HAN E:

Inspection of irradiated fuel rods to determine the condition of the fuel assemblies following irradiation.

MMARY:

I. No. FSAR Section 9.1.4.2.7.1, 9.1.4.3.7, 9.1.4.5.1, 9.1.5 and 15.7.4 were reviewed. The fuel inspections are not covered in the FSAR. The review evaluated handling fuel, dropping equipment in the spent fuel pool, and interaction with plant auxiliary systems which are similar to events previously evaluated in the FSAR.

II. No. Potential new events evaluated were, (1) radiation streaming on the refueling floor, (2) damage to the refueling floor curb, (3) damage to a fuel rod during inspection, (4) hot particles and, (5) a dropped fuel rod. The use of the inspection equipment or techniques does not create any hazard other than those previously evaluated.

IH. No.

~ ~ Technical Specifications 3/4.9.6, 3/4.9.7 and 3/4.9.9 were reviewed. The proposed fuel inspections will not reduce any margin of safety because (1) the equipment is designed to perform fuel inspections, (2) a minimum of 7 feet of water shielding will be maintained above irradiated components, (3) the radiological consequences of fuel damage that could occur during the inspection are less than those previously evaluated in the FSAR, and 4) all fuel handling operations will be performed in accordance with RE-081-102, 'Fuel Movements Within Vault and Pool Locations, which requires all Technical Specifications related to fuel handling to be complied with prior to and during all operations related to moving irradiated fuel in the spent fuel pool.

NL-91-012, Rev. 1 Hot tap of Unit 1 TBCCW and RBCCW headers to support Reactor Building and turbine Building sample station replacement.

BlggLAAZ:

I. No. After reviewing FSAR Sections 15.1 thru 15.8, it was determined that the hot tapping of RBCCW and TBCCW does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the SAR.

The review of FSAR Section 9.2.3 indicates that the TBCCW system performs no safety-related function. The review of FSAR Section 9.2.2 indicates that the RBCCW System containment penetrations and isolation valves perform containment isolation functions which are safety related. The operability of the containment isolation valves will not be jeopardized due to the location of the hot ta P s.

II.

~ No.~ In accordance with FSAR Sections 9.2.2 and 9.2.3, the RBCCW and TBCCW lines are not required to function after an accident.

III. No. Technical Specification Sections 3/4.6.1 and 3/4.6.3 contain requirements for the RBCCW Primary Containment Isolation Valve leakage integrity and operability.

The RBCCW hot taps which are being performed do not make any direct changes to these valves.

The Technical Specifications do not address the TBCCW system since it has no function which is important to safety. Hot tapping of this system will not reduce the plant margin of safety as defined in the basis for any Technical Specifications.

R REFERE E, NL-91-011, Rev. 1 Removal of all irradiated reactor vessel components (excluding spent fuel) from the Unit 1 and Unit 2 spent fuel pools.

~V~ARY:

I. No. Accident analyses contained in the FSAR for a Fuel Handling Accident (FSAR 15.7.4) address accidents far more severe than similar accidents postulated for this project. Section 15.7.5 in the FSAR addresses a Spent Fuel Cask Drop Accident and concludes that this event is not credible due to the use of redundant rigging and a single failure proof crane. The Fuel Pool Cleanout Project activities are therefore bounded by the FSAR analyses noted.

II. No. Irradiated hardware will be loaded and transported in casks similar to spent fuel casks. Cask drop accidents are not considered a credible accident per FSAR Section 15.7.5 due to redundant rigging and a single failure proof crane, preventing a single failure.

Loss of water shielding in the cask storage pit is a credible scenario for evaluation. The airborne radiological consequences of this event would be bounded by the Fuel Handling Accident analyzed in FSAR Section 15.7.4.

III. No. All Fuel Pool Cleanout Project activities will be conducted in accordance with plant Technical Specifications. The following Technical Specifications were reviewed: 3/4.3.2, 3/4.3.7, 3/4.9.6, 3/4.9.7, 3/4.9.9, 3/4.11.3, 3/4.11.4, 5.6.2, 6.9.1.8 and 6.13.

DCP 90-9103, Rev. 0 Split the different functions of the Circulating Water Pumps (1P501 A,B,C,D) Aux Control Circuit (scheme 1C1026) to assure receiving a local alarm on panel 1C511 in the Circ. Water Pumphouse and to assure control power to scheme 1C1026.

I. No. The DC Power Systems design basis and function, as previously evaluated in FSAR section 8.3.2 and MJREG 0776 Section 8.3.2 is not altered by this modification.

II. No. The proposed modification does not alter the function or the design basis of the Circulating Water System or its annunciation system. Circulating Water System reliability is enhanced by separately fusing the switchgear breaker aux relays from the start and stop interlock relays on scheme 1C1026. The non-Class 1E 125V DC power supply is isolated from the Class 1E power supply in accordance with FSAR Section 8.1.6.lq.5(iii). This design has no different failure modes or effects than those previously analyzed for this system.

III. No. The Circulating Water System is a system that does not have a safety function and its operability does not contribute to the margin of safety as defined in the basis of the Technical Specifications.

The relay and the indicating lights required for this change require DC power supply from the 125V DC subsystems whose operability is governed by Technical Specification Section 3/4.8.2. The addition of a DC load of approx. 260 mA does not degrade the performance of the 125V DC subsystems.

'CP 90-3060C, Rev. 0 Replace the non-class lE control switch HS-28723 A&B which operates the evaporator water circulating pump 2P217 A&Band the condenser water circulating pump 2P235 A&Bin Reactor Building Chillers to prevent circuit breaker trip instantaneously on high inrush current.

I. No. The replacement control switch is specified to the same criteria as the existing control switch. The function of the Reactor Building Chilled Water System is not modified by this action. However, the switching logic will be improved, eliminating unnecessary tripping of pumps 2P235A (2P235B) and 2P217A (2P2178) when control switch HS-28723A (HS-28723B) is moved from "Auto" to "Start" position.

II. No. These modifications add no new function to the Reactor Building Chilled Water System as described in FSAR Sections 9.2.2 and 9.2. 12.3. These modifications change switching logic to prevent unnecessary tripping of the evaporator water pump and the condenser water pump, which improves chiller reliability.

III. No. The operation of the Reactor Building Chilled Water System has no safety related function. Although isolated portions of chilled water piping of the containment are safety related, these portions are not affected by this modification.

These modifications affect control switch HS-28723A (HS-28723B) which is part of the non-safety portion of the Reactor Building Chilled Water System, which is not addressed in the Technical Specification.

/ERIN.: 91-154

DCP 90-3058C and 90-3064C, Rev. 0 DE Rl F H E:

Replace the non-class lE control switch HS-18723 A&B which operates the evaporator water circulating pump 1P217 A&Band the condenser water circulating pump 1P235 A&Bin Reactor Building Chillers to prevent circuit breaker trip instantaneously on high inrush current.

SllKAfARY:

I. No. The replacement control switch is specified to the same criteria as the existing control switch. The function of the Reactor Building Chilled Water System is not modified by this action. However, the switching logic will be improved, eliminating unnecessary tripping of pumps 1P235A (1P235B) and 1P217A (1P217B) when control switch HS-18723A (HS-18723B) is moved from "Auto" to "Start" position.

H. No. These modifications add no new function to the Reactor Building Chilled Water System as described in FSAR Sections 9.2.2 and 9.2. 12.3. These modifications change switching logic to prevent unnecessary tripping of the evaporator water pump and the condenser water pump, which improves chiller reliability.

IH. No. The operation of the Reactor Building Chilled Water System has no safety related function. Although isolated portions of chilled water piping of the containment are safety related, these portions are not affected by this modification. These modifications affect control switch HS-18723A (HS-18723B) which is part of the non-safety portion of the Reactor Building Chilled Water System, which is not addressed in the Technical Specification.

DCP 90-3093A, Rev. 0 Install a second isolation valve on Elevation 656'f the Turbine Building between the recycle isolation valves HV-11615A thru G, and the sluice/recycleldrain header of the Condensate Demineralizer System to provide a positive means of isolating the demineralizer vessels when the existing butterfly valves fail to provide adequate isolation.

I. No. FSAR Chapter 15 (Accident Analysis) and FSAR Sections 9.2.10 (Condensate Transfer and Storage) and 10.4.6 (Condensate Cleanup System) have been reviewed. The accident analysis contained in Chapter 15 are unaffected by the proposed modification. Section 9.2.10 does not specifically discuss the Condensate Demineralizer System. Section 10.4.6.1 states that the design basis of the Condensate Cleanup System is to maintain condensate purity by removing various contaminants. This section also states the not-to-exceed effluent quality for the demineralizers. The design basis for the system is not affected by the proposed action.

II.

~ No.~ There is no equipment important to safety and no interface with safety equipment or systems associated with this change.

III. No. This modification does not reduce any margin of safety defined in the basis of the Technical Specifications. This modification is non-safety related and does not affect systems having Technical Specification requirements. This change adds valves which are designed and built to the same standards and design conditions as the condensate demineralizer system. The new valves will not in any way degrade the operation of the condensate demineralizer system.

ggR Pgl.: 91-156

DCP 90-3065, Rev. 0 Addition of a Data Acquisition System (DAS) for each of the Unit 2 Reactor Building Chillers 2K206A and 2K206B to monitor the chiller and its support systems on a continuous basis.

I. No. The proposed action does not affect safety related portions of the RBCW System.

The addition of the Data Acquisition System will only affect portions of the SW and RBCW Systems associated with chillers 2K206A and 2K206B. Both a single failure and double failure of the reactor building chillers is analyzed in FSAR Section 9.2.12.3.2.3. Thus the failure modes of this installation are covered by those accidents previously evaluated in the FSAR. The proposed DAS adds no new function to the RBCW system as described in FSAR Section 9.2.12.3.

II. No. The proposed modification will install Data Acquisition Systems and associated wiring to instruments which will not alter the intended function of the SW system and the RBCW System as described in FSAR Section 9.2.1.1 and 9.2.12.3.

The proposed action does not affect operability of the RBCW System nor introduce any new mode of operation. The present operation of the chiller is not changed. The failure modes of this installation are covered by those accidents previously evaluated in the FSAR.

III. No. The RBCW System and the SW System operation does not contribute to the margin of safety as defined in the basis of any Technical Specification. SSES Technical Specifications (Unit 2) Section 3.6.1.7 and Table 3.6.3-1 and their bases were reviewed for applicability. Section 3.6.1.7 contains the requirements for average drywell temperatures during reactor operating conditions 1, 2 and 3; These requirements are not affected by this modification. Table 3.6.3-1 includes the RBCW primary containment isolation valves, which will not be affected by the proposed modification.

The proposed modification affects the non-safety related portions of the RBCW and SW Systems, which are not addressed in the Technical Specifications.

~jER~N.: 91-157 R RFE  : DCP 90-3064D, Rev. 0 DE TPTI F Existing root valves (1" globe valves) will be replaced with new full port 1" ball valves on the Reactor Building Chiller Condenser Service Water Sensing lines to reduce the collection of service water silt and debris in the root valves for the condenser pump low flow switch instrument sensing lines on Unit 1 chiller 1K206B.

RMMMARY:

I. No. As discussed in FSAR Section 9.2.12.3, the only safety related portions of the RBCW System are the primary containment piping penetrations and containment isolation valves. The proposed action does not affect safety related portions of the RBCW System. The installation of the new root valves will affect only a portion of the RBCW System associated with the chiller condenser water circulating pump. The proposed modification adds no new function to the RBCW System as described in FSAR Section 9.2.12.3.

II.

~ No.~ The new valves do not alter the intended function of the system and the RBCW System as described in FSAR Section 9.2. 12.3.2.3.The proposed action does not affect operability of the RBCW System nor introduce any new mode of operation.

The present operation of the chiller is not changed and no new failures or malfunctions are created by installation of the new root valves.

III. No. The RBCW System operability does not contribute to any margin of safety as defined in the basis of any Technical Specification. SSES Technical Specifications (Unit 1) Section 3.6.1.7 and Table 3.6.3-1 and their bases were reviewed for applicability. Section 3.6.1.7 contains the requirements for average drywell temperatures during reactor operating conditions 1, 2, and 3. Table 3.6.3-1 includes the RBCW primary containment isolation valves, which will not be affected by the proposed modification.

The proposed. modification affects the non-safety related portions of the RBCW System, which are not addressed in the Technical Specifications.

/ERIN.: 91-158 R REFE  : DCP 90-3060D, Rev. 0 DE R?

Existing root valves (1" globe valves) will be replaced with new full port 1" ball valves on the Reactor Building Chiller condenser Service Water Sensing Lines to reduce the collection of service water silt and debris in the root valves for the condenser pump low flow switch instrument sensing lines on Unit 2 chiller 2K206A.

SBRLRR.

I. No. As discussed in FSAR Section 9.2.12.3, the only safety related portions of the RBCW System are the primary containment piping penetrations and containment isolation valves. The proposed action does not affect safety related portions of the RBCW System. The installation of the new root valves will affect only a portion of the RBCW System associated with the chiller condenser water circulating pump. The proposed modification adds no new function to the RBCW System as described in FSAR Section 9.2.12.3.

II.

~ No.

~ The new valves do not alter the intended function of the system and the RBCW System as described in FSAR Section 9.2. 12.3.2.3 The proposed action does not affect operability of the RBCW System nor introduce any new mode of operation.

The present operation of the chiller is not changed and no new failures or malfunctions are created by installation of the new root valves.

IH. No. The RBCW System operability does not contribute to any margin of safety as defined in the basis of any Technical Specification. SSES Technical Specifications (Unit 2) Section 3.6.1.7 and Table 3.6.3-1 and their bases were reviewed for applicability. Section 3.6.1.7 contains the requirements for average drywell temperatures during reactor operating conditions 1, 2, and 3. Table 3.6.3-1 includes the RBCW primary containment isolation valves, which will not be affected by the proposed modification.

The proposed modification affects the non-safety related portions of the RBCW System, which are not addressed in the Technical specifications.

DCP 90-3058D, Rev. 0 Existing root valves (1" globe valves) will be replaced with new full port,l" ball valves on the Reactor Building Chiller Condenser Service Water Sensing Lines to reduce the collection of service water silt and debris in the root valves for the condenser pump low fiow switch instrument lines on Unit 1 chiller 1K206A.

I. No. As discussed in FSAR Section 9.2.12.3, the only safety related portions of the RBCW System are the primary containment piping penetrations and containment isolation valves. The proposed action does not affect safety related portions of the RBCW System. The installation of the new root valves will affect only a portion of the RBCW System associated with the chiller condenser water circulating pump. The proposed modification adds no new function to the RBCW System as described in FSAR Section 9.2.12.3.

II.

~ No.~ The new valves do not alter the intended function of the system and the RBCW System as described in FSAR Section 9.2.12.3.2.3. The proposed action does not affect operability of the RBCW System nor introduce any new mode of operation. The present operation of the chiller is not changed and no new failures or malfunctions are created by installation of the new root valves.

III. No. The RBCW System operability does not contribute to any margin of safety as defined in the basis of any Technical Specification. SSES Technical Specifications (Unit 1) Section 3.6.1.7 and Table 3.6.3-1 and their bases were reviewed for applicability. Section 3.6.1.7 contains the requirements for average drywell temperatures during reactor operating conditions 1, 2, and 3. Table 3.6.3-1 includes the RBCW primary containment isolation valves, which will not be affected by the proposed modification.

The proposed modification affects the non-safety related portions of the RBCW System, which are not addressed in the Technical Specifications.

0

~ER~.: 91-160 DCP 89-9057, Rev. 0 Install a permanent battery cell charging facility for maintaining spare 24, 125 and 250 V DC Class 1E battery cells in the fully charged state for immediate replacement of any failed cells on the Class 1E battery banks.

I. No. FSAR 8.3.1, AC Power System; 8.3.2, DC Power Systems; 9.4.4, Turbine Building Ventilation System and Chapter 15 were reviewed. Chapter 15, Accident Analysis, does not address the charging of spare battery cells in the Turbine Building. The proposed modification does not affect any accident mitigation or initiation as described in Chapter 15. The modification does not interface, directly or indirectly, with any equipment that is important to safety.

II. No. The installation of the spare battery cell charging facility will be installed in a non-seismic Category I structure (Unit 2 Turbine Building) in an area in which no Class 1E equipment is located. The spare battery cell charging facility is not required to function or interface with any equipment that is required to function during and after a design basis event. The installation and design will be consistent with accepted PP&L standards and practices.

HI. No. The Technical Specification Bases have been reviewed, specifically Section 3/4.8.1. Neither the referenced section nor any other Technical Basis was found to be affected by the proposed change.

No safety related systems are affected, thus they will continue to meet single failure criteria and their ability to function in the event of a SSE, LOCA/LOOP, or other DBA.

NL-91-025, Rev. 0 Evaluate the interim operation of the Unit 1 RBCCW System as a contaminated system until the source of the contamination can be identified and the system decontaminated.

I. No. Affected systems have no safety related functions. Performance of the system is in no way affected by operation as a contamiiiated system. Rupture of the system willbe contained and not released to the environment. Per the FSAR 9.2.1.3 and 9.2.2.3, failure of the system (SW and RBCCW) will not compromise any safety-related system or component or prevent a safe shutdown of the plant.

II. No. System operability is not affected. System has no safety related functions.

III. No. Affected systems are not addressed in Tech Specs.

QgK~N).: 91-162

DCP 91-9064, Rev. 0 Weld additional bracing to the leg angles of the Reactor Building Closed cooling Water (RBCCW) Head Tank. This will strengthen and stiffen the legs of the RBCCW Head Tank on Unit 2.

/gal AKY:

I. No. The proposed change to the RBCCW Head Tank's legs and its piping does not change any of the original design parameters or operational modes of the RBCCW system. The proposed modification willhave no involvement or impact on the accident analysis already evaluated in FSAR Chapter 15.

II. No. The proposed action will assure that the RBCCW Head Tank does not become a Safety Impact Item, thereby, satisfying FSAR Section 9.2.2.3. i.e., "failure of the system will not compromise any safety related system or component or prevent a safe shutdown of the plant".

III. No. The RBCCW system is not addressed in the Technical Specifications and the cooling water supplied to other systems (or lack of it) cannot affect the basis of any Technical Specification.

$ ER Pg).: 91-163

DCP 91-9063, Rev. 0 Weld additional bracing to the leg angles of the Reactor Building Closed Cooling Water (RBCCW) Head Tank. This will strengthen and stiffen the legs of the RBCCW Head Tank on Unit 1.

SBQUAE I. No. The proposed change to the RBCCW Head Tank's legs and its piping does not change any of the original design parameters or operational modes of the RBCCW system. The proposed modification will have no involvement or impact on the accident analysis already evaluated in FSAR Chapter 15.

II. No. The proposed action will assure that the RBCCW Head Tank does not become a Safety Impact Item, thereby, satisfying FSAR Section 9.2.2.3, i.e., "failure of the system will not compromise any safety related system or component or prevent a safe shutdown of the plant".

III.

~ No.~

~

~

The RBCCW system is not addressed in the Technical Specifications and the cooling water supplied to other systems (or lack of it) cannot affect the basis of

~ ~

any Technical Specification.

/ERIN.: 91-164 E: DCP 90-9061, Rev. o DE RI HAN E:

Install new differential pressure control loop components in the Turbine Building HVAC return air, and filtered exhaust air systems to better detect pressure changes and to more accurately control the dampers which modulate the systems air flow for Unit 2.

SGbgEAR Y:

I. No. FSAR Sections 9.4.4, Turbine Building Ventilation System; 11.5.2.1.2, Turbine Building and Radwaste Vent Exhaust Sampler and Sample Radiological Monitoring Systems; 11.5.4.2, Process Monitoring and Sampling Systems, 12.3.4, Area Radiation and Airborne Radioactivity Monitoring Instrumentation; and Section 15.0, Accident Analysis; were reviewed. FSAR Section 15.7, Radioactive Release from Subsystems and Components; explicitly Section 15.7.1.3, Failure of Air Ejector Lines; was reviewed for the potential of being impacted by changes made via this proposed modification. The proposed modification does not impact the design basis of the reviewed systems as related to nuclear safety.

The accidents described in FSAR Section 15.7 do not require any intervention or action by Turbine Building HVAC systems to mitigate the consequences of those accidents. This modification does not alter this relationship by adding requirements necessitating the Turbine Building HVAC system support systems important to safety during an accident, III. No. The Technical Specification Bases have been reviewed, specifically Sections 3/4.3.7.11, Radioactive Gases Effluent Instrumentation, and 3/4.11.2, Radioactive Effluent - Gaseous Effluent. The modification does not alter the ability of the affected systems to perform their intended function and satisfy the Technical Specification Basis.

~ER~N.: 91-165 R REFERE E DCP 90-9060, Rev. 0 DE RIPTI Install new differential pressure control loop components in the Turbine Building HVAC<return air, and filtered exhaust air systems to better detect pressure changes and to more accurately control the dampers which modulate the systems air flow for Unit 1.

KKMARY:

I. No. FSAR Sections 9.4.4, Turbine Building Ventilation System; 11.5.2.1.2, Turbine Building and Radwaste Vent Exhaust Sampler and Sample Radiological Monitoring Systems; 11.5.4.2, Process Monitoring and Sampling Systems, 12.3.4, Area Radiation and Airborne Radioactivity Monitoring Instrumentation; and Section 15.0, Accident Analysis; were reviewed. FSAR Section 15.7, Radioactive Release from Subsystems and Components; explicitly Section 15.7.1.3, Failure of Air Ejector Lines; was reviewed for the potential of being impacted by changes made via this proposed modification. The proposed modification does not impact the design basis of the reviewed systems as related to nuclear safety.

The accidents described in FSAR Section 15.7 do not require any intervention or action by Turbine Building HVAC systems to mitigate the consequences of those accidents. This modification does not alter this relationship by adding requirements necessitating the Turbine Building HVAC system support systems important to safety during an accident.

III. No. The Technical Specification Bases have been reviewed, specifically Sections 3/4.3.7.11, Radioactive Gases Effluent Instrumentation, and 3/4.11.2, Radioactive Effluent - Gaseous Effluent. The modification does not alter the ability of the affected systems to perform their intended function and satisfy the Technical Specification Basis.

DCP 89-3003D, Rev. 0 DE I Modifications to the intake structure instrument air system (ISIA) including replacement of drain trap, prefilters, reorientation of membrane dryers, relocation of tap for air receiver OT-086, reduction of system maximum design pressure and modification of compressor discharge check valves all to improve performance of the system.

I. No. FSAR Section 9.3.1.4 was reviewed in consideration of this question. This section states that failure of the intake structure instrument air system will not endanger the operation of any safety-related instruments and controls.

The River Water Make-Up (RWMU) System and its support systems (Screen Wash, etc.) are the only systems that rely on the ISIA system. FSAR Sections 9.2, 9.5 and 10.4 were reviewed to assure no indirect adverse affects to equipment important to safety would be caused by the proposed actions. The RWMU system is not discussed in the FSAR as confirmed by the review of the above mentioned sections. In addition, no safety related equipment relies directly upon the RWMU system.

II. No. The reliability of both the ISIA system and the RWMU system is being improved and there are no systems that rely directly on the RWMU system.

III. No. The ISIA system nor the RWMU system are mentioned in the basis for any technical specification. The basis for the service water systems (ESW &

RHRSW, Section B3/4.7. 1), which use river water as makeup, are not affected by the proposed action. The spray pond at its minimum tech spec level has enough water for the design basis accident.

SCP J91-1027 & J91-1028, Rev. 0 Increase in the maximum allowable setpoint of the Unit 1 and Unit 2 Turbine Building HVAC supply filter high differential pressure switches to reduce the frequency of filter changes and the attendant maintenance costs.

I. No. The setpoint changes require no physical changes to the plant design, function, accuracy, response time, reliability, actuation logic or to any other component or system. The setpoint changes also do not require any physical modification to the switches or to any other components. The setpoint changes are intended to allow greater pressure differentials across the filters and thereby reduce the costly maintenance requirements associated with frequent filter changes. Calculation determines that the increased differential has no adverse effect on the filter.

II. No. The setpoint changes do not affect the design or function of the high differential pressure switches or of any other components. Also, the changes do not cause any component or system to operate outside the performance boundaries described in the FSAR.

III. No. Neither the Turbine Building Supply Systems nor the high differential pressure switches of these systems are governed by the Technical Specification.

SCP E911049, Rec. 0 Achieve a better coordination of instantaneous overcurrent trip units of breaker OA550 downstream of the start-up Bus 10 breaker ¹52-10308.

I. No. The proposed action increases availability of the equipment by insuring that upstream breaker 52-10308 of Bus 10 (OA103) does not trip for a fault on the downstream breaker OA550 feeder. Setpoint changes of the relays are routinely performed in the plant. Since there is no change in the design except for a setpoint change, there is no increase of the probability of occurrence of an accident evaluated in the FSAR.

II. No. The SCP does not propose any physical change to the plant design or any change of circuit design except the setpoint. Therefore, the proposed action does not create a possibility for an accident or malfunction of a different type than evaluated previously in the FSAR.

There are two main categories of equipment affected: Make-Up Water System equipment and Diesel Generator-E auxiliary load equipment. No technical specifications are associated with the Make-Up Water System. Diesel generators are covered by Technical Specification Section 3.8.1.1 and 3.8.1.2. The proposed action does not involve any change to the Diesel Generator design.

ging goal.: 91-169 DCP 90-3093B For Unit 1, installation of a second isolation valve just upstream of valve HV-11618, "Condensate Demineralizer drain valve to LRW" to provide a positive means of isolating the drain header when the butterfly valve fails to provide adequate isolation.

I. No. Section 9.2.10 does not specifically discuss the Condensate Demineralizer System and, therefore, does not need to be addressed in this section. Section 10.4.6.1 states that the design basis of the Condensate Cleanup System is to maintain condensate purity by removing various contaminants. This section also states the not-to-exceed effluent quality for the demineralizers. The design basis for the system is not affected by the proposed action. The accident Analysis contained in Chapter 15 is unaffected by the proposed modification.

II. No. In this modification, there is no equipment important to safety and no interface with safety equipment or systems. This modification merely supplies redundant isolation capability to the condensate demineralizer drain line to LRW.

III. No. This modification is non-safety related and does not affect systems having Technical Specification requirements. This change adds a valve which is designed and built to the same standards, codes and design conditions as the condensate demineralizer system. The new valve will not in any way degrade the operation of the condensate demineralizer system.

~ER~N.: 91-170 R REFERE E: DCP 90-3064A, Rev. 0 DE RIPTI N F HA E:

Removal of the existing Unit 1 Reactor Building Chiller "B" vent and drain lines and replace them with vent and drain lines designed to alleviate clogging problems.

KMMAilY:

I. No. The proposed modification adds no new function to the RBCW System as described in FSAR Section 9.2.12.3. The proposed modification to the equipment drains does not modify the design parameters of the Equipment and Floor Drainage System as described in FSAR Section 9.3.3.3.

II. No. After the modification, the RBCW System willperform its design bases function as described in FSAR Section 9.2.12.3.1. The proposed modification to the equipment drains does not modify the design parameters of the Equipment and Floor Drainage System as described in FSAR Section 9.3.3.3. No new functions are added to the Liquid Waste Management System as described in FSAR Section 11.2.

The RBCW System operability does not contribute to any margin of safety as defined in the basis of any Technical Specification. SSES Technical Specifications (Vnit 1) Section 3.6.1.7 and Table 3.6.3-1 and their bases were reviewed for applicability. Section 3.6.1.7 contains the requirements for average drywell temperatures during Reactor Operating Conditions 1, 2, and 3. Table 3.6.3-1 includes the RBCW primary containment isolation valves, which will not be affected by the proposed modification.

DCP N-3QEDA, R . 0 Removal of the existing Unit 2 Reactor Building Chiller "A" vent and drain line and replace them with vent and drain lines designed to alleviate clogging problems.

I. No. The proposed modification adds no new function to the RBCW System as described in FSAR Section 9.2.12.3. The proposed modification to the equipment drains does not modify the design parameters of the Equipment and Floor Drainage System as described in FSAR Section 9.3.3.3.

II. No. After the modification, the RBCW System willperform its design bases function as described in FSAR Section 9.2.12.3.1. The proposed modification to the equipment drains does not modify the design parameters of the Equipment and Floor Drainage System as described in FSAR Section 9.3.3.3. No new functions are added to the Liquid Waste Management System as described in FSAR Section 11.2.

The RBCW System operability does not contribute to any margin of safety as defined in the basis of any Technical Specification. SSES Technical Specifications (Unit 2) Section 3.6.1.7 and Table 3.6.3-1 and their bases were reviewed for applicability. Section 3.6.1.7 contains the requirements for average drywell temperatures during Reactor Operating Conditions 1, 2, and 3. Table 3.6.3-1 includes the RBCW primary containment isolation valves, which will not be affected by the proposed modification.

~ER~N.: 91-172 RRRRRRRRllRNNllD:D99I83038A, R . D DE RlPT1 F HA E:

Removal of the existing Unit 1 Reactor Building Chiller "A" vent and drain lines and replace them with vent and drain lines designed to alleviate clogging problems.

888RMARY:

I. No. The proposed modification adds no new function to the RBCW System as described in FSAR Section 9.2.12.3. The proposed modification to the equipment drains does not modify the design parameters of the Equipment and Floor Drainage System as described in FSAR Section 9.3.3.3.

II. No. After the modification, the RBCW System willperform its design bases function as described in FSAR Section 9.2.12.3.1. The proposed modification to the equipment drains does not modify the design parameters of the Equipment and Floor Drainage System as described in FSAR Section 9.3.3.3. No new functions are added to the Liquid Waste Management System as described in FSAR Section 11.2.

The RBCW System operability does not contribute to any margin of safety as defined in the basis of any Technical Specification. SSES Technical Specifications (Unit 1) Section 3.6.1.7 and Table 3.6.3-1 and their bases were reviewed for applicability. Section 3.6.1.7 contains the requirements for average drywell temperatures during Reactor Operating Conditions 1, 2, and 3. Table 3.6.3-1 includes the RBCW primary containment isolation valves, which will not be affected by the proposed modification.

/ERIN.: 91-173 R REFE E: SCP J91-1016 & 1017, Rev. 0 DE R1PTI F H E:

Lower the low alarm setpoint of pressure switches PSHL-17192 and PSHL-27192 to reduce nuisance alarms for the charcoal absorber train outlet pressure.

KMhfARY:

I. No. Switches PSHL-17192 and PSHL-27192 do not provide input to any safety functions. Since the switches do not provide input to safety functions, the FSAR has not evaluated these switches as devices that impact the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety. SpecificaHy, FSAR chapters 11.3.2 and 15 have been reviewed and found to be unaffected.

II. No. FSAR Table 3.2-1 classifies the major components of the condenser off-gas system as quality class D. However, switches PSHL-17192 and PSHL-27192 are classified non-safety. The setpoint changes change the low alarm setpoint of switches PSHL-17192 and PSHL-27192 only and do not affect any safety functions.

HI. No.

~ ~ The low alarm setpoint of switches PSHL-17192 and PSHL-27192 is not governed by Technical Specifications.

JERKIN.: 91-174 R REFBRE E: DCP 90-3060B, Rev. 0 E RlPTI Installation of sensors required for the new Unit 2 Reactor Building Chiller "A" diagnostic system.

$ Qg Y:

I. No. As discussed in FSAR Section 9.2.12.3, the only safety related functions of the RBCW System are performed by the primary containment piping penetrations and containment isolation valves. The proposed action does not affect safety related portions of the RBCW System. The addition of the sensors will only affect portions of the SW and RBCW systems associated with chiller 2K206A. The proposed modification adds no new function to the RBCW System as described in FSAR Section 9.2. 12.3.

II. No. The proposed modification will add sensors required for a diagnostic system which do not alter the intended function of the SW system and the RBCW System as described in FSAR Section 9.2.1.1 and 9.2.12.3. The sensors willbe designed and installed in accordance with the construction codes and standards listed in FSAR Section 3.

The proposed action does not affect operability of the RBCW System nor introduce any new mode of operation. The present operation of the chiller is not changed and no new failures or malfunctions are created by addition of the sensors.

HI. No. SSES Technical Specifications (Unit 2) Section 3.6.1.7 and Table 3.6.3-1 and their bases were reviewed for applicability. Technical Specification Section 3.6.1.7 contains the requirements for average drywell temperatures during reactor operating conditions 1, 2 and 3. These requirements are not affected by this modification. Technical Specification Table 3.6.3-1 includes the RBCW primary containment isolation valves, which will not be affected by the proposed modification.

DCP 89-3006B, Rev. 0 D 1 Installation of a second isolation valve in the feedwater heater tube side vent piping.

SgdgfkkR.':

The new valves interface with the feedwater system which is a non-Q system and not required for the safe shutdown of the plant. FSAR Section 10.4.7 and Chapter 15 were reviewed for this analysis. This modification does not jeopardize the function or alter the operation of any safety related equipment.

This modification will significantly reducelcompletely eliminate leakage of feedwater into the radwaste system, and enhance system availability by providing a backup isolation capability.

This modification is non-Q and only interfaces with non-Q systems. In addition, the installation and design will be in accordance with all original codes and standards. All components being replaced or added are of similar physical characteristics to the original components and, therefore, any type of postulated component failure will not change.

III. No. This modification is non-safety related and does not affect systems having Technical Specification requirements. This change adds and replaces valves which are designed and built to the same standards and design conditions as the feedwater system. The new valves will not in any way impact the operation of the feedwater system.

$ER~.: 91-176

NL-91-011, Rev. 2 E R1 Fuel Pool Cleanout Project to remove all irradiated reactor vessel components (excluding spent fuel) from the Unit 1 and Unit 2 spent fuel pools.

I. No. Accident analyses contained in the FSAR for a Fuel Handling Accident (FSAR 15.7-4) address accidents far more severe than similar accidents postulated for this project. Section 15.7.5 in the FSAR addresses a Spent Fuel Cask Drop Accident and concludes that this event is not credible due to the use of redundant rigging and a single failure proof crane. The Fuel Pool Cleanout Project activities are therefore bounded by the FSAR analyses noted.

II. No. Two types of accidents are postulated: a loaded cask drop scenario and draining of the cask storage pit with a significant inventory of irradiated hardware becoming uncovered. Cask drop accidents are not considered a credible accident per FSAR Section 15.7.5 due to redundant rigging and a single failure proof crane, preventing a single failure.

Loss of water shielding in the cask storage pit is a credible scenario for evaluation. The airborne radiological consequences of this event would be bounded by the Fuel Handling Accident analyzed in FSAR section 15.7.4.

IG. No. All Fuel Pool Cleanout Project activities will be conducted in accordance with plant Technical Specifications.

R N .: 91-177 R REFE  : DCP 90-3075, Rev. 1 D RI Installation of an ultrasonic flow monitoring system for both of the Unit 1, 96 inch pipes supplying circulating water to tlie condenser.

BVMMARY:

I. No. As discussed in FSAR Section 10.4.5, the Circulating Water System has no safety related functions. Section 15.2.5 of the FSAR analyzes a loss of condenser vacuum whereby there is a loss of one or more of the Circulating Water Pumps.

This modification willnot increase the probability of a loss of condenser vacuum.

In addition, any type of water leakage due to the failure of penetration assemblies would be less severe than a Circulating Water System Rupture discussed in Section 10.4.1 of the FSAR.

II. No. In this modification, there is no equipment important to safety, and there is no interface with safety equipment or systems. With the transducers, pull boxes (and shields), and piping welded onto the pipes per PP&L Specification, and ASME Section IX, there should be no concern with the ultrasonic flow monitoring system affecting another component or system. As analyzed, the failure modes of this installation are covered by those accidents previously evaluated in the FSAR.

HI. No. This modification affects the Circulating Water System which has no safety-related function. In particular, flow measurement of circulating water has no direct interface with any safety function in the plant.91-178 SCP J91-1044, Rev. 0 Change in setpoints for TSH-17702A2 and TSH-17702B2 from 130'F to 123 F for temperature switches associated with drywell unit coolers 1V411A and 1V411B to illiminate numerous false alarms.

IUMNNRY:

I. No. None of the design basis accidents discussed in FSAR section 15 are affected by this change. This change does not alter the failure modes associated with either TSH-17702A2 & TSH-17702B2 or the drywell unit coolers 1V411A & 1V411B.

These switches provide alarms to indicate either failed chilled water flow through the unit coolers or high intake air temp. There are other Reactor Building Cooling Water/Drywell Cooler system alarms which provide the same information.

Further, FSAR Sections 9 4.5, 7.3.1.1b.8.5.6 and tables 9.4-8, 9.4-9, 9.4-)0 provide details of Drywell Cooling System. These sections and tables were reviewed for applicability. The proposed change has no impact on these sections and tables.

II. No. The function and performance of the drywell unit coolers 1V411A&B is not affected by this change. This change does not alter the failure modes of this equipment and it does not modify plant response to a design basis accident.

Further, FSAR Section 9.4.5.1 discusses the design basis for drywell cooling system including unit coolers 1V411A&B. This change does not alter the function and the performance of any of the above drywell unit coolers and it does not alter the design basis discussed in the FSAR. This change will eliminate false alarms.

III. No. SSES Technical Specification sections 3/4.6.1.7 was reviewed for applicability.

This change neither reduces the operability of any equipment nor requires additional surveillance to ensure plant safety. This change does not change the function or performance of drywell unit coolers 1V411A&B.

~BR lg.:~~ 91-179 R E SCP J9ILlOI3R*,. 0 Increase the setpoint of the 'instrument failure'unction of FSL-17630A/B, Setpoint ¹2, to prevent unnecessary trips of Emergency Switchgear and Load Center Rooms Cooling System on high air flow. Decrease the setpoint for the 'low flow'unction of FSL-17630 A/B, Setpoint

¹1, to reduce the instrument response time on system start-up.

BUMNARY:

I. No. There are no design base accidents which specifically address the operation of the Emergency Switchgear and Load Center Room Cooling System. However, it is defined as an Engineering Safety Feature by the FSAR (Section 7.3.1), and the system is required to keep 1E equipment within defined temperature limits. This setpoint change increases the reliability of the system by preventing unnecessary trips when both trains of fans are operating, and during system start-up of a single fan. The reliability of the Control Structure ventilation systems is enhanced by the increased reliability of the Emergency Switchgear and Load Center Room Cooling System.

This setpoint change does not'affect the design or operation of the Emergency Switchgear and Load Center Room Cooling System or any of the Control Structure Habitability Systems. This change increases system reliability by eliminating the possibility of unnecessary trips during two fan operation and system start-up of a single fan. There is no basis for the value of the 'low flow'rip or the existence of a high flow trip in this system. Therefore, decreasing the

'low flow'etpoint and redefining two of the trips as an instrument supervisory trip does not impact any safety function III. No. The control Room Emergency Outside Air System is the only Control Structure Habitability System addressed by the Technical Specifications. It is not affected by the operation of the Emergency Switchgear and Load Center Room Cooling System or the Control Structure Chillers. The operation and instrumentation of the Emergency Switchgear and Load Center Room Cooling System is not addressed specifically in the Technical Specifications. This change increases the reliability of the Emergency Switchgear and Load Center Room Cooling System and therefore Control Structure Chillers during emergency operation.

0

/ERIN.: 91-180 R REFEREN E: DCP 91-3023B, Rev. 0 DE RlPTI F HAN E:

Replace existing target-type flowmeter in Unit 2 Cooling Tower Blowdown Flow Instrumentation with an electromagnetic flowmeter for improved reliability.

RRMMARY:

I. No. Radioactive Releases from Subsystems and Components and I.iquid Radwaste System Accidents are addressed in Section 15.7 and Section 15.7.2, respectively, of the FSAR. Engineered Safety Features are addressed in Chapter 6 of the FSAR. Circulating Water System and River Water Makeup System are addressed in Section 10.4.5 of the FSAR. Liquid Waste Management System is addressed in Section 11.2 of the FSAR.None of the referenced systems willbe impacted by the proposed modification.

II. No. This modification only replaces the existing target-type flow element and transmitter with a new electromagnetic flow element and transmitter to upgrade the performance of a non-safety related instrument loop. The new instrumentation will perform all the same functions as the existing instrumentation. No new functions or interfaces with other devices or systems will be introduced. The upgraded non-safety related instrument loop will continue to provide the 5000 gprn minimum blowdown flow interlock that is a permissive for radwaste discharge.

Instrument loop failure is not specifically addressed in the FSAR; however, a failure of the upgraded loop will have the same effect as that of the existing loop.

III. No. Radiological releases are discussed in sections 3.3.7.10, 4.3.7.10, and 3.11.1.1 of the Plant Technical Specifications, which provides the bases and limiting conditions for operation for liquid effluent releases and associated instrumentation. The upgraded instrument loop accuracy is equivalent to the existing instrument loop accuracy; therefore, the upgraded loop will continue to use 5000 gpm for the minimum blowdown flow permissive to ensure minimum dilution flow is available prior to allowing radwaste effluent discharge valves to be opened. The upgraded loop will ensure that the minimum requirements for cooling tower blowdown dilution flow as contained in the ODCM are maintained.

/ERIN.: 91-181 R  : DCP 91-3023A, Rev. 0 DE Replace existing target-type flowmeter in Unit 1 Cooling Tower Blowdown Flow Instrumentation with an electromagnetic flowmeter for improved reliability.

SgdgEARY:

I. No. Radioactive Releases from Subsystems and Components and Liquid Radwaste System Accidents are addressed in Section 15.7 and Section 15.7.2, respectively, of the FSAR. Engineered Safety Features are addressed in Chapter 6 of the FSAR. Circulating Water System and River Water Makeup System are addressed in Section 10.4.5 of the FSAR. Liquid Waste Management System is addressed in Section 11.2 of the FSAR. None of the referenced systems will be impacted by the proposed modification.

II. No. This modification only replaces the existing target-type flow element and transmitter with a new electromagnetic flow element and transmitter to upgrade the performance of a non-safety related instrument loop. The new instrumentation will perform all the same functions as the existing instrumentation. No new functions or interfaces with other devices or systems will be introduced. The upgraded non-safety related instrument loop will.

continue to provide the 5000 gpm minimum blowdown flow interlock that is a permissive for radwaste discharge. Instrument loop failure is not specifically-addressed in the FSAR; however, a failure of the upgraded loop will have the same effect as that of the existing loop.

III. No. Radiological releases are discussed in sections 3.3.7.10, 4.3.7.10, and 3.11.1.1 of the Plant Technical Specifications, which provides the bases and limiting conditions for operation fpr liquid effluent releases and associated instrumentation. The upgraded instrument loop accuracy is equivalent to the existing instrument loop accuracy; therefore, the upgraded loop will continue to use 5000 gpm for the minimum blowdown flow permissive to ensure minimum dilution flow is available prior to allowing radwaste effluent discharge valves to be opened. The upgraded loop will ensure that the minimum requirements for cooling tower blowdown dilution flow as contained in the ODCM are maintained.

/ERIN.: ~~ 91-182 DCP 90-9040, Rev. 0 E RI To isolate the flow switch from the pump control circuit in the event of a fire, this change provides a separate fuse for the flow switch components in each Control Structure Chilled Water circulating pump control circuit.

Y:

I. No. This modification does not increase the probability of occurrence or the consequences of the accidents described in FSAR Chapters 6 and 15. FSAR Section 9.2.12.1.2.3 states that one of the two chilled water trains will be in operation during all modes of plant operation including LOCA. Under normal conditions, one chilled water train willbe operating and the other train will be on standby.

Failure in the new circuit (e.g. blown fuse in flow switch circuit) will only disable the low chilled water flow automatic start of that loop along with sending an alarm to inform operators of circuit failure. Manual start capability of the chilled water pump will be unaffected.

II. No. FSAR Sections 9.4.2.2.2 and 9.2.12.1.3 imply that the Control Structure Chilled Water System is single failure proof; however, a fire induced failure involving the existing OP162A/B control circuitry may cause a complete loss of the Chilled Water System. The Control Structure Chilled Water System will be placed in its design basis configuration by the proposed action and willpreclude the possibility of the single failure described above.

III. No. The Control Structure Chilled Water System is not mentioned in the bases for any Technical Specifications. Tech. Spec. Basis 3/4.8.1, 3/4.8.2, and 3/4.8.3, "A.C.

Sources, D.C. Sources and onsite power distribution" - "The operability of the A.C. and D.C. power sources and associated distribution systems during operation ensures that sufficient power willbe available for (1) the safe shutdown of the facility and (2) the mitigation and control of accident conditions within the facility." The proposed action eliminates a single failure and thus improves the reliability of the room cooling system for the A.C. power sources. This will reduce the possibility of malfunctions to this equipment.

ER N .: 91-183

DCP 91-3005, Rev. 0 Installation of a double bay panel in the control room to house the electronic drawers for the Reactor Recirculation Pump Vibration Monitoring System for Unit 1 and 2. Also, the SPING insert is removed from panel OC669 and put in the second bay of the new double bay panel.

I. No. The Reactor Recirculation Pump Vibration Monitoring System is Non-Q, Non-Class 1E and does not perform any control functions. This modification simply removes the components from a temporary panel and installs them in a permanent control room panel. The Reactor Recirculation Pump Vibration Monitoring System is not used in any FSAR Chapter 6 or 15 accident analysis.

Any credible electrical failure of the system would not affect any safety system or the ability of the safety system to perform its safety functions. The Vent Stack Radiation Monitoring System is also Non-Q, non-Class 1E and does not perform any control functions. This modification removes the components from the SPING console and installs them in the second bay of the new, permanent control room panel. This modification does not alter the capability of the radiation monitoring system to perform as functionally described in the FSAR.

II. No. This modification removes the equipment for two existing systems (Vent Stack Radiation Monitoring System and Reactor Recirculation Pump Vibration Monitoring System) from their separate control panels and installs them in a single, double bay, control panel. The functional design and operation of the two systems remain unchanged since all equipment except the enclosures for the equipment are reused for this modification. The failure modes after this modification remain the same as that for the same systems before the modification and would in no way affect any safety related systems or components.

IH. No. A review of the applicable portions of the Technical Specification, Sections 3/4.3.7.11 and 3/4.11.2.1 indicates that the intent of the Vent Stack Radiation Monitoring System is to ensure compliance with the guidelines of 10CFR Part 20.

In that this modification neither improves nor affects the capability of the Vent Stack Radiation Monitoring System to perform its function, the modification does not change the margin of safety. The Reactor Recirculation Pump Vibration Monitoring System is not addressed in the Technical Specifications. The system performs no safety functions and is isolated from equipment which perform safety related functions.

~ER~.: 91-184 R RE  : DCP 89-3027 & 3028, Rev. 0 D RT Replacement of existing 3" butterfly and 2" globe valves on RCIC, HPCI, CS and RHR cooling coil supply lines and RCIC, HPCI and CS cooling coil return lines with stainless steel ball valves to improve maintainability. Also, installation of new stainless steel flow orifice and spool pieces to provide more accurate long term flow measurements.

SgdfMARY I. No. The design basis for the ESW System is discussed in Chapter 9, Section 9.2.5 of the FSAR. An evaluation of components in the system affected by this modification shows that the ESW System design basis will not change as a result of these modifications. This modification is required to simply upgrade the existing cooling coil isolation/throttle valves, interconnecting pipe and flow elements for increased reliability and maintainability of this equipment. This modification will provide greater assurance that the ESW System is performing its design function by increasing the reliability/accuracy of flow measurements at the cooling coils.

The upgrading of isolation valves and replacement of flow orifices/spool pieces with upgraded materials does not create a possibility for an accident or malfunction of a different type than any previously evaluated in Chapter 15 of the FSAR. The ESW System design basis'ill not change as a result of this modification.

III, No. The only technical specification potentially affected by this modification would be Section 3/4.7.1 paragraph 3.7.1.2, because that section deals with operability of ESW system and this modification will replace valves and flow elements that are used during ESW operation. The difference in pressure drop due to the new ball valves was evaluated against the valves assumed in the design calculations and it was found insignificant. Therefore, the system performance and calculations important to safety will not be affected by these changes.

/ERIN.: 91-185 R REFERE E: DCP 91-3011, Rev. 0 DE RIPTI A Replacement of existing underrated fuses in the non-class 1E battery chargers'elay enclosure, motor control centers, and uninterruptible power supplies within the 250V DC system.

RQMMARY:

I. No. The function of the connected load systems are not modified by this action.

Additionally, the failure mode and effect analysis provided in FSAR Table 8.3-22 for the 250V DC System is not impacted by this change. The replacement fuses and fuse blocks are specified to the same criteria as the existing equipment with the exception that they are tested to a minimum of 300V DC and 20,000 amps interrupting capacity in order to assure interruption of faulted and/or overloaded circuits and prevent further propagation of the failures, II. No. Installation of these replacement fuses and fuse blocks does not alter the function or loads of the Non-Class 1E loads connected to the Class 1E busses. This modification does not generate any adverse effects on the loads supplied by safety systems. The proposed modification will not introduce any failure modes different than already analyzed in FSAR Table 8.3-22.

III. No. This modification is implemented to decrease the challenge on the upstream interrupting device by increasing the assurance that the local equipment fuses will perform their design function during equipment faults and/or overloads. This limits to the extent practical the effect of Non-Class 1E equipment faults and overloads on the Class 1E 250V DC system. Therefore, the proposed action does not reduce the margin of safety as defined in the basis for any Technical Specification.

/ERIN.: 91-186 R RE  : DCP 90-3093C, Rev. 0 DE RIPTI F HA E:

This modification installs a manual 3" ball valve in the common 3" header between the 20" Condensate Effluent Header and the Feedwater Pump Seal Water System. The addition of the 3" ball valve will reduce maintenance time by providing positive isolation for repair of the seal water system components and eliminate the requirement for costly freeze seals.

SBMLL'.

No. FSAR Chapter 15 (Accident Analysis) and FSAR Section 10.4.7 (Condensate and Feedwater) hav'e been reviewed. The Accident Analyses contained in Chapter 15 are unaffected by the proposed modification. The new valve, located in the Turbine Building, elevation 676,'nterfaces with the feedwater pump seal water system which is a non-Q system and not required for the safe shutdown of the plant. This modification does not jeopardize the function or alter the operation of any safety related equipment or increase the radiological consequences of a design basis accident. This modification will significantly improve the seal water isolation capability without any adverse impact on reactor feed pump design or function.

II. No. The installation and design will be in accordance with all original codes and standards. All components being replaced or added are of similar physical characteristics to the original components and, therefore, any type of postulated component failure will not change.

III. No. This modification is non-safety related and does not affect systems having Technical Specification requirements. This change adds a valve which is designed and built to the same standards, codes and design conditions as the condensate system. The new valve will not in any way degrade the operation of the condensate system, or more specifically, the Reactor Feed Pump Seal Water System.

R REFERE E: DCP 89-3006C, Rev. 0 DE RIPTI F H E:

Installation of a second isolation valve upstream of the existing vent or drain line isolation valve in the Unit 1 feedwater heater shellside vent and drain piping to facilitate maintenance and prevent leak-by.

SVMfAllY I. No. The new valves, located in the Turbine Building between elevations 656'nd 699', interface with the feedwater system which is a non-Q system and not required for the safe shutdown of the plant. FSAR Section 10.4.7, 10.4.10 and Chapter 15 were reviewed for this analysis. This modification does not jeopardize the function or alter the operation of any safety related equipment.

This modification will significantly reduce/completely eliminate leakage of feedwater into the radwaste system, and enhance system availability by providing a backup isolation capability.

H. No.

~ ~ This modification is non-Q and only interfaces with non-Q systems. In addition, the installation and design will be in accordance with all original codes and standards. All components being replaced or added are of similar physical characteristics to the original components and, therefore, any type of postulated component failure will not change.

III. No. This modification does not reduce any margin of safety defined in the Basis of the Technical Specifications. This modification is non-safety related and does not affect systems having Technical Specification requirements. This change adds valves which are designed and built to the same standards and design conditions as the feedwater heater drain and vent system. The new valves will not in any way change the operation of the feedwater heater drain and vent system.

R REFERE: NL-91-026, Rev. 0 RIPTI Provide new bounding conditions for either the former or future Mobile Processing System Facility decontamination system supplied by Plymouth Diversified Services.

SlgMARY I. No. The location of the Mobile Processing System Facility outside the plant buildings, i.e. in the environment, does not have a direct affect on the potential accidents as evaluated in FSAR Chapter 15. Nuclear Safety Operational Analysis (NSOA) in FSAR Section 15A.6.2.3.1. Radioactive Material Release Control, describes.

the significance of preventing excessive release of radioactive materials to the environment. Safety action required is a monitoring system to ensure limits given in 10CFR20 and 10CFR50 are not exceeded.

II. No. The Mobile Processing System Facility is not physically connected to the plant.

This facility will be used to process liquid oily waste or petroleum based materials. An accident or malfunction of a different type than any evaluated previously in the SAR include a fire which consumes total liquid inventory of facility, releasing activity to the air. Waste oil trailer fire offsite dose analysis and concentration limits, shows that at a maximum activity limit of 1.05E3 p, Ci, Technical Specification 3.11.2.1.b limit of 1500 mrem/yr, is not exceeded.

III. No. Bases for Technical Specification 3/4.7.2 ensures that Control Room personnel will not exceed 5 rem whole body in accordance with 1OCFR50, Appendix A, Criteria 9. Unrestricted areas and members of the public are protected from concentrations of gaseous effluents exceeding 10CFR20, Appendix B, Table 11, Column 1 . Member of the public dose is kept "as low as reasonably achievable" (ALARA)at site boundary in accordance with 10CFR20 and 10CFR50 Appendix I. Limits to minimize dose rate from gaseous effluents shall be included in procedures. Mobile Processing System Facility total activity is limited to lm Ci.

Liquid oily waste or Petroleum based materials containing noble gasses (i.e. Kr Xe) and I-131 as determined by isotopic analysis will not be processed. Facility fire detection and protection shall be available prior to and during processing.

~Eel.: 91-189

NL 90-029, Rev. 1 Review of proposed use of the Temporary Laundry Facility that willbe used to sort, store, ship and receive protective clothing.

I. No. The Temporary Laundry Facility is a stand alone, temporary facility. The facility has no safety-related functions, and the facility does not interface with any safety-related systems. The laundry trailers do not play a critical role in plant operations.

II. No. Accident analyses in the FSAR are not performed for scenarios involving,the processing of contaminated protective clothing. However, FSAR 15.7 contains analysis on Radioactive Release from Subsystems and Components. A worst case accident scenario for the Temporary Laundry Facility could result in radioactive release of up to 36.4 millicuries (assuming all activity is CO-60). A dose evaluation indicates a maximum dose at the site boundary of 372 millirem for a release of 36.4 millicuries. This dose does not require a change to the FSAR.

III. No. The Facility is a temporary stand-alone laundry processing area. The facility does not interfere with any other system or component governed by the Technical specifications.

~ER~.: 91-190 R REFEREN E: NL-91-026, Rev. 1 DE RIPTI N F E:

Provide new bounding conditions for either the former or future Mobile Processing System Facility decontamination system supplied by Plymouth Diversified Services.

ltLMMARY:

I. No. The location of the Mobile Processing System Facility outside the plant buildings, i.e. in the environment, does not have a direct affect on the potential accidents as evaluated in FSAR Chapter 15. Nuclear Safety Operational Analysis (NSOA) in FSAR Section 15A.6.2.3.1. Radioactive Material Release Control, describes the significance of preventing excessive release of radioactive materials to the environment. Facility atmosphere and process vessel vent system willbe sampled and analyzed to determine if release limits were exceeded.

H. No. The Mobile Processing System Facility is not physically connected to the plant.

This facility will be used to process liquid oily waste or petroleum based materials. An accident or malfunction of a different type than any evaluated previously in the SAR include a Fire which consumes total liquid inventory of facility, releasing activity to the air. Waste oil trailer fire offsite dose analysis and concentration limits, shows that at a maximum activity limit of 1.05E3 p Ci, Technical Specification 3.11.2.1.b limit of 1500 mrem/yr. is not exceeded.

III. No. Bases for Technical Specification 3/4.7.2 ensures that Control Room personnel will not exceed 5 rem whole body in accordance with 1OCFR50, Appendix A, Criteria 9. Unrestricted areas and members of the public are protected from concentrations of gaseous effluents exceeding 10CFR20, Appendix B, Table 11, Column 1 . Member of the public dose is kept "as low as reasonably achievable" (ALARA)at site boundary in accordance with 10CFR20 and 10CFR50 Appendix I. Limits to minimize dose rate from gaseous effluents shall be included in procedures. Mobile Processing System Facility total activity is limited to lm Ci.

Liquid oily waste or Petroleum based materials containing noble gasses (i.e.

Kr,Xe) and I-131 as determined by isotopic analysis will not be processed.

Facility fire detection and protection shall be available prior to and during processing.

h

/ERIN.: 91-191 R REFEREN E: SCP J-91-1046 DE R?PTI N F HAN E:

Decrease the setpoint of the Unit 1 Condensate Pump Room High Temperature TSH-17442 to prevent instrument inaccuracies from causing the temperature of the controlled space to be exceeded and to avoid nuisance alarms.

SEMLAIIF:

I. No. The temperature switch, the Condensate Pump Room Cooling System (CPRCS) and the Condensate System are considered as not important to safety as evaluated in FSAR Sections 9.4.4.2, 9.4.4.3, 10.4.7.1. The reduced temperature switch setpoint causes cooling equipment to begin operation at temperatures well below the maximum allowable limit identified in the FSAR.

The room temperature may be monitored to prevent the maximum allowable limit from being exceeded. As time permits, faulty cooling equipment would be repaired and the fourth unit returned to standby. The change does not affect the design function, accuracy, reliability, response time or actuation logic of the switch or any other component or system. The setpoint change also does not require any physical modification'to the switch or other components. The lower temperatures enhance rather than degrade, system performance.

II. No. The setpoint change does not affect the design or function of the switch or any other component or system. Also, the change does not cause any component or system to operate outside the performance boundaries described in the FSAR or Technical Specifications.

III. No. The temperature switch, the CPRCS and the Condensate System are not safety-related. Furthermore, the Condensate Pump Room Temperature High Setpoint is not governed by Technical Specifications.

/ERIN.: 91-192

DCP 90-3029 Removal of Valve Body Drain line for Reactor Recirculation Pump Suction Valve to eliminate the potential for drain valve leakage and eliminate the possibility of fatigue failure at the drain nozzle connection for Unit 1.

~f~gY:

I. No. The applicable design basis accident in the FSAR which was reviewed for potential impact by this change is described in FSAR Section 15.6, Decrease in Reactor Coolant Inventory". The specific evaluation is described in FSAR Section 15.6.5, "Loss-of-Coolant Accidents (Resulting from spectrum of postulated Piping Breaks within the Reactor Coolant Pressure Boundary) - Inside Containment". This change will not increase the probability of occurrence of an accident since all standards of design, material and construction, namely the ASME Boiler and Pressure Vessel Code,Section III, Class I, will be maintained for the modified piping.

II.

~ No.~ The change does not introduce any new modes of failure for the affected system, nor does it impact the system in such a manner that the probability of any type of accident would be increased.

III. No. The Tech. Spec. limit which is applicable, to this system with respect to this change is described in Tech. Spec. Section 3/4.4.8 "Structural Integrity".

The surveillance requirement for verifying the structural adequacy of the reactor coolant system as described in FSAR Section 5.2.4 and Tech. Spec. 4.0.5 is the periodic inspection and pressure testing of the pressure retaining components of the reactor coolant pressure boundary, in accordance with the requirements of Section XI of the ASME Boiler and Pressure Vessel Code and 10CFR50. This change does not affect the bases for the Tech. Spec. limits or programmatic requirements identified.

~ER~N).: 91-193 R REFEREN E NL 91-027, Rev. 0 DE RIPTI F PAN E:

Test to attempt to unplug the Liquid Radwaste Collection Tank (LRWCT) eductors to unblock them.

fQMil ARY:

I. No. FSAR section 11.2 was reviewed against the proposed test. This section delineates all safety mechanisms available to the liquid radwaste system. It states the collection system has no nuclear safety related function as a design basis.

II. No. FSAR section 15.7.3, Postulated Radioactive Release Due to Liquid Radwaste Tank Failure, provides a basis for the release of inventory of high activity Concentrates Waste Tank, ultimately causing airborne release through the plant vent. Failure of any equipment would have no effect on a safety related system.

A spill is bounded by the Concentrates Tank scenario.

III.

~ No.~ Technical Specification 3.11.1.3 provides basis and LCOs for the LRW system.

The operation of collection tank recirculation flow paths is not altered by this test.

The only change to the system is to remove a flow orifice in an attempt to unblock tank eductors.

@ERIN.: 91-194 R RE  : NL-91-011, Rev. 3 Removal of all irradiated reactor vessel components (excluding spent fuel) from the Unit 1 and Unit 2 spent fuel pools.

I. No. Accident analyses contained in the FSAR for a Fuel Handling Accident (FSAR 15.7.4) address accidents far more severe than similar accidents postulated for this project. Section 15.7.5 in the FSAR addresses a Spent Fuel Cask Drop Accident and concludes that this event is not credible due to the use of redundant rigging and a single failure proof crane. The Fuel Pool Cleanout Project activities are therefore bounded by the FSAR analyses noted.

II. No. Irradiated hardware will be loaded and transported in casks similar to spent fuel casks. Cask drop accidents are not considered a credible accident per FSAR Section 15.7.5 due to redundant rigging and a single failure proof crane, preventing a single failure.

Loss of water shielding in the cask storage pit is a credible scenario for evaluation. The airborne radiological consequences of this event would be bounded by the Fuel Handling Accident analyzed in FSAR Section 15.7.4.

IH. No. All Fuel Pool Cleanout Project activities will be conducted in accordance with plant Technical Specifications. The following Technical Specifications were reviewed: 3/4.3.2, 3/4.3.7, 3/4.9.6, 3/4.9.7, 3/4.9.9, 3/4.11.3, 3/4.11.4, 5.6.2, 6.9.1.8 and 6.13.

DCP 91-3024, Rev. 0 Provide connection from the Training Center addition to Domestic Water, Sanitary Drainage and Fire Protection.

Ilggf44Z I. No. Section 9.2.11 of the FSAR deals specifically with Potable Water and Sanitary Waste Systems, and section 9.5.11 deals with Fire Protection System. No functional changes to any systems will be made by this modification. The existing systems will be capable of providing the required services to the new facility with no adverse affects to other connected site facilities.

II. No. The connections are all performed on non-safety related systems located outside the plant security fence. No feature of the design of the proposed modification or locations for installation create any mechanism that would create the possibility of an accident or malfunction of a different type than previously analyzed in the FSAR.

I The Technical Specifications have been reviewed and the Domestic Water and the Sanitary Drainage are not specifically addressed. Section 3/4.7.6 provides the basis for the Technical Specification on the Fire Suppression System. This modification is only to tie into and use existing systems to service the new facility. No functional changes to any systems will be made by this modification.

The existing systems willbe capable of providing the required services to the new facility with no adverse affects to other connected site facilities.

DCP 90-3034G, Rev. 1 DE RI E:

Replace Condensate Demineralizer Vessel Underdrain System on Unit 2.

R:

I. No. The accident analyses addressed in Chapter 15 are not affected by the proposed action. FSAR Section 9.2.10 provides no discussion on the condensate demineralizer system, and therefore does not need to be addressed in this section.

This design basis of the Condensate Cleanup System (as described in Section 10.4.6.1) is to maintain the condensate at the required purity by removing various containments. This section also states the not-to-exceed effluent quality for the demineralizers. The design basis of the system is not affected by the proposed action. Section Section 10.4.6.3 identifies the presence of an effluent strainer at the outlet of each vessel to prevent a massive discharge of resins in the event of an underdrain failure. The strainers are not being affected by the proposed action and therefore will still be available to mitigate this type of malfunction.

II.

~ No.~ A malfunction not discussed in FSAR Chapter 15 or section 10.4.6 is where ionically exhausted resin, also containing a large amount of crud, remains in the vessel after resin transfer, causing poor quality feedwater. This type of malfunction occurs occasionally and is caused, in part, by poor flow distribution through the underdrain laterals. The proposed action will improve flow distribution through the underdrains, thus decreasing the likelihood of occurrence of this type of malfunction. An additional malfunction not previously evaluated in the FSAR is a sight glass failure which could cause personnel injury and/or a radioactive spill incident. Installing the site glasses creates a possibility of this type of malfunction; however, the possibility is extremely remote since the sight glass chosen has a design pressure of 750 psig.

III. No. Tech Spec Basis 3/4.4.4 states that water chemistry limits are established to prevent damage to the reactor materials in contact with reactor coolant. This basis identifies that conductivity measurements are required continuously since changes in this parameter are an indication of an abnormal condition. The conductivity monitoring equipment will not be affected by the proposed action.

JERE.: 91-197 RE  : DCP J91-1050, Rev. 0 Establish the low level alarm setpoint for the liquid radwaste collection tanks, samples tanks and surge tanks at approximately 2% tank level to reduce the potential for spurious alarms during normal operation and to provide consistency in the setpoints of the various tanks.

I. No. FSAR Section 15.7.3 provides an analysis of a liquid radwaste tank failure to show compliance with off-site dose criteria The tank containing the largest quantities of significant radionuclides in the liquid radwaste system is assumed to fail. No credit is taken for operator action or automatic actions. This setpoint change does not affect any assumptions used in this accident analysis.

II. No. This proposed change to the low level alarm setpoint for the subject radwaste tanks affects a non safety-related monitoring function only. The possibility for an accident or malfunction of a different type than any previously evaluated is not created by this change.

The only Technical Specification considered applicable to this change is TS 3/4.11.1.3, which requires the liquid radwaste treatment system to be operable in order to reduce the amount of radioactive materials in liquid waste prior to discharge. This proposed change will improve operation of the system by eliminating nuisance alarms.

DCP 90-3094, Rev. 0 Installation of flanges on each side of Contromatics check valves 141F024A, B, C and D inside containment to facilitate maintenance of the check valves in Unit 1.

I. No. This change will not increase the probability of occurrence of an accident since aH standards of design, material and construction will be met for the equipment being modified. Furthermore, this modification does not change overall performance of any system. This change will not increase the probability of a malfunction of equipment important to safety since no new failure modes will be introduced by this modification. This modification only installs flanges in existing piping in order to facilitate the maintenance of safety related check valves. The piping including the new flanges will remain in conformance with the loading combinations specified in FSAR Table 3.9-17 as originally designed.

~ ~ ~

H. No.

~ ~ This modification does not alter the function, change the operation or introduce

~

new failure modes for the Containment Instrument Gas System, the MSIVs or any other safety related components.

III. No. There are no requirements for the Containment Instrument Gas System in the Technical Specifications. This modification does not alter the basis for the technical specification for the MSIVs (3/4.4.7) or any other safety related systems.

$ gR gg7.: 91-199 R REPE  : DCP 90-3035E, Rev. 0 Addition of a RWCU pump handling system to remove the motor subassembly from the mounting location and transporting it to the maintenance shop for work as required.

$ Q@g4RY:

I. No. The design bases for the Reactor Building structure were reviewed in FSAR section 3.8.4.2 and SSES Safety Evaluation Report (NUREG 07761 section 3.8.2.

NUREG 0776 Supplements 2 and 6, section 9.4.1 discuss Heavy Load reviews and approves PP&L program for heavy load reviews. The load drop analysis concluded that the consequences of the drop do not result in the loss of the safety function of Reactor Building.

H. No. The RWCU handling system allows for removal of one RWCU pump while the RWCU system is pressurized. The dropping of the motor subassembly does not introduce any new failures, or failure modes, nor does it effect the safety function of Reactor Building structure.

IIL No. Technical Specification sections 3.3.2, 4.3.2.1, 4.3.2.2 and 4.3.2.3 are applicable to Reactor Water Cleanup system. Table 3.3.2-2 lists the isolation actuation instrumentation trip setpoints for the system. These sections and their implied basis were examined and it was determined that the margin of safety as defined in this basis is not affected by this modification.

JERKIN>.: 91-200 R F  : DCP 90-3035A, Rev. 0 Addition of a new purge water system and a local control panel to house the instrumentation and controls required to support operation of the new sealless RWCU replacement pumps.

I. No. The design basis accidents in the FSAR Chapter'15 and Reload Nuclear Analysis were reviewed and it was found that none of the design basis accidents are affected by these modifications. FSAR Sections 5.4.8.1, 5.4.8.2 and 5.4.8.3 provide the design bases, system description and system evaluation for the Reactor Water Cleanup system. FSAR Table 5.4-2 provides Reactor Water Cleanup System Equipment Design Data. FSAR Section 9.2.10 provides design basis, system description and safety evaluation for Condensate Storage and TransferSystem. Thesesectionsandtablewerereviewed forapplicability. The proposed modifications to support the replacement pumps do not alter the function of the Reactor Water Cleanup System.

II.

~ No. ~ The failure modes associated with the RWCU pump and the new supporting systems are not initiating events for any design basis accident and do not affect any safety systems. These changes do not modify the plant response to a design basis accident. These changes have no effect on safety functions, new purge water piping meets the high energy pipe break criteria provided in the FSAR Section 3.6.2.1.3 and the effects of the break from the new moderate energy line is bounded by the breaks in other larger lines.

III. No. This safety evaluation addresses only impacts of the potential leakage from the RWCU system on the new purge water skid and the Condensate Transfer System.

There are no sections of the Technical Specification applicable to the Condensate Storage and Transfer System, relating to flow capacity of the system.

R REFERE E: NL-91-011, Rev. 3 DE RIPTI N F HA E:

Fuel Pool Cleanout Project to remove all irradiated reactor vessel components (excluding spent fuel) from the Unit 1 and Unit 2 spent fuel pools.

SIIMMARY:

I. No. Accident analyses contained in the FSAR for a Fuel Handling Accident (FSAR 15.7.4) address accidents far more severe than similar accidents postulated for this project. Section 15.7.5 in the FSAR addresses a Spent Fuel Cask Drop Accident and concludes that this event is not credible due to the use of redundant rigging and a single failure proof crane. The Fuel Pool Cleanout Project activities are therefore bounded by the FSAR analyses noted.

II. No. Two types of accidents are postulated: a loaded cask drop scenario and draining of the cask storage pit with a significant inventory of irradiated hardware becoming uncovered. Cask drop accidents are not considered a credible accident per FSAR Section 15.7.5 due to redundant rigging and a single failure proof crane, preventing a single failure.

Loss of water shielding in the cask storage pit is a credible scenario for evaluation. The airborne radiological consequences of this event would be bounded by the Fuel Handling Accident analyzed in FSAR Section 15.7.4.

III. No. All Fuel Pool Cleanout Project activities will be conducted in accordance with plant Technical Specifications.

$ ER~.: 91-202

DCP 90-9114, Rev. 0 E RTPTI E:

Installation of a sampling device to be utilized for protecting the presence of biofouling mollusks in the river water make-up water supply system.

I. No. The proposed action affects only the RWMU system by adding a sample tank and associated 1/2" piping. The RYE system is discussed in the FSAR as confirmed by a review of FSAR Sections 9.2, 9.5 and 10.4, however, no safety functions are related to the RWMU system. In addition, no safety related equipment relies directly upon the RWMU system to perform its safety function.

There are Tech. Spec. limits on spray pond level to insure the spray pond inventory is adequate in the case of a design basis accident; however, continuous makeup is not required to assure adequate spray pond level, and a continuous removal of 1 gpm won't affect the make-up capability of the RWMU system.

II. No. There are no plant safety systems that rely directly upon the RWMU system, and the reliability/availability of the RWMU system is not affected.

III. No. The RWMU system is not mentioned in the basis for any Technical Specification.

The basis for the service water systems (ESW & RHRSW), which use river water as makeup (Section B3/4.7.1), are not affected by the proposed action, since failure of the proposed modification will not affect the capability of the RWMU system.

$ ER~,: 91-203 R EFERE E: DCP 90-3089, Rev. 0 DE RIPTI F HA E:

Replacement of existing Pacific wedge gate valves on the feedwater heater and drain cooler startup vent lines and the same valves on the bypass drain lines for the extraction steam to the 5A/B/C feedwater heaters on Unit I to prevent thermal binding.

KKK!lLY:

I. No. FSAR section 15.6.4 addresses steam system piping breaks outside containment.

The bounding fault event for breaks outside containment is the complete severance of one of the four main steam lines. The calculated exposure for this accident is illustrated in FSAR Table 15.6-9 and represents only a small fraction of the 10CFR100 guidelines. NUREG 0776 Section 15.3.3 concludes that the design provides adequate protection against such an accident. This accident scenario envelops the same type failure for the pipe being modified. The replacement valves are being purchased, installed, and tested in accordance with the same design codes as the original valves. The valves, and associated piping where applicable, are being welded into the existing system using approved SSES procedures in order to minimize the chance of leakage. The valves are located in the turbine building and no safety related equipment is installed below them.

II. No. The replacement valves are being purchased, installed and tested in accordance with ANSI B31.1, Power Piping requirements as required by FSAR section 3.2.

The replacement valves, and piping where applicable, will be welded into the existing system as are the original valves and piping in order to minimize the risk of leakage in accordance with FSAR section 10.4.10.3.

III. No. This modification is non-safety related and does not affect systems having Technical Specification requirements. This modification adds valves which are to be designed, built, and installed to the same standards as the existing valves.

The new valves and piping do not change the function or operation of the system in which they are to be installed.

0

/ERIN.: 91-204 R REFEREN E: DCP 90-3004, Rev. 1 DE RIPTI F HAN E:

Replacement of the Common Recombiner Vessel, the catalyst and the Common Recombiner Condenser and upgrade the heat trace system for the Common Recombiner Vessel.

ZQNNARY:

I. No. Section 15.7.1 of the FSAR evaluates the consequences of an Offgas System boundary rupture resulting from a seismic event, which is the most probable and only conceivable event which could cause significant damage to the system. The replacement recombiner vessel and recombiner condenser will be designed in accordance with the existing design requirements of the respective components with the exception that recombiner condenser will be designed to ASME Section VIIIinstead of ASME Section III, and will be more resistant to stress corrosion cracking. This modification will not adversely affect the ability of the Gaseous Radwaste System to reduce the release of radioactive materials in gaseous effluents to limits below the concentrations specified in 1OCFR Part 20.

II.

~ No.~ Section 15.7.1 of the FSAR evaluates the consequences of a breach of the Offgas System as a result of several different initiating events, the most probable of which is a seismic event. The improvement in the replacement recombiner vessel material, heat trace system and the recombiner condenser material to reduce their susceptibility to cracking reduces the probability of failure from these events.

III. No. The modification performed will not affect any of the system performance requirements concerning radioactive effluents and explosive mixtures discussed in the bases for Technical Specification Section 3/4.11.2. The replacement recombiner vessel and recombiner condenser will perform in exactly the same manner as the current components in meeting the Technical Specification requirements.

The change to upgraded materials and the upgrade of the heat trace system for the recombiner vessel will reduce the susceptibility to corrosion cracking which will increase the margin of safety over the current designs. The addition of flanges to the Gaseous Radwaste Recombiner Closed Cooling Water connections on the condenser, the replacement of the recombiner vessel drain line and the addition of the isolation valves to the condensate cooler outlet lines will not affect the operation of the system in meeting the Technical Specification requirements.

~ER~N).: 91-205 R REFEREN E: NL 91-028, Rev. 0 E RIPTI F HA E:

Reorganization of the Nuclear Department following implementation of the department Operational Effectiveness Review (OER).

MMARY:

I. No. Chapter 12.1, 12.5, 13.1, 13.4 and 17.2 of the FSAR were reviewed. The effects of organizational changes were reviewed from a safety standpoint. No line or support functions have been deleted. There is no overall reduction in staff. Clear lines of authority and responsibility have been maintained. Nuclear Quality Assurance, Nuclear Safety Assessment, Nuclear Training and Health Physics continue to enjoy organizational freedom to be independent of operational pressures.

II. No. The change from a focus on a discipline oriented organization to a key functional focus will allow local managers the ability to solve their own problems and minimize the number of interfaces necessary to accomplish normal functions.

III. No. The new organization meets all the aspects of organizational structures as stated in Tech Spec 6.2 Organization. All aspects of the organization charts which are important to safety have remained in the Tech Specs.

~ER~N.: 91-206 R REFEREN E: NL-91-011, Rev. 4 DE RIPTI N F HAN E:

Removal of all irradiated reactor vessel components (excluding spent fuel) from the Unit 1 and Unit 2 spent fuel pools.

@@@MARY:

I. No. Accident analyses contained in the FSAR for a Fuel Handling Accident (FSAR 15.7.4) address accidents far more severe than similar accidents postulated for this project. Section 15.7.5 in the FSAR addresses a Spent Fuel Cask Drop Accident and concludes that this event is not credible due to the use of redundant rigging and a single failure proof crane. The Fuel Pool Cleanout Project activities are therefore bounded by the FSAR analyses noted.

II. No. Irradiated hardware will be loaded and transported in casks similar to spent fuel casks. Cask drop accidents are not considered a credible accident per FSAR Section 15.7.5 due to redundant rigging and a single failure proof crane, preventing a single failure.

Loss of water shielding in the cask storage pit is a credible scenario for evaluation. The airborne radiological consequences of this event would be bounded by the Fuel Handling Accident analyzed in FSAR Section 15.7.4.

IH. No. All Fuel Pool Cleanout Project activities will be conducted in accordance with plant Technical Specifications. The following Technical Specifications were reviewed: 3/4.3.2, 3/4.3.7, 3/4.9.6, 3/4.9.7, 3/4.9.9, 3/4.11.3, 3/4.11.4, 5.6.2, 6.9.1.8 and 6.13.

JERKIN.: 91-207 R REFERE E: DCP 90-3035D, Rev. 0 E RlPTI F HA of existing RWCU pump E'eplacement 1P221B with a new sealless pump to reduce the need to repair and remove the pump from service.

BVMMAilY:

I. No. FSAR Sections 5.4.8.1, 5.4.8.2 and 5.4.8.3 provide the criteria, system description and system evaluation for the Reactor Water cleanup system. FSAR Tables 3.2-1 & 5.4-2 provide Reactor Water Cleanup System Equipment Design Basis and Data. FSAR Section 7.3.1.1A.2.4.1.9 addresses high flow isolation instrumentation.

These sections and tables and the design basis were reviewed for applicability.

The proposed modification to replace the existing pumps does not alter the function of the Reactor Water Cleanup System. The other function of the RWCU system affected by this modification is the leak detection capability. FSAR Section 7.3.1.1a.2.4.1.9 provides a description and the bases for the leakage detection system for the RWCU system. The design of the leakage detection features was not affected by this modification.

II. No. The replacement pump, piping and other associated components are in compliance with FSAR criteria and regulatory requirements and the leakage detection capability is not affected. A possible malfunction of a different type could be created by a change that adds a new single failure. These changes do not add any new single failure.

III. No. Technical Specification sections 3.3.2, 4.3.2.1, 4.3.2.2 and 4.3.2.3 are applicable to Reactor Water Cleanup system. Table 3.3.2-2 lists the isolation actuation instrumentation trip setpoints for the system. These sections and their implied basis were examined and it was determined that the margin of safety as defined in this basis is not affected by this modification.

~ER NO : .91-208 R REFERE E: DCP 90-3035C, Rev. 0 RI F A E:

Replacement of existing RWCU Pump 1P221A with a new sealless pump to reduce the need to repair and remove the pump from service.

sjlMNARY:

I. No. FSAR Sections 5.4.8.1, 5.4.8.2 and 5.4.8.3 provide the criteria, system description and system evaluation for the Reactor Water cleanup system. FSAR Tables 3.2-1 & 5.4-2 provide Reactor Water Cleanup System Equipment Design Basis and Data. FSAR Section 7.3.1.1A.2.4.1.9 addresses high flow isolation instrumentation.

These sections and tables and the design basis were reviewed for applicability.

The proposed modification to replace the existing pumps does not alter the function of the Reactor Water Cleanup System. The other function of the RWCU system affected by this modification is the leak detection capability. FSAR Section 7.3.1.1a.2.4.1.9 provides a description and the bases for the leakage detection system for the RWCU system. The design of the leakage detection features was not affected by this modification.

H. No.'he replacement pump, piping and other associated components are in compliance with FSAR criteria and regulatory requirements and the leakage detection capability is not affected. A possible malfunction of a different type could be created by a change that adds a new single failure. These changes do not add any new single failure.

IH. No. Technical specification sections 3.3.2, 4.3.2.1, 4.3.2.2 and 4.3.2.3 are applicable to Reactor Water Cleanup system. Table 3.3.2-2 lists the isolation actuation instrumentation trip setpoints for the system. These sections and their implied basis were examined and it was determined that the margin of safety as defined in this basis is not affected by this modification.

~ER~.: 91-209 R REFEREN: DCP 91-3014Z, Rev. 0 DE RI F A E:

Addition of multi-pin quick disconnect connectors (QDC) at the interface of the field cables and raceway system with the valve operator limit switch compartment of the Drywell MOVs for RBCCW, CIG, Main Steam, RWCU, Reactor Recirculation and RCIC to facilitate removal of the motor actuator.

$ QMM'BARF:

I. No. The addition of multi-pin and single-pin quick disconnect connectors on the motor actuator of valves inside the primary containment does not affect any of the postulated initiating events identified in Chapter 6 and 15 of the FSAR or NUREG-0776. Thus, the proposed action does not increase the probability of occurrence of an accident. The probability of an increase in the consequences of an accident is not affected by the proposed action. The safety function of all of the valves is not changed by the addition of the multi-pin and single-pin quick disconnect connectors.

II. No. The review did not identify a postulated initiating event which would create the possibility for an accident of a different type. Electrical separation of the Class 1E and affiliated power and control and the non-Class 1E control and instrumentation is maintained in accordance with PP&L Specification E-1012 and FSAR Section 8.3.1.11.4.

III. No. Tech Spec sections 3/4.6.3, 3/4.4.1 and 3.8.4.2.1 were reviewed. The proposed action associated with these valves does not affect the operability requirements, surveillance requirements or any existing margin of safety in any Technical Specification.

/ERIN.: 91-210 R REFERE E: DCP 90-3026, Rev. 2 DE RIPTI F HAN E:

Existing Unit 1 Reactor Building Sample Station (RBSS) (1C210) and Turbine Building Sampling Station (ASS) (1C132) are being replaced to update instrumentation and monitoring systems.

/@+MARY:

I. No. FSAR Sections 9.2.2.1, 9.3.1.1.1, 9.2.9.1, 9.4.4.1 and 9.3.2.3 were reviewed.

They state that all systems, except Reactor Building ventilation and RBCCW, have no safety-related functions. The reference sections were reviewed and it was determined that this change does not change the function of the systems or the operator's interface with these systems or any safety system. The review of FSAR Sections 6.2.3.2.3.1, 6.2.4.3.3.5 and 9.2.23 indicate that the RBCCW System containment penetrations and isolation valves perform containment isolation functions which are important to safety. The installation of new piping under this change does not change the probability of occurrence or the consequences of an accident or malfunction of these components over that which was previously evaluated.

IL No. Malfunction of the sample stations or the modulating damper does not have an adverse affect on the safety function of the reactor building recirculation mode.

Based on the review of FSAR Section 9.4.4 the Turbine Building HVAC system is a non-safety related system which is not required to operate to mitigate the consequences of an accident. The additional electrical load added by this change does not exceed the load margin provided in FSAR tables 8.3.1 through 8.3.5.

III. No. This modification does not reduce the margin of safety since the interfaces being installed for the TBSS and RBSS are not safety-related and do not affect any margins of safety described in the Technical Specifications.

/ERIN.: 91-211 R REFERE E: DCP 91-9012 F thru W, Rev. 0 D RlPTI F PAN E:

Replacement of Unit 2 Non-1E oversized 125V DC circuit breakers and DC rated replacement subfuses.

~jggi~RY:

I. No. The replacement breakers are identical replacements for original equipment except that the trip ratings match the derated ampacities of the connected cables in order to minimize to the extent practicable the effects of a sustained limited magnitude fault. The replacement fuses are identical replacements for original equipment and are sized to maintain coordination with the lower trip point breakers.

H. No. Installation of properly sized and dedication tested replacement breakers and properly sized fuses does not modify the non-safety related functions of the system loads. The proposed modification does not introduce any failure modes different than already analyzed in FSAR Table 8.3-21.

IH. No.

~ ~ The proposed action does not modify the function or capacity of any connected non-safety related systems and increases the probability of systems remaining functional in the event of a sustained limited magnitude fault above the cable ampacity by opening the affected circuit. This limits, to the extent practical, damage to adjacent cables of other non-safety systems in the same raceway.

Therefore the proposed action does not reduce the margin of safety as defined in the basis for any Technical Specification.

/ERIN.: 91-212 R REFEREN E: DCP 91-9004Z, Rev. 0 DE RIPTI N F HA E:

Replacement of Unit 1 Class 1E Oversized 125V DC circuit breakers.

~~jQ}/MARY:

I. No. The replacement breakers are dedicated replacements for original equipment with trip ratings which match the derated ampacities of the connected cables in order to minimize to the extent practicable the effects of a sustained limited magnitude fault. The failure mode and effects analysis for the replacement breakers is bounded by existing analysis provided in FSAR Table 8.3-21 for 1D614, 1D624, 1D634, & 1D644 branch circuit breakers. The functions of the connected load systems are not modified by this action.

II. No. Installation of properly sized and dedication tested replacement breakers and properly sized & qualified fuses does not modify the functions of the system loads. The proposed modification does not introduce any failure modes different than already analyzed in FSAR Table 8.3-21.

Based upon review of Section 8.0 Electric Power Systems of NUREG 0776 Safety Evaluation Report including Supplements 1 through 7 the proposed action does not affect any Technical Specification/Bases and does not reduce the margin of safety as defined in the basis for any Technical Specification.

$ ER~N.: 91-213 R REFEREN E: DCP 89-9153 A, B & C, Rev. 1 E RIPTI N F HA E:

Removal of existing seals and installation of new RHR pump "Hot" seals and seal piping. Also isolation of existing seal coolers from the RHR and ESW fluid systems. Removal of ESW piping seal coolers and associated hardware made obsolete by this change. Also, addition of flanges in ESW supply/return lines at branch connections to RHR pump seal and motor oil coolers.

KSfAilY:

I. No. The design has been reviewed against Section 5.4.7 (RHR), Section 6.2 (containment cooling) Section 6.3 (ECCS-RHR/LPC1) and Section 9.2.5 (ESW) to confirm that the design changes will not increase the possibility of an accident nor affect the subject systems or any other systems relying on their function/operation.

H. No. Installing the new "HOT" seal and deleting ESW cooling to the new seal will not affect the function of the RHR or ESW systems or any component or system relying on the subject systems. The new seal shall eliminate the RHR/ESW interface thus tending to simplify the operation of the systems and reduce the potential for operational error or radioactive releases to the environment.

III. No. Technical Bases 3/4.4.9 and 3/4.9.11 (RHR), 3/4.5.1 & 3/4.5.2 (ECCS-Operating and Shutdown), 3/4.7.1 Service Water Systems and 3/4.6.1 Primary Containment were reviewed. This change will not alter RHR pump functional capability for LPCI operation. Also, the elimination of ESW cooling water to the RHR pump seal cooler reduces the cooling capacity load on the system, thus increasing availability of cooling water for operation of the safety related equipment.

/ERIN.: 91-214 R REFEREN E: DCP 90-3014, Rev. 0 DE RIPTI F PAN E:

Installation of a wear strip on the lower valve guide of HPCI valve HV155F003 to prevent valve leakage which occurs when wear on the valve guide results in misalignment of the valve disc in the seat.

M RY:

I. No. Failure of the wear strip could result in the valve not fully closing. Single active component failures (SACF) such as failure of a valve to open or close are moderate frequency accidents as stated in FSAR Section 15A.3.3.2. Single failure of any single electrical device is similarly evaluated to have a moderate frequency. Failure of the valve to close whether it be from the wear strip binding the valve, a mechanical problem with the actuator, or an electrical supply failure to the operator each have the same probability of occurrence as defined in FSAR Section 15.0.3 and result in the same consequence.

H. No.

~ ~ In the worst case, the introduction of the wear strip into the HPCI steam supply line could lead to damage or failure of the HPCI turbine buckets, Single active component failures can render individual systems within the ECCS inoperable without effecting the reliability of the ECCS system as discussed in FSAR Section 6.3.2.5.

The inboard HPCI isolation valve HV-155FOO2 is to be closed, deenergized, and tagged to prevent operation during this modification in order to ensure that an Operation with Potential for Draining reactor Vessel/Cavity (OPDRV/C) does not exist. These measures will preclude a reactor cavity drain-down event and an accident of a different type will not be created.

III. No. This modification will increase the valve reliability in providing HPCI steam line isolation in the event of a steam supply line break or low reactor pressure. The addition of wear strips to valve HV155FOO3 does not affect the automatic isolation closure time of the valves as specified in Table 3.6.3-1 of the Technical Specifications. HPCI is not required to be operable during modes 4 or 5 when this modification will be performed. Therefore, the margin of safety as defined in the design basis for Section 3/4.6.3 Primary Containment Isolation Valves is not reduced.

/ERIN.: 91-215 R REFEREN E: NL-91-023, Rev. 1 E R1PTI N F HAM E Inspection of irradiated fuel rods to determine the condition of the fuel assemblies following irradiation.

$ QMil ARY:

I. No. FSAR Sections 9.1.4.2.7.1, 9.1.4.3.7, 9.1.4.5.1, 9.1.5 and 15.7.4 were reviewed. The fuel inspections are not covered in the FSAR. The review evaluated handling fuel, dropping equipment into the spent fuel pool and interaction with plant auxiliary systems which are similar to events previously evaluated in the FSAR.

II. No. Potential new events evaluated were, (1) radiation streaming on the refueling floor, (2) damage to the refueling floor curb, (3) damage to a fuel rod during inspection, (4) hot particles and, (5) a dropped fuel rod. The use of the inspection equipment or techniques does not create any hazard other than those previously evaluated.

Technical 'Specifications 3/4.9.6, 3/4.9.7 and 3/4.9.9 were reviewed. The proposed fuel inspections will not reduce any margin of safety because (1) the equipment is designed to perform fuel inspections, (2) a minimum of 7 feet of water shielding will be maintained above irradiated components, (3) the radiological consequences of fuel damage that could occur during the inspection are less than those previously evaluated in the FSAR, and 4) all fuel handling operations will be performed in accordance with RE-081-102, 'Fuel Movements Within Vault and Pool Locations, which requires all Technical, Specifications related to fuel handling to be complied with prior to and during all operations related to moving irradiated fuel in the spent fuel pool.

E'ER NO :.91-216 R REFERE E: DCP 90-3082. Rev. 0 DE RIPTI F HAN E:

Installation in Unit 2 of a SF6 gas injection system into the Circulating Water System used to identify condenser tube leaks in the main condenser water boxes. Also, installation of condenser hotwell HP, IP and LP sample lines to the Turbine Building Sample Station (TBSS) for water chemistry monitoring of condensate water.

SIIMNARY:

I. No. "

FSAR Sections 10.4.5, 9.3.2.3, 10.4.7 and 9.2.9.1 state that the mechanical, electrical and sampling systems have no safety-related functions. The referenced sections were reviewed and it was determined that this change does not affect the function of the systems or the operator's interface with these systems or any safety system.

The impact of the equipment installation in this change on the diesel generator has been evaluated and determined to have no effect on the safety functions of the diesel generators.

The process sampling system does not perform any safety functions. The sample lines installed and made operational by this change are only for chemistry monitoring and provide no direct control function. Loss of sample line flow due to blockage or breakage, results in loss of chemistry monitoring until corrected but does not create any possibility for unevaluated accidents or malfunction.

Breakage of lines of this size is addressed in FSAR Section 15.6.2 and results found acceptable. The SF6 gas can act as an asphyxiant in a high concentration.

Calculations determined that 20 lbs. is the maximum amount that can be stored and maintain a safe environment.

III. No. The first modification, SF6 injection, is tapped into the Circulating Water System piping which is not part of any safety-related system or pressure boundary and thus is not discussed in the basis for any Technical Specification. The second modification, the condenser hotwell HP, IP and LP samples interfaces are to the and the condensate piping. The TBSS does not have any affect on the 'BSS margin of safety described in the Technical Specification. The condensate piping which is the location of the sample nozzle taps is not discussed in the basis for any Technical Specification. The taps are located downstream of the Nuclear Boiler isolation valves. The Reactor Cooling System chemistry as noted in the Technical Specification section 3/4.4.4 will not be affected.

R REFEREN E: DCP 90-3081, Rev. 0 DE RIPTI F HAN E:

Installation in Unit 1 of a SF6 gas injection system into the Circulating Water System used to identify condenser tube leaks in the main condenser water boxes. Also, installation of condenser hotwell HP, IP and LP sample lines to the Turbine Building Sample Station (TBSS) for water chemistry monitoring of condensate water.

$ QMMARF:

I. No. FSAR Sections 10.4.5, 9.3.2.3, 10.4.7 and 9.2.9.1 state that the mechanical, electrical and sampling systems have no safety-related functions. The referenced sections were reviewed and it was determined that this change does not affect the function of the systems or the operator's interface with these systems or any safety system.

The impact of the equipment installation in this change on the diesel generator has been evaluated and determined to have no effect on the safety functions of the diesel generators.

H. No.

~ ~ The process sampling system does not perform any safety functions. The sample lines installed and made operational by this change are only for chemistry monitoring and provide no direct control function. Loss of sample line flow due to blockage or breakage, results in loss of chemistry monitoring until corrected but does not create any possibility for unevaluated accidents or malfunction.

Breakage of lines of this size is addressed in FSAR Section 15.6.2 and results found acceptable. The SF5 gas can act as an asphyxiant in a high concentration.

Calculations determined that 20 lbs. is the maximum amount that can be stored and maintain a safe environment.

IH. No. The first modification, SF6 injection, is tapped into the Circulating Water System piping which is not part of any safety-related system or pressure boundary and thus is not discussed in the basis for any Technical Specification. The second modification, the condenser hotwell HP, IP and LP samples interfaces are to the TBSS and the condensate piping. The TBSS does not have any affect on the margin of safety described in the Technical Specification. The condensate piping which is the location of the sample nozzle taps is not discussed in the basis for any Technical Specification. The taps are located downstream of the Nuclear Boiler isolation valves. The Reactor Cooling System chemistry as noted in the Technical Specification section 3/4-4.4 will not be affected.

R REFEREE E: DCP 90-3080, Rev. 0 E RIPTI F HAN E:

Installation of an Ion Chromatograph gC) Unit in the Unit 1 Turbine Building Sample Station (TBSS).

$ QigM4,RY:

I. No. The FSAR sections 9.3.2.3 and 9.3.3.3 contain the function and requirements of the interfacing and Process Sampling Systems. They state that the electrical and sampling systems have no safety-related functions. The referenced sections were reviewed and it was determined that this modification does not affect the function of the systems or the operator's interface with these systems or any safety system.

II. No. The process sampling system described in FSAR section 9.3.2.3 does not perform any safety functions. The sample lines installed and made operational by this change are only for chemistry monitoring and provide no direct control function.

Loss of sample line flow due to blockage or breakage results in loss of chemistry monitoring until corrected but does not create any possibility for unevaluated accidents or malfunction. Breakage of lines of this size is addressed in FSAR Section 15.6.2 and results found acceptable, III. No. The TBSS does not have any effect on the margin of safety described in the Technical Specification. The Reactor Cooling System chemistry as noted in the Technical Specification Section 3/4.4.4 will not be affected as a result of the modification performed by this change.

/ERIN.: 91-219 R REFEREN E: DCP 90-3042, Rev. 0 DE RIPTI N F HAN E:

Retirement of Unit 2 Reactor Building and Turbine Building Sample Stations (RBSS and TBSS) that were abandoned under a previous change for recovery of plant space'nd prevention of inadvertent operation of the equipment.

/~MARY:

I. No. None of the equipment or piping that is being removed from the plant by this change has an affect on equipment or systems that are important to safety. All of the equipment and piping that are being affected by this change were abandoned under previous changes.

II. No. This change does not create a possibility of a different type of accident or malfunction because the equipment is isolated and not functional. The removal of this equipment actually reduces the possibility for an accident or malfunction of any type because it eliminates unnecessary equipment and valves that could be inadvertently operated. The FSAR will be updated as a result of this change.

This update will delete the equipment that is being eliminated from the plant.

IH. No. RBCCW and the Reactor Building Zone IIventilation system are the only systems that are addressed in Technical Specifications. Technical Specification 3/4.6.3 requires that the RBCCW containment isolation valves shall be operable. This change will have no effect on the operability of these valves since the work is being completed on a small branch line that will be isolated from the system while the work is being completed. Technical Specification 3/4.6.5 requires that secondary containment integrity be maintained. A register is being installed in the ductwork that is being disconnected from the RBSS to assure that proper airflow balance is maintained. Technical Specification 3.4.11.2.5 discusses the operability of the ventilation exhaust gas treatment systems. Implementation of this change does not require that the Zone II equipment compartment exhaust system (Filtered exhaust) be taken out of service, therefore maintaining the system Technical Specification integrity.

/ERIN.: 91-220 R REFERE E: DCP 90-3041, Rev. 0 DE RIPTI F PAN E:

Retirement of Unit 1 Reactor Building and Turbine Building Sample Stations (RBSS and TBSS) that were abandoned under a previous change for recovery of plant space and prevention of inadvertent operation of the equipment.

~AR Y:

I. No. None of the equipment or piping that is being removed from the plant by this change has an affect on equipment or systems that are important to safety. All of the equipment and piping that are being affected by this change were abandoned under previous changes.

II. No. This change does not create a possibility of a different type of accident or malfunction because the equipment is isolated and not functional. The removal of this equipment actually reduces the possibility for an accident or malfunction of any type because it eliminates unnecessary equipment and valves that could be inadvertently operated.

The FSAR will be updated as a result of this change. This update will delete the equipment that is being eliminated from the plant.

III. No. RBCCW and the Reactor Building Zone I ventilation system are the only systems that are addressed in Technical Specifications. Technical Specification 3/4.6.3 requires that the RBCCW containment isolation valves shall be operable. This change will have no effect on the operability of these valves since the work is being completed on a small branch line that will be isolated from the system while the work is being completed.

Technical Specification 3/4.6.5 requires that secondary containment integrity be maintained. A register is being installed in the ductwork that is being disconnected from the RBSS to assure that proper airflow balance is maintained.

Technical Specification 3.4.11.2.5 discusses the operability of the ventilation exhaust gas treatment systems. Implementation of this change does not require that the Zone I equipment compartment exhaust system (Filtered exhaust) be taken out of service, therefore maintaining the system Technical Specification integrity.

JERKIN.: 91-221 E: DCP 90-3108 E, J & K, Rev. 0 HAN E:

Modification to existing pipe support/restraint configuration that are required in order to reduce the total number of mechanical snubbers located on Diesel Generator 'E', Feedwater - Inside Containment, and RPV Instrumentation Piping Systems.

R:

I. No. The systems included in this proposed modification were reanalyzed to justify the removal of unnecessary snubbers and their pipe support configurations are to, be modified accordingly. The reanalysis is in accordance with the original design basis, and has been performed in accordance with accepted industry codes. All aspects of the existing qualification calculations have been addressed, all original interface design parameters such as equipment allowables have been considered and all system design requirements as specified in the Design Specifications and FSAR have been addressed in the reanalysis effort. As in the existing piping analysis, the applicable code design limits have been met to ensure piping integrity and system function; FSAR Section 15.0.3.1, 15.0.3.5.

II. No. The accident events/causes in FSAR Chapter 15.0 that are applicable to this proposed action involve pipe breaks. The proposed action only reduces the number of seismic and hydrodynamic restraints (snubbers) on selected piping systems by reanalyzing each line using optimum restraint configurations. The original design basis is being followed. The original allowable equipment interface loads are still being met and the pipe break criteria remains unchanged, IH. No. Technical Specifications 3/4.8.1, 3/4.3.1 thru 3.5, 3/4.5.1, 3/4.4.8, 3/4.6.1, 3/4.6.3 and 3/4.7.4 were reviewed. The proposed action reduces the overall snubber population for SSES Unit 1. In cases of snubber size change the same type of snubber as currently installed, Pacific Scientific mechanical snubbers, will be used where necessary - i.e., no new type of snubber will be introduced by these modifications. The bases, as defined in the Technical Specifications, deal with snubber test requirements and are not directly affected by this proposed action.

/ERIN).: 91-222

DCP 91-9068, A thru X, Rev. 0 DE RlPTI HAN E:

Replacement of Unit 1'on-1E oversized 125V DC circuit breakers.

MMARF:

I. No. The replacement breakers are identical replacements for original equipment except that the trip ratings match the derated ampacities of the connected cables in order to minimize to the extent practicable the effects of a sustained limited magnitude fault. The replacement fuses meet or exceed the performance characteristics of the original equipment and are sized to maintain coordination with the lower trip point breakers. The non-safety related functions of the connected load systems are not modified by this action.

II. No. Installation of properly sized replacement breakers and properly sized fuses does not modify the non-safety related functions of the system loads. The proposed modification does not introduce any failure modes different than already analyzed in FSAR Table 8.3-21.

The proposed action does not modify the function or capacity of any connected non-safety related systems and increases the probability of systems remaining functional in the event of a sustained limited magnitude fault above the cable ampacity by opening the affected circuit. This limits, to the extent practical, damage to adjacent cables of other non-safety systems in the same raceway.

~ER~N.: 91-223 R REFERE E. NL-91-031, Rev. 0 DE RlPTI F EM Procedure change to OP-139-001 to allow flow through Unit 1 condensate demineralizers in excess of 5760 gpm up to 6200 gpm until flows can be restored to normal values because of dirty resin traps on the outlet of other demin vessels.

RYYMMARY:

I. No. Section 15.6.6 of the FSAR describes a feedwater line break outside containment.

Failure of a condensate demineralizer pressure boundary would produce an accident which is similar to this analyzed event. However, the increased flow does not increase the probability of failure of the pressure boundary. All components which experience increased flow (the demin vessel, the piping, the flow instrumentation, and the resin trap) can accommodate the increased flow without exceeding their design limits for structural integrity. The consequences are also not increased. The effect of a loss of feedwater is not significantly changed by changes to demin flows.

II.

~ No. ~ Chapter 15 of the FSAR was reviewed. No accident willbe created which is not already addressed. No failures are expected which could cause an accident.

HI. No. The basis section of the Technical Specifications was reviewed. The margin of safety is not reduced as defined in the basis. Condensate demineralizer fiows are not mentioned in the basis.

R REFERE E: NL-91-032, Rev. 0 DE RIPTI F HAN E:

Installation of pressure monitoring instrumentation and a strip chart recorder for the Radwaste Solidification Sample System (RWSS) to monitor system pressures during normal system operation.

JtlMh/ARF:

I. No. Upon review of FSAR section 15.7.3 Postulated Radioactive Releases Due To Liquid Radwaste Tank Failure it has been determined that this procedure has no impact on the postulated events of cracks in the tanks or operator error. System operation does not change and the probability of a tank crack is not influenced by implementation of this procedure.

H. No. The only difference in operation of the system will be the installation of a pressure transmitter and the monitoring of the pump logic. In the event of a pressure transmitter failure the result would be minimal leakage of sludge onto floor. For the monitoring of the pump logic the strip chart recorder channel could either cause an open circuit or a short circuit. If an open circuit were to occur only the channel of the strip chart recorder monitoring would be affected.

Should a short circuit develop the result would be a blown fuse in the RWSS Sample Pump logic which would terminate the transfer.

III. No. Monitoring of system pressure and a contact point's voltage will not affect any margin of safety of the system. With regards to Tech Specs, this system does not have affect on any other system other than itself and this system is not Tech Spec related.

R RE E: DCP 91-3019, Rev. 0 DE R?PTI F HA E:

Modification of RCIC Steam Admission Valve and control circuit to resolve startup overspeed concerns.

R:

I. No. The modification does not increase the probability of occurrence of an accident as previously evaluated in the FSAR, because a total failure of the steam admission valve or its control circuit would not result in an accident as previously evaluated in the FSAR. The modification does not increase the consequences of an accident as previously evaluated in the FSAR. The FSAR Chapter 15 accident analyses were reviewed for impact but none was found. RCIC may be used to help maintain reactor water level in case of a LOCA, however, per FSAR Section 7.4.2.1.2.1.8 RCIC meets the single-failure criterion on a network basis with HPCI. It is not necessary for RCIC alone to meet the single failure criterion since its function is duplicated or backed up by other systems. Regardless, this modification only improves RCIC system reliability.

The new electrical components (auxiliary relays and time delay relay) being installed by this modification are highly reliable. However, should one of these components fail, the only operating scenario that might occur which is not similar to what is seen today is that the steam admission valve might remain stuck at about 10% open leaving the turbine turning at idle speed. This low probability event is enveloped by failure of the RCIC System as discussed FSAR Section 5.4.6.2.5.4. Similarly, this modification does not create the possibility of a malfunction of a different type than any evaluated previously in the FSAR. As stated above the newly installed electrical components and cage plug only increase the reliability of the RCIC System and have no negative impact on any other plant systems.

III. No. RCIC System operability is addressed in the Technical Specifications in the Plant Systems Section 3/4.7.3 and Bases Section 3/4.7.3. This modification has no impact on RCIC operability as is described in these sections. In addition RCIC System actuation instrumentation is addresses in the Technical Specifications in Instrumentation Section 3/4.3.5 and Bases Section 3/4.3.5. This modification does not involve or impact in any way this instrumentation.

/ERIN.: 91-226 R REFE E: DCP 91-9079, Rev. 0 DE RTPTI F HAN E:

Replace the first butt-welded elbow on the 4" recirculation line directly off each LRW collection tank with flanged elbows to allow easy access to the piping close to and inside each tank for cleaning.

~i/MARY:

I. No. This modification does not alter the design basis or operation as described within FSAR section 11.2.1, "Liquid Waste Management Systems". Replacing the butt-welded elbow fitting with a flanged elbow on each of the LRW collection tanks creates no change to system operation or function. The possibility of a pipe rupture has been evaluated against the Accident Analyses in Chapter 15 of the FSAR, specifically section 15.7.2 and 15.7.3, Since the components affected by this change will comply with the original ANSI piping code there is no decrease to the design integrity of the system.

H. No.

~ ~ Implementation of this modification will not change the normal operation of the Liquid Radwaste Management Systems, specifically the liquid radwaste processing subsystem. The system will continue to operate in accordance with its original design bases, and the piping modification to the elbows complies with the original construction codes (ANSI B31.1).

III. No. The flanges installed per this modification will enhance the Liquid Waste Management System by providing easy access to recirc. piping but in no way affects system operability and/or radioactive material release rates as described in Technical Specification 3/4.11.1.

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