ACCELERATED DISTRIBUTION TIONtrip solenoid valves, 20-1/OPC and 20-2/OPC, and associated circuits....

ACCELERATED DISTRIBUTION DEMONS TION SYSTEM

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9309280400 DOC.DATE: NOTARIZED: NOFACIL:50-389 St. Lucie Plant, Unit 2, Florida Power & Light Co.

AUTH.NAME AUTHOR AFFILIATIONSAGER,D.A. Florida Power 6 Light Co.

RECIP.NAME RECIPIENT AFFILIATION

DOCKET05000389

SUBJECT: "St Lucie Unit 2 Rept of Changes Made to Facility U rProvisions of 10CFR50.59 for Period 911'007-930325. W/930922ltr.

DISTRIBUTION CODE: IE47D COPIES RECEIVED:LTR L ENCL / SIZE: 7TITLE: 50.59 Annual Report of Changes, Tests or Experiments Made W/out

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NOTES:

RECIPIENTID CODE/NAME

PD2-2 PD

COPIESLTTR ENCL

1 0

RECIPIENTID CODE/NAME

NORRIS,J

COPIESLTTR ENCL

2 2

/A

DINTERNAL: ACRS

AEOD/DSP/ROABNRR/DRCH/HHFBRGN2 FILE 01

EXTERNAL: NRC PDR

6 61 '

1 11 1

1 1

AEOD/DOA~AEOD DSP/TPABREG - E 02

NSIC

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NOTE TO ALL"RIDS" RECIPIENTS

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PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK,ROOM P 1-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTIONLISTS FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 17 ENCL 16

P.O. Box 128, Ft. Pierce, FL 34954-0128

APL September 22 1993

L'-93-24610 CFR 50.59

U. S. Nuclear Regulatory CommissionAttn: Document Control DeskWashington, D. C. 20555

Gentlemen:

Re: St. Lucie Unit 2Docket No. 50-389Re ort of 10 CFR 50.59 Plant Chan es

Pursuant to 10 CFR 50.59 ('b)(2), the enclosed report contains abrief description and summary of the safety evaluation of PlantChanges/Modifications (PCMs) which were made, and are reportablepursuant to 10 CFR 50.59. Included with the brief description ofeach PCM is a summary of the safety evaluation completed byFlorida Power & Light (FPL) for that PCM. This report includesPCMs completed for the period of October 7, 1991 to March 25,1993, and correlates with the information included in Amendment 8of the Updated Final Safety Analysis Report submitted underseparate cover.

Should there be any questions on this information, please contactus ~

Very truly yours,

D. A.ViceSt. Lu

geresidentie Plant

DAS/CDW/kw

Enclosure

cc: Stewart D. Ebneter, Regional Administrator, Region II, USNRCSenior Resident Inspector, USNRC, St. Lucie Plant

DAS/PSL 75994-931

A

9309280400 930325PDR ADQCK 05000389 ~4R PDR

an FPL Group company

RE: St. Lucie PlantDocket No. 50-38910 CFR 50.59 Report

St. Lucie Unit 2Report of Changes Made to the Facility

Under the Provisions of10 CFR 50.59

for the period October 7, 1991to March 25, 1993

NOTE: The safety evaluations in this report are chronologically arranged startingwith those created more recently. Please note that the level of detail ofsafety evaluations from earlier years do not reflect current practices.

~ 93OgZBO)pp

Plant Change/Modifications reportable pursuant to10 CFR 50.59 for St. Lucie Unit 2

Number Su lement Title

'52-292146-292140-292

120-292

097-992067-292053-292

023-292

563-291

543-291

510-291500-291486-291421-291

419-291418-291309-291

247-291

092-291

091-291256-290176-290311-289

00-10

00-30-1

0-1000

0-100

0-1

01-200

Turbine Trip Controls ModificationFire Barrier Drawing EnhancementsCCW Heat Exchanger Temperature ControlValve Minimum Stop SetpointFixed Incore Detector Replacement andMovable Incore Detector Assembly ModificationSouth Craft Lunchroom FacilityFuel Reload for Cycle 7Main Generator Inadvertent EnergizationProtectionCondenser Air Evacuation System HeaderSeparationInstrument Inverter Trip and Alarm CircuitImprovementsDeenergizing of Waste Management Heat TraceCircuitsNRC Generic Letter 89-10 MOV Thrust ValuesSpent Fuel Pool Visual Level IndicatorRemoval of Turbine RunbackEmergency Diesel Generator Shaft CouplingGuardsAtmospheric Dump Valves Actuator ModificationContainment Spray Vent Valve InstallationContainment Hydrogen Analyzer SystemEnhancementsReplacement of RCP Upper & Lower OilReservoir Level Measurement SystemFisher & Porter Indicating ControllersReplacement125 VDC Arc Suppression AdditionFuel Reload for Cycle 6BoricAcid Concentration Reduction ModificationObsolete Smoke Detectors Replacement

PC/M 152-292 Supplement 0

ABSTRACT

This Engineering Package (EP) implements portions of the recommendations ofWestinghouse Customer Advisory Letter 92-02, 'Operation, Maintenance, Testing of, and

'ystem Enhancements to Turbine Overspeed Protection System', as addressed in FPLInteroffice Correspondence 'Westinghouse CAL 92-02 Overspeed Protection System'.The Customer Advisory Letter was issued in response to the destructive turbine-generatoroverspeed of Public Service Electric and Gas Company Salem Unit 2 on November 9,1991, as documented by INPO Significant Event Report 7-92.

Indicator lights will be installed to monitor the electrical continuity of the turbine auto stoptrip solenoid valve, 20/AST, emergency trip solenoid valve, 20/ET, overspeed protectiontrip solenoid valves, 20-1/OPC and 20-2/OPC, and associated circuits. This will provideindication of capability to accomplish manual or automatic turbine trip.

A trip test switch will be installed at the turbine front standard to allow on-line electricalactuation testing of the auto stop trip solenoid valve, 20/AST, without causing turbine trip.This will allow demonstration of the operability of the valve and of the capability toaccomplish manual or automatic turbine trip.

A seal-in circuit will be installed for the coil of the emergency trip solenoid valve, 20/ET.This willcause the emergency trip solenoid to remain energized following actuation by theturbine trip push button and will ensure that the turbine remains tripped.

Surge suppression diodes willbe installed across the coil of the emergency trip solenoid.

All existing manual and automatic trip functions will remain in effect.

The control systems affected by this EP perform no safety related function. However,since modifications will be made to RTGB-201, which has been seismically qualified, thisEP has been classified as Quality Related.

A walkdown of the turbine control system revealed a discrepancy between installedcircuits and controlled drawings. An auxiliary relay associated with turbine and feedwaterpump trip on high-high steam generator level was found to have been connected in amanner which did not ensure reliable trip operation. The relay wiring is modified, and theaffected drawings revised to reflect the corrected wiring configuration.

The safety evaluation of this EP has shown that the implementation of this PCM does notconstitute an unreviewed safety question as defined in 10CFR50.59 and does not requirea change in the Plant Technical Specifications. This PCM has no adverse impact on plantsafety or operation and may be implemented without prior NRC approval.


SAFETY EVALUATION

As defined in 10 CFR 50.59, an unreviewed safety question exists; (i) if the probability ofoccurrence or the consequences of an accident or malfunction of equipment important

. to safety previously evaluated in the Safety Analysis Report (SAR) may be increased; or(ii) if a possibility of an accident or malfunction of a different type than any previouslyevaluated in the SAR may be created; or (iii) if the margin of safety as defined in the basisfor any Technical Specification is reduced.

In accordance with 10 CFR 50.59, the following evaluation serves to determine whetherthis modification constitutes an unreviewed safety question:

Does the proposed change increase the probability of occurrence of an accidentpreviously evaluated in the SAR?

FSAR Section 15.2.1.2, Limiting Reactor Coolant System Pressure Event-Isolationof Turbine, evaluates plant response to an isolation of turbine at 102% power, andFSAR Section 15.2.2.1, Limiting Offsite Dose Event-Isolation of Turbine With aStuck Open Main Steam Safety Valve, evaluates response to isolation of turbineat 20% power.

Although the proposed changes result in a slight increase in the probability of aninitiating event, turbine trip due to circuit failure, the changes also providecompensating effects through the addition of turbine control system test andmonitoring functions which enhance the capability to assess the operability of aprotective system prior to the system being required to operate, and throughmodification of control operation to provide assurance that a manual turbine tripwillbe maintained. In accordance with 'Nuclear Engineering Department Guidancefor Performing-10CFR50.59 Safety Evaluations', Revision 0, Section 4.7, theproposed changes therefore do not increase the probability of occurrence of anaccident previously evaluated as defined in 10CFR50.59.

Does the proposed change increase the consequences of an accident previously evaluatedin the SAR?

The modifications of this PCM do not prevent any safety related equipment from performingits intended function. The consequences of a fault in the modified system will remainbounded by the existing FSAR analyses.

3. Does the proposed change increase the probability of an occurrence of a malfunction ofequipment important to safety previously evaluated in the SAR?


SAFETY EVALUATION(continued)

The modifications of this PCM do not prevent any safety related equipment from performingits intended function. The only potential interaction with Safety Related equipment is in theRTGB. Since the components are mounted in the RTGB in accordance with seismic designcriteria, the probability of occurrence of a malfunction of equipment important to safetypreviously evaluated in the SAR is not increased by this modification.

Does the proposed change increase the consequences of a malfunction of equipmentimportant to safety previously evaluated in the SAR?

No failure of the turbine control system alters the designed operation of any safety relatedequipment. Therefore, the consequences of malfunction of equipment important to safetypreviously evaluated in the SAR are not increased by this modification.

Does the proposed change create the possibility of an accident of a different type than anypreviously evaluated in the SAR?

This modification does not change the design bases of any structure, system, orcomponent important to safety. No new failure modes or conditions are created that canbe postulated to cause an accident different than those previously analyzed in the SAR.Therefore, the possibility of an accident of a different type than any previously evaluatedin the SAR is not created by this modification.

Does the proposed change create the possibility of a malfunction of equipment importantto safety of a different type than any previously evaluated in the SAR?

The manual and automatic turbine trip functions of the existing turbine control systemremain in effect. No failure modes are created that can be postulated to cause amalfunction of equipment important to safety different than those previously analyzed in theSAR. Therefore, the possibility of a malfunction of equipment important to safety which isof a different type than any previously evaluated in the SAR is not created by thismodification.

Does the Proposed change reduce the margin of safety as defined in the basis for anyTechnical Specification?

The Technical Specification requirements and Technical Specification Bases are notaffected by this modification. Therefore, the margin of safety as defined in the bases forany Technical Specification is not reduced by this modification.



The changes made by this EP are not to the level of detail required by the SAR, and provide forthe conduct of tests not described in the SAR. 10CFR50.59(a)(1) permits changes which providefor the conduct of tests not described in the SAR when such changes do not involve a changeto the Technical Specifications or an unreviewed safety question.

The foregoing discussion constitutes, per 10CFR50.59(b), the written safety evaluation whichprovides the bases that these modifications do not impact the safe operation of'the plant,constitute an unreviewed safety question, or require a change to the plant TechnicalSpecifications. As such, prior NRC approval for the implementation of this PC/M is not required.

PC/M 146-292 Supplement 0-1

ABSTRACT

This Engineering Package (EP) provides for the enhancement of drawings for the existing Unit 2Fire Barriers in response to QA audit QSL-OPS91-794. The audit was performed to investigate't. Lucie Plant engineering documentation and drawing updates concerning fire barriers andresulted in a finding which indicated that current plant configurations were not completely reflectedon the fire barrier drawings. Therefore, revisions to enhance these drawings and the creation ofadditional drawings is necessary.

The,'intent of this EP is to enhance the drawings of existing fire barriers and assign a FireArea/Fire Zone designation to one stairwell inside the Reactor Auxiliary Building (RAB). A SafeShutdown Analysis (SSA) has been performed to demonstrate that fire protection equipmentrequired to protect Safety Related equipment, or required to maintain the integrity of a fire barriernecessary to protect Safety Related equipment has not been adversely affected. Accordingly, thisEP has been classified as Quality Related. A safety evaluation has beeri performed in accordancewith 10 CFR 50.59 and is documented in Section 3.0 of this EP. This evaluation demonstratesthat implementation of this modification does not involve an unreviewed safety question and doesnot require a change to the Plant Technical Specifications. In addition, this modification has nodetrimental effects on plant safety or operation. Based upon the above, prior NRC approval isnot required for the implementation of this modification.

Su lement 1

Supplement 1 to this EP deletes references to the fire protection related Technical Specificationswhich were deleted by License Amendment 55 subsequent to the issuance of Supplement 0.

SAFETY EVALUATION

With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shallbe deemed to involve an unreviewed safety question: (i) if the probability of occurrence or theconsequences of an accident or malfunction of equipment important to safety previously evaluatedin the Safety Analysis Report (SAR) may be increased; or (ii) if the possibility for an accident ormalfunction of a different type than any evaluated previously in the Safety Analysis Report maybe created or (iii) if the margin of safety as defined in the basis for any technical specification isreduced.

In accordance with 10CFR50.59, the following serves to determine whether this modificationconstitutes an unreviewed safety question:

1. Does the proposed activity increase the probability of occurrence of an accident previouslyevaluated in the SAR?

QePC/M 146-292 Supplement 0-1


The proposed activity does not increase the probability of occurrence of an accidentbecause no physical changes are being made to the plant. No combustibles are addedby this EP.

Does the proposed activity increase the consequences of an accident previously evaluatedin the SAR?

"The proposed activity does not increase the consequences of an accident. No fire barriersare modified by this EP and no combustibles are added. The fire suppression capabilitiesare not affected.

Does the proposed activity increase the probability of occurrence of a malfunction ofequipment important to safety previously evaluated in the SAR?

The proposed activity does not increase the probability of occurrence of a malfunction ofequipment important to safety, because no equipment important to safety is affected by thisEP.

4. Does the proposed activity increase the consequences of a malfunction of equipmentimportant to safety previously evaluated in the SAR?

The consequences of a malfunction of equipment important to safety are not affectedbecause no equipment is affected. The EP merely changes the drawings to reflect thepresent plant configuration. The addition of the stairwells as fire zones will add newresponses to the SSA; however, all these cables already appear in the zones adjacent tothe stairwells. The drawings currently reflect the stairwells as part of the adjacent zones;therefore, the operator response would be the same.

5. Does the proposed activity create the possibility of an accident of a different type thanpreviously evaluated in the SAR?

The proposed activity does not create any possibility of an accident of a different typebecause no failure modes are created.

Does the proposed activity create the possibility of a different type of malfunction ofequipment important to safety than any previously evaluated in the SAR?

The proposed activity does not create the possibility of a different type of malfunction ofequipment important to safety. This EP does not change any equipment or the operatorresponse to a fire.

PC/M 146-292 Supplement 0-1II


7. Does the proposed activity reduce the margin of safety as defined in the basis for anytechnical specification?

The proposed activity does not reduce the margin of safety as defined in the basis for anyTechnical Specification, because the Technical Specification will not be affected by thischange.

The foregoing constitutes, per 10CFR 50.59(b), the written safety evaluation which provides thebases that this change does not involve an unreviewed safety question nor a change to the PlantTechnical Specifications and prior NRC approval for the implementation of this modification is notrequired.


ABSTRACT

This Engineering Package (EP) allows setting the minimum open stop limit on the pneumaticcontrollers of Temperature Control Valves TCV-14-4A and TCV-14-4B as low as 8 percent open.

. TCV-14-4A and TCV-14-4B regulate the ICW flow through CCW heat exchangers 2A and 2Brespectively by maintaining the CCW outlet temperature of the heat exchanger at its requiredsetpoint. The ICW system serves as a heat sink for the Component Cooling Water System, theTurbine Cooling Water System and the Steam Generator Blowdown System. The ICW Systempiping supplies cooling water to these three systems in parallel.

The purpose of this modification is to increase the intake cooling water flow to the Turbine CoolingWater (TCW) heat exchangers during normal plant operations. Increased intake cooling waterflowto the TCW heat exchangers is desired during peak summer periods due to increased foulingof these heat exchangers and high intake cooling water temperatures. Since the CCW heatexchangers are sized for Design Basis Accident (DBA) conditions they are significantly oversizedfor normal plant operating conditions and maximum expected peak summer periods. Therefore,during reduced heat loads on the CCW heat exchangers such as during normal plant operationsthe CCW outlet temperature setpoint is maintained by the controllers which throttle the valves.The excess ICW flowwill be diverted to the TCW heat exchangers. This modification only affectsthe valves'inimum open position during normal plant operations and does not prevent the valvefrom opening to the full open position to perform its Safety Related function.

An Engineering Evaluation was performed to address concerns of waterhammer on the Safety.Related portion of the ICW system with TCV-14-4A and TCV-14-4B set as low as 8 percent open,coincident with an ICW pump stop and restart during an Emergency Diesel Generator loadingsequence. This analysis demonstrates that this modification will not adversely affect thecomponents of the ICW system and is acceptable. In addition, precautionary measures are takenin this EP to preclude valve degradation from potential cavitation due to long term operation ata minimum open position as low as 8 percent.

This modification affects the waterhammer analysis of the Safety Related piping of the ICWsystem, therefore, this PCM is being classified as Nuclear Safety Related.

A safety evaluation of this modification has been performed in accordance with 10 CFR 50.59.This evaluation has shown that implementation of this Engineering Package does not have anadverse effect on plant safety, security or operation, does not constitute an unreviewed safetyquestion and does not require a change to Plant Technical Specifications. Therefore, prior NRCapproval for implementation of this modification is not required.


SAFETY EVALUATION

With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shallbe deemed to involve an unreviewed safety question: (i) if the probability of occurrence or theconsequences of an accident or malfunction of equipment important to safety previously evaluatedin the Safety Analysis Report may be increased, or (ii) if a possibility for an accident or malfunctionof a different type than any evaluated previously in the Safety Analysis Report may be created, or(iii) ifthe margin of safety as defined in the bases for any Technical Specification is reduced. Themodification included in this engineering package does not involve an unreviewed safety questionbecause of the following reasons:

Does the proposed activity increase the probability of occurrence of an accident previouslyevaluated in SAR?

2.

The proposed activity does not increase the probability of occurrence of an accidentpreviously evaluated because the Temperature Control Valves modified by this EP are notconsidered to initiate any acctdent.

e


I

The proposed activity does not increase the consequences of an accident since asdiscussed in section 1 there are no accidents previously analyzed which are attributed tothe Temperature Control valves.

3. Does the, proposed activity increase the probability of occurrence of a malfunction ofequipment important to safety previously evaluated in the SAR?

The proposed activity does not increase the probability of a malfunction of equipmentimportant to safety because neither the functionality of the Temperature Control Valves orthe CCW heat exchanger outlet temperature setpoint or the Safety Related signal thatgenerates isolation of the non essential from the essential ICW header are affected by thismodification. As demonstrated in the Failure Modes and Effects Analysis section thismodification does not degrade the reliability or increase challenges, directly or indirectly forequipment important to safety.

4. Does the proposed activity increase the consequences of a malfunction of equipmentimportant to safety previously evaluated in the SAR?

The proposed activity does not increase the consequences of a malfunction of equipmentimportant to safety because the ICW and CCW system remain capable of performing theirSafety Related function during a DBA. No new failure modes are introduced by thismodification.



5. Does the proposed activity create the possibility of an accident of a different type than anypreviously evaluated in the SAR?

This modification does not degrade the reliability or increase challenges, directly orindirectly for equipment important to safety nor does it add or delete any equipmentimportant to safety. The redundancy and separation of the ICW and CCW system are notaffected by this EP, thereby, a loss of ultimate heat sink event is not created. For thesereasons, the proposed activity does not create the possibility of an accident of a differenttype than any previously evaluated in the SAR.

6. Does the proposed activity create the possibility of a malfunction of equipment importantto safety of a different type than any previously evaluated in the SAR?

The proposed activity does not create the possibility of a malfunction of equipmentimportant to safety of a different type than any previously evaluated because as statedpreviously, no new failure modes are created by this modification and all failures analyzedin the FSAR for the ICW and CCW system remain unchanged. This modification does notdegrade the reliability or increase challenges, directly or indirectly for equipment importantto safety.

7. Does the proposed activity reduce the margin of safety as defined in the bases for anyTechnical Specification?

The proposed activity does not reduce the margin of safety as defined in the basis for anyTechnical Specification, because neither is the integrity or is the flow capability of the ICWand CCW system affected by this EP. In addition, this EP has no impact on the ICW orCCW system as bounded by existing Technical Specifications.

The foregoing discussions provided in Sections 2 and 3 of this EP constitutes, per 10 CFR50.59(b), the written safety evaluation which provides the bases that these modifications do notimpact the safe operation of the plant, constitute an unreviewed safety question, or require achange to the Plant Technical Specifications. As such, prior NRC approval for the implementationof this PC/M is not required.


ABSTRACT

This Engineering Package (EP) provides the engineering justification and design details necessaryto allow use of Fixed Incore Detectors of the "Active Tail" design. This engineering packageinvolves the following tasks:

1) The Fixed Incore Detector locations will be replaced with either a Passive Tail detectorassembly or an Active Tail detector assembly.

2) 'he Instrument Assembly background detector correction factors and sensitivities will bechanged in the plant's DDPS computer system to reflect the new values for each detector.

3) The interconnecting tubing for the Moveable Incore Detection System will be removed fordetector locations where the Active Tail design detector assembly is installed.

The design bases for the Incore Instrumentation System is to monitor neutron flux distributionwithin the reactor core. The data generated is used in the analysis of core conditions. The IncoreInstrumentation System consists of 56 fixed incore detector assemblies and two moveable incoredetectors. Each incore detector assembly consists of four rhodium detectors, one chromel-alumelthermocouple and a Moveable Incore Detector path. The moveable incore detectors are routedby transfer machines and tubing to the detector path in the incore detectors. Information of theneutron flux distribution is provided at selected core locations.

The new Fixed Incore Detector Assemblies have been modified to incorporate minor changes anddo not provide a moveable incore detector path as a result of the current manufacturer'sfabrication methods and applicable ABB-CE specification. These changes do not affect thequalification or function of the incore instrument assembly (Ref. 6.11). The Incore InstrumentationSystem is classified as Not Nuclear Safety. However, the detector assemblies being replaced alsohouses the Core Exit Thermocouples (CETs) which is classified as Safety Related (Ref. 6.1). TheFixed Incore Detectors have inputs to the Digital Data Processing System (DDPS) and the CoreExit Thermocouples have inputs to the Qualified Safety Parameter Display System (QSPDS) whichare used for post accident monitoring. Therefore, this Engineering Package is classified as SafetyRelated.

A safety evaluation for this replacement has been performed in accordance with 10 CFR 50.59.This evaluation concludes that the implementation of this Engineering Package does not involvean unreviewed safety question nor a change to Plant Technical Specifications and has nodetrimental effect on plant safety or operation. Therefore, prior NRC approval for implementationof this modification is not required.


SAFETY EVALUATION

With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shallbe deemed to involve an unreviewed safety question: (i) if the probability of occurrence or theconsequences of an accident or malfunction of equipment important to safety previously evaluatedin the Safety Analysis Report may be increased, or (ii) if a possibility for an accident or malfunctionof a different type than any evaluated previously in the Safety Analysis Report may be created, or(iii) if the margin of safety as defined in the bases for any Technical Specification is reduced. Thereplacement of the Fixed Incore Detector Assemblies with either Passive Tail or Active Tail designdoes not involve an unreviewed safety question'because of the following reasons:

Does the proposed activity increase the probability of occurrence of an accident previouslyevaluated in the SAR?

The proposed activity does not increase the probability of occurrence of an accidentpreviously evaluated in the SAR since the use of either Passive Tail or Active Tail designdetector assemblies meets the design intended for the Incore Instrumentation System asdescribed in the FSAR, Section 7.7.1.1.8 and Technical Specification 3.3.3.2. The FixedIncore Instrumentation System is considered to be a non-safety related system and theIncore Detectors are not considered accident initiators. In addition, the replacement of theincore detectors meets all form, fit and function requirements of the Incore InstrumentationSystem. This replacement will not affect the overall performance and operation of theIncore Instrumentation System.


The proposed activity does not increase the consequences of an accident since the useof either Passive Tail or Active Tail detector assemblies willnot change, degrade or preventactions described or assumed for any accident as discussed in the SAR. The IncoreInstrumentation System does not perform any safety related function and is not requiredfor safe shutdown or to mitigate the consequences of an accident. In addition, thereplacement of the Fixed Incore Detector Assemblies does not alter any assumptionspreviously made in evaluating the radiological consequences of an accident described inthe FSAR.

3. Does the proposed activity increase the probability of occurrence of a malfunction ofequipment important to safety previously evaluated in the SAR?

The proposed activity does not increase the probability of occurrence of a malfunction ofequipment important to safety, because the use of either Passive Tail design or Active Taildesign detector assemblies meets the required design intended for the IncoreInstrumentation System as described in the FSAR, Section 7.7.1.1.8 and TechnicalSpecification 3.3.3.2. The Incore Instrumentation System is a non-safety related system



and no new components or equipment are introduced that could interact with anyequipment important to safety. In addition, the CETs which are housed in the new detectorassemblies are not affected by this changes to the detector assemblies. The replacementdetector assemblies meet all seismic requirements as well as environmental qualificationrequirements.

Does the proposed activity increase the consequences of a malfunction of equipmentimportant to safety previously evaluated in the SAR?

The proposed activity does not increase the consequences of a malfunction of equipmentimportant to safety because the use of either Passive Tail design or Active Tail designdetector assemblies meets the required design intended for the Incore InstrumentationSystem. In addition, the Incore Instrumentation System is a non-safety related system andis not used to prevent or mitigate the consequences of an accident. The replacement ICIdetectors are similar in design such that interfaces with other equipment are not changed,with the exception of the change to the non-safety MICDS.

Does the proposed activity create the possibility of an accident of a different type thanpreviously evaluated in the SAR?

The proposed activity does not create any possibility of an accident of a different type thanpreviously evaluated in the SAR since the use of either Passive Tail design or Active Taildesign detector assemblies does not change the design function of the IncoreInstrumentation System. The replacement detector assemblies will perform the samefunctions as the existing detector assemblies. The Incore Instrumentation System does notperform any safety related functions and is not used to prevent or mitigate theconsequences of an accident. A review of the existing Failure Modes analysis for theIncore Instrumentation System (FSAR, Section 7.7.1.1.8) has been performed. There areno new failure modes introduced as a result of this modification. The removal of theInterconnecting tubing for the Moveable Incore Detection System does not create any newFailure Modes since the interconnecting tubing'is being removed and therefore can notcreate any accident of a different type. No new components or equipment are beinginstalled that could create and accident of a different type that has not already beenevaluated in the SAR. In addition, the overall seismic integrity of the Fixed Incore DetectorAssemblies and Moveable Incore Detection Transfer Assemblies will not be degraded bythis modification.

Does the proposed activity create the possibility of a different type of malfunction ofequipment important to safety than any previously evaluated in the SAR?

po PC/M 120-292 Supplement 0


The proposed activity does not create the possibility of a different type of malfunction ofequipment important to safety because the use of either Passive Tail design or Active Taildesign detector assemblies willnot introduce any new Failure Modes. However, the ActiveTail design detector assemblies will require that the MICDS interconnecting tubing beremoved since the new style detector does not use the interconnecting tubing. Theremoval of the interconnecting tubing does not introduce any new Failure Modes since theinterconnecting tubing is not relied upon for any structural integrity. The replacement ofthe Fixed Incore Detector assemblies and removal of the interconnecting tubing for theMoveable Incore Detection System has not affected the systems interfaces, reliability orperformance of the Incore Instrumentation System.

7 Does the proposed activity reduce the margin of safety as defined in the basis for anyTechnical Specification?

The proposed activity does not reduce the margin of safety as defined in the basis for anyTechnical Specification, since the use of either Passive Tail design or Active Tail designdetector assemblies does not change the required design functions of the IncoreInstrumentation System or the Core Exit Thermocouples as described in TechnicalSpecifications 3.3.3.2, 3.3.3.6. However, the Active Tail design detector assembly does notuse the interconnecting tubing for the Moveable Incore Detection System and theinstallation of this type of detector requires that the interconnecting tubing be removed.The removal of the interconnecting tubing will decrease the number of detector locationsthat the MICDS can position its probe in. This is acceptable since the Fixed IncoreDetectors, are used as the primary instrumentation for monitoring neutron flux within thereactor core and the MICDS are used as the secondary instrumentation. The IncoreInstrumentation System and CETs will still be capable of complying with all therequirements of the Technical Specifications.

The foregoing constitutes, per 10 CFR 50.59(b), the written safety evaluation which provides thebases that this change does not involve an unreviewed safety question nor a change to the PlantTechnical Specifications and prior NRC approval for the implementation of this modification is notrequired.


ABSTRACT

This Engineering Package (EP) provides for the addition of a new Craft Lunchroom Facility (CLF).This facility will replace the existing craft lunchroom which is to be renovated for craft workshops'nd permit the construction of a new Administration Building south of the Unit 2 Reactor AuxiliaryBuilding. The CLF will also house craft toilet facilities and ice house/laundry facilities.

The Craft Lunchroom Facility will be a prefabricated metal building located north of the existingF5 warehouse. The facility will require utility tie-ins for power, potable water and sanitary sewer.The structure will not require a fire sprinkler system, however, the existing fire water supply lineto the F5 warehouse and fire hydrant 34 will be relocated. The facility will also be connected tothe plant paging system (Gai-Tronics) to provide page/party line communications and enableemergency alarms and announcements to be heard within the building.

The Craft Lunchroom Facility does not perform any nuclear safety function as it is a structurewhich provides a personnel facility outside the power block area. The fire protection system andthe paging system, providing audible emergency alarms and announcements, do not performsafety related functions. The section of the fire protection system affected by this modificationdoes not protect safety related equipment, is located outside the Appendix R fire areas and canbe isolated from the main fire loop if failure occurs. Therefore, the fire line relocation is "NotNuclear Safety". Extending the plant paging system (Gai-Tronics) to provide page/party linecommunications and audible emergency alarms and announcements within the building does notaffect any safety related equipment. Therefore, the addition of a page/party line and speakerstation in the CLF is "Not Nuclear Safety". A safety evaluation has been performed in accordancewith 10 CFR 50.59 and is documented in Section 3.0 of this engineering package. This evaluationdemonstrates that implementation of this modification does not involve an unreviewed safetyquestion and does not require a change to the Plant Technical Specifications. In addition, thismodification has no detrimental effects on plant safety or operation. Based upon the above, priorNRC approval is not required for the implementation of this modification.

SAFETY EVALUATION

As defined in 10 CFR 50.59, an unreviewed safety question exists; (i) if the probability ofoccurrence or the consequences of an accident or malfunction of equipment important to safetypreviously evaluated in the Safety Analysis Report (SAR) may be increased; or (ii) if a possibilityof an accident or malfunction of a different type than any previously evaluated in the SAR may becreated; or (iii) if the margin of safety as defined in the basis for any Technical Specification isreduced.

In accordance with 10 CFR 50.59, the following serves to determine whether this modificationconstitutes an unreviewed safety question:

Qe PC/M 097-992 Supplement 0


Does the proposed change increase the probability of occurrence of an accident previouslyevaluated in the SAR?

This modification does not affect any equipment whose malfunction is postulated in theSAR to initiate an accident. In addition the new Craft Lunchroom Facility does not affectany equipment required to prevent an accident from occurring. Therefore, the probabilityof occurrence of an accident previously evaluated in the SAR is not increased.

2. Does the proposed change increase the consequences of an accident previously evaluatedin the SAR?

'henew facility is located such that it will not adversely affect any structure, system, or

component that functions to mitigate the consequences of an accident, to contain or detectthe release of radioactivity, or to provide post-accident shielding. The facility andmodifications to the fire protection system and the plant paging system (Gai-Tronics) donot perform any nuclear safety function and do not interact with any safety related item.Therefore, the consequences of an accident previously evaluated in the SAR will notincrease as a result of this modification.

Does the proposed change increase the probability of occurrence of a malfunction ofequipment important to safety previously evaluated in the SAR?

The construction of the new facility does not affect any equipment whose malfunction isevaluated in the SAR. The section of the fire protection system affected by this modificationdoes not protect safety related equipment, is located outside the Appendix R fire areas andcan be isolated from the main fire loop if failure occurs. The new Gai-Tronics equipmentadded by this modification performs no nuclear safety function and does not interact withany safety-related components. Therefore, the probability of occurrence of any equipmentmalfunction important to nuclear safety previously evaluated in the SAR will not beincreased.


The new facility is located such that it does not adversely affect any structure, system, orcomponent that functions to mitigate the consequences of an accident, to contain or detectthe release of radioactivity, or to provide post-accident shielding. The modification to thefire protection system is designed to ensure that a high level of fire protection is availablefor structures, systems, and components important to safety and is in compliance with theapplicable codes and FSAR requirements for all fire protection equipment. The

extension'o

the Gai-Tronics system will have no impact to nuclear safety. Therefore, theconsequences of a malfunction of equipment important to safety previously evaluated in



the SAR is not changed.

5. Does the proposed change create the possibility of an accident of a different type than anypreviously evaluated in the SAR?

This modification does not change the function or design bases of any structure, system,or component important to safety as described in the SAR. As discussed in the FailureModes and Effects Analysis section, construction of the Craft Lunchroom Facility and theassociated modifications to the fire protection system and the plant paging system do notcreate any new failure modes that can be postulated to cause an accident different thanthose previously analyzed in the SAR. Furthermore, postulated construction accidents donot impact nuclear safety. Therefore, there is no possibility that an accident may becreated that is different from any previously evaluated in the SAR.


The Craft Lunchroom Facility, fire protection system and plant paging system do notperform any nuclear safety functions. Based on the location of the modifications,interaction does not occur with any structure, system, or component important to safety.The section of the fire protection system affected by this modification does not protectsafety related equipment, is located outside the Appendix R fire areas and can be isolatedfrom the main fire loop if failure occurs. No new failure modes are created that can bepostulated to cause a malfunction of equipment important to safety different than thosepreviously analyzed in the SAR. Therefore, the possibility of a malfunction of equipmentimportant to safety which is of a different type than previously evaluated in the SAR is notcreated.

7. Does the proposed change reduce the margin of safety as defined in the basis for anyTechnical Specification?

The Technical Specification requirements applicable to this modification are discussed inSection 3.4 and are not affected. Therefore, this modification does not reduce the marginof safety as defined in the bases for the Technical Specifications.

I

Based on the previous discussion, this modification does not impact safe operation of the plant,constitute an unresolved safety question or require a change to the Technical Specifications.Therefore, this modification does not require prior NRC approval.


ABSTRACT

This engineering package (EP), prepared in accordance with Ql Supplement 3.1-8, provides thereload core design of St. Lucie Unit 2 Cycle 7 developed by Florida Power 8 Light Co. The

. original Cycle 7 energy requirement was 10,725 EFPH, +/- 100 EFPH, based upon a nominalCycle 6 length of 11,400 EFPH. Cycle 6 achieved an EOC exposure of 11,819 EFPH. Thisincreased Cycle 6 exposure will require a coastdown at EOC 7 to meet the Cycle 7 target cyclelength.

r

The primary design change to the core for Cycle 7 is the replacement of 68 irradiated assemblieswith 68 fresh Region J (CE-3) fuel assemblies. The fuel is arranged in a low leakage pattern withno significant differences between the Cycle 7 loading pattern and the Cycle 6 design. Themechanical design of Region J is nearly identical to that of Region H (Cycle 6) and Region G(Cycle 5) reload fuel.

The safety analysis of this design was performed by Asea Brown Boveri Combustion EngineeringNuclear Power, Inc. (ABB/CE) and independently reviewed by Florida Power and Light Co. It hasbeen determined that the operation of the Cycle 7 reload core does not pose an unreviewedsafety question and can be implemented with no changes to the St. Lucie Unit 2 TechnicalSpecifications. Therefore, prior NRC approval is not required for implementation.

M

The implementation of this EP will not adversely impact plant safety or operation.

SUPPLEMENT 1

The purpose of this revision is to include data into the original package that was not available atthe time qf the initial issue. This data is required to support initial startup, power ascension andbeginning of cycle full power operation.

e

SUPPLEMENT 2

Cycle 7 can be operated with a minimum Safety Injection Tank (SIT) pressure of 500 psig.The safety analyses results to support reduction of the SIT minimum allowable pressurefrom 570 psig to 500 psig are presented in Reference 65. However, modification of the.plant to reduce the SIT minimum pressure can only be implemented via MEP 153-292M.This Technical Specification change is currently under review by the NRC.

The measured CEA drop times may exceed the values in the groundrules (Reference 14)provided that they are within the constraints of the analysis results presented in References63 through 65.

0 PC/M 067-292 Supplement 0-3

ABSTRACT (continued)

SUPPLEMENT 3

The purpose of this revision is to include operation data into the package that was not availableprior to issuance of previous Supplements.

SAFETY EVALUATION

In accordance with 10 CFR 50.59 the following discussion demonstrates that there is nounreviewed safety question associated with this reload:

Does the proposed activity increase the probability of an accident previously evaluated inthe FSAR?

The Cycle 7 reload at St. Lucie 2 does not change the overall configuration of the plant.The mode of operation of the plant remains unchanged. Therefore, the probability ofoccurrence of an accident previously evaluated in the FSAR is not changed.

2. Does the proposed activity increase the consequences of an accident previously evaluatedin the FSAR?

The Cycle 7 design remains bounded by the FSAR analyses. The consequences ofaccidents analyzed in the SAR remain unchanged. Therefore, the Cycle 7 reload does notincrease the consequences of an accident previously evaluated in the FSAR.

3. Does the proposed activity increase the probability of an occurrence of a malfunction ofequipment important to safety previously evaluated in the FSAR?

Plant refueling is a normal plant operation. Probabilities of equipment failure during refuelinghave already been incorporated into the plant design basis. Implementation of the St. LucieUnit 2, Cycle 7 reload does not increase the probability of equipment malfunction.Therefore, the probability of occurrence of any equipment malfunction important to safetypreviously evaluated in the FSAR will not increase.

Does the proposed activity increase the consequences of a malfunction of equipmentimportant to safety previously evaluated in the FSAR?

Since Cycle 7 limiting design parameters remain bounded by the existing analyses,consequences of accidents resulting from malfunction of equipment remain unchanged.Therefore, the Cycle 7 reload does not increase the consequences of a malfunction ofequipment important to safety previously evaluated in the FSAR.



5. Does the proposed activity create the possibility of an accident of a different type than anypreviously evaluated in the FSAR?

Fuel reload is a normal plant evolution. Possible accidents have already been postulatedand analyzed in the FSAR. Thus, no new accidents are created. Therefore, the Cycle 7reload does not create the possibility of an accident of a different type than any previouslyevaluated in the FSAR.

6. Does the proposed activity create the possibility of a malfunction of equipment importantto safety of a different type than any previously evaluated in the FSAR?

Fuel reload is a normal plant evolution. Possible equipment malfunctions have already beenpostulated and analyzed in the FSAR. Thus, no new equipment failures are created.Therefore, the Cycle 7 reload does not create the possibility of a malfunction of equipmentimportant to safety of a different type than any previously evaluated in the FSAR.

7. Does the proposed activity reduce the margin of safety as defined in the basis for anyTechnical Specification?

The Cycle 7 design parameters and safety analyses remain bounded by the existinganalyses. Therefore, the margin of safety as defined in the basis for any TechnicalSpecification is not reduced by the Cycle 7 reload.

The above discussion illustrates that there is no unreviewed safety issue. Analysis and review hasalso shown that there is no Technical. Specification change required. Thus, as per 10 CFR 50.59,the reload can be implemented without prior NRC approval.


ABSTRACT

The existing St. Lucie Unit No. 2 turbine generator relay protection scheme does not precludecertain inadvertent non-synchronized connections to the power system. Such events could result'n extensive damage to the main generator and/or turbine.

This Engineering Package (EP) encompasses the engineering/design details for the installationof new relaying and control equipment for the St. Lucie Unit No. 2 main generator as follows:

1. Protection against inadvertent non-synchronized connection to the power system.

2. Revision of the tripping logic of the under-frequency relays.

3. Addition of a synchrocheck relay to supervise closing of the generator breakers.

4. All interconnecting cabling and raceway for the above equipment, as required.

The relays/systems affected by this EP perform no Nuclear Safety Related function. However,since it involves modifications to the RTGB 201 which contains Nuclear Safety Related equipment,and also involves a new circuit from a non-safety section of 125V dc safety bus 2AB, this EP hasbeen classified Quality Related.

The safety evaluation of this EP has shown that the implementation of this PCM does notconstitute an unreviewed safety question as defined in IO CFR 50.59 and does not require achange in the Plant Technical Specifications. This PCM has no adverse impact on plant safetyor operation; thus this PCM can be implemented without prior NRC approval.

Supplement 1

This supplement provides additional evaluation of the safety classification of the PCM and batteryloading. The Safety Analysis has been updated to reflect this evaluation but the conclusionsremain unaffected.

SAFETY EVALUATION

With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shallbe deemed to involve an unreviewed'safety question: (i) if the probability of occurrence or theconsequences of an accident or malfunction of equipment important to safety previously evaluatedin the Safety Analysis Report (SAR) may be increased, or (ii) if a possibility for an accident ormalfunction of a different type than an evaluated previously in the Safety Analysis Report may becreated, or (iii) if the margin of safety as defined in the bases for any Technical Specification isreduced.



The modifications have been evaluated under 10CFR50.59 and it has been determined that themodifications included in this EP does not involve an unreviewed safety question as demonstratedby the answers to the questions below:

1. Does the proposed activity increase the probability of occurrences of an accidentpreviously evaluated in the Safety Analysis Report (SAR)?

This modification does not increase the piobability of occurrence of an accident previouslyevaluated in the safety analysis report since the inadvertent energization relaying ismanually isolated (key switch) with the plant in modes 1 and 2. No accidents evaluatedin the SAR involve any equipment/systems modified by this PC/M. The relayingmodifications enhance main generator protection with no impact on existing accidentanalyses.

2. Does the proposed activity increase the consequences of an accident previously evaluatedin the Safety Analysis Report?

3.

The equipment modified or added by this PC/M will not prevent safety-related equipmentfrom performing their intended functions. As generator protection has no bearing on anyaccidents previously analyzed in the SAR, the implementation of these modifications cannotincrease the consequences of an accident previously evaluated.

Does the proposed activity increase the probability of occurrence of a malfunction ofequipment important to safety previously evaluated in the Safety Analysis Report?

The addition or modification of equipment by this PC/M will not prevent safety-relatedequipment from performing their intended functions. As mentioned above, theimplementation of these modifications cannot increase the probability of occurrence of amalfunction of equipment previously evaluated in the SAR since generator protection is notaddressed therein.

Does the proposed change increase the consequences of a malfunction of equipmentimportant to safety previously evaluated in the Safety Analysis Report?

Modifications performed by this PC/M will not prevent safety-related equipment fromperforming their, intended functions. Since generator protection instrumentation isnon-safety related and has no effect on Nuclear Safety Related equipment, theimplementation of these modifications cannot increase the consequences of a malfunctionof equipment previously evaluated in the SAR.



5. Does the proposed activity increase the possibility of an accident of a different type thanany previously evaluated in the Safety Analysis Report?

The equipment added/modified by this EP is not required during an accident condition norwill it prevent safety related equipment from performing their functions. This modificationdoes not affect any safety related equipment. A failure can only cause turbine trip (modes1 and 2), which is analyzed in FSAR Section 15.2, or a loss of Off-Site Power, which isanalyzed in FSAR Section 15.3. Therefore, the possibility of an accident of a different typethan any evaluated previously in the SAR is not created.

6. Does the proposed activity create the possibility of a malfunction of equipment importantto safety of a different type than any previously evaluated in the SAR?

The possibility of a malfunction of equipment of a different type than evaluated previouslyis not created. The equipment added/ modified by this EP (inadvertent energizationprotective relaying) is powered from a non-safety section of the safety related 125V dc bus2AB. The circuit is provided with electrical double isolation (breaker and fuse in series) andis located in the section of the bus physically isolated (steel barrier) from the safety relatedcircuits. Therefore, failure of the subject circuit, i.e., short circuit, will be isolated from thesafety bus via an actuation of the protective devices, therefore, no Safety Relatedequipment will be affected and no nuclear safety related equipment will be prevented fromperforming their design basis functions.


The margin of safety as defined in the basis for any Technical Specification is not reducedby this modification since the equipment added/modified by this EP does not form thebasis of any Technical Specification. Also, no Plant Technical Specification systemavailability or surveillance requirement is affected by the implementation of this PC/M.

The foregoing constitutes, per 10CFR50.59(b), the written safety evaluation which provides thebases that this change does not involve an unreviewed safety question or a change to plantTechnical Specifications and prior Nuclear Regulatory Commission approval for the implementationof this EP is not required.


ABSTRACT

This Engineering Package (EP) includes the engineering and design necessary to divide thecondenser Air Evacuation System (AES) such that each condenser (2A and 2B) will haveindependent air removal capability. The purpose of the AES is to remove the non-condensiblegases from the condenser, which may blanket condenser tubes from the turbine exhaust steam.The two condensers tend to operate on a lead/lag basis, governed by backpressure, with respectto air removal capability, such that the condenser with the higher backpressure provides themajority of the takeoff gases. In this situation, the air ejectors remove more steam (also, lessnon-condensible gases) and the subcooling required for efficient operation of the ejectors is notachieved. The separation of the two condensers will help to increase the efficiency of the AESby reducing the lead/lag effects of backpressure differences between the condensers.

This EP involves the installation of piping tie-ins and various valves which will provide the abilityto separate the 2A and 2B condenser air evacuation piping. It will also involve the installation oftaps for local pressure indication on each of the four AES takeoff lines as well as attendant valvesfor the steam jet air ejectors to accommodate possible future modifications. The implementationof this EP must occur during an outage since it requires that condenser vacuum be broken toinstall the various components.

This Engineering Package is classified as Not Nuclear Safety, since the affected portions of thecondenser and air evacuation systems as described in the Unit 2 FSAR Sections 10.4.1 and10.4.2, Table 10.4-1, and Figure 10.1-1f perform no safety related functions;

A safety evaluation of this modification has been performed in accordance with 10 CFR 50.59.This evaluation indicates that implementation of this Engineering Package does not involve anunreviewed safety question nor a change to Plant Technical Specifications and has no detrimentaleffect on plant safety or operation. Therefore, prior NRC approval for implementation of thismodification is not required.

SAFETY EVALUATION

With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shallbe deemed to involve an unreviewed safety question: (i) if the probability of occurrence or theconsequences of an accident or malfunction of equipment important to safety previously evaluatedin the Safety Analysis Report may be increased, or (ii) ifa possibility for an accident or malfunctionof a different type than any evaluated previously in the Safety Analysis Report may be created, or(iii) if the margin of safety as defined in the bases for any technical specification is reduced. Themodification included in this Engineering Package does not involve an unreviewed safety questionbecause of the following reasons:

1. Does the proposed activity increase the probability of occurrence of an accident previouslyevaluated in the SAR?



The proposed activity does not increase the probability of occurrence of an accidentpreviously evaluated in the SAR since the portions of the components and systems affectedby this modification do not serve a safety function. They are not required for safeshutdown or to mitigate the effects of a LOCA. This modification does not adversely affectany accident initiating components.


The proposed activity does not increase the consequences of an accident previouslyevaluated in the SAR since this modification will not affect any components or systemsrequired to mitigate the consequences of an analyzed accident. This modification will notaffect the air ejector radiation monitor or any other radiation monitoring equipment.


The proposed activity does not increase the probability of occurrence of a malfunction ofequipment important to safety as previously evaluated in the SAR since this modificationwillnot affect any components or systems which serve a safety function. The componentsand systems affected are not required for safe shutdown or to mitigate the effects of aLOCA.

4.,Does the proposed activity increase the consequences of a malfunction of equipmentimportant to safety previously evaluated in the SAR?

The proposed activity does not increase the consequences of a malfunction of equipmentimportant to safety previously evaluated in the SAR since this modification does not addor modify any equipment which is important to safety. The affected components andsystems are not required for safe shutdown or to mitigate the effects of a LOCA and theirfailure willnot result in the release of significant uncontrolled radioactivity. This modificationwill not adversely affect the air ejector radiation monitors or any other radiation monitoringequipment.

Does the proposed activity create the possibility of an accident of a different type thanpreviously evaluated in the SAR?

The proposed activity does not create any possibility of an accident of a different type thanthat previously evaluated in the SAR since the portions of the components and systemsaffected by this modification do not serve a safety function. They are not required for safeshutdown or to mitigate the effects of a LOCA. This modification does not add any newfailure modes or any equipment capable of initiating an accident. The piping, valves and



pipe supports added by this modification will in no way interact with any safety relatedequipment. The operation of this modified system will not result in any unanalyzed safetyrelated system configurations.

6. Does the proposed activity create the possibility of a different type of malfunction ofequipment important to safety than any previously evaluated in the SAR?

The proposed activity does not create the possibility of a different type of malfunction ofequipment important to safety than any previously evaluated in the SAR because thismodification does not impact any safety related equipment and does not introduce any newfailure modes. The air ejector radiation monitor, which is used to indicate the presence ofa steam generator tube leak, is not impacted by this modification.

7. Does the proposed activity reduce the margin of safety as defined in the basis for anytechnical specification?

The proposed activity does not reduce the margin of safety as defined in the basis for anyTechnical Specification, because the air evacuation system is not included in the basis ofany Technical Specification.

The foregoing constitutes, per 10 CFR 50.59(b), the written safety evaluation which provides thebases that this change does not involve an unreviewed safety question nor a change to the PlantTechnical Specifications. Prior NRC approval for the implementation of this modification is notrequired.


ABSTRACT

This Engineering Package (EP) provides the engineering and design details necessary to add a'igh AC Output Voltage alarm to the 2A, 2B, 2C, and 2D 120 VAC instrument inverters at St.Lucie Unit 2. The need for this new alarm circuit was identified when a 2D" instrument invertermalfunction resulted in a high AC output voltage condition which went undetected for severaldays. In order to facilitate the new alarm, an existing alarm for DC Input Breaker Trippedcondition willbe disconnected. The alarm for DC Input Breaker Tripped can be deleted since thisalarm is enveloped by the Low DC Voltage Alarm. In addition to the new alarm circuit beingadded, this EP is also removing the trip function'of the instrument inverters'C input breaker onLow DC Voltage, High DC Voltage, and Low AC Output Voltage to eliminate automatic power lossto safety related instrumentation during an inverter degraded operating condition. The setpointsfor the instrument inverter alarm circuits are being adjusted (or established if not specified) inaccordance with equipment protection and/or system operational requirements.

This EP involves modifications to the control and alarm circuits of the Class IE 120 VAC instrumentinverters. The instrument inverters are required to achieve and maintain normal safe shutdownconditions and to mitigate the consequences of an accident. Therefore, this PCM is classified asSafety Related.

The safety evaluation of this EP has determined that this PCM does not constitute an unreviewedsafety question as defined in 10 CFR 50.59 and does not require a change in the plant TechnicalSpecifications. This PCM has no adverse impact on plant safety or operation. Therefore, thisPCM can be implemented without prior NRC approval.

SAFETY EVALUATION

With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shallbe deemed to involve an unreviewed safety question: (i) if the probability of occurrence or theconsequences of an accident or malfunction of equipment important to safety previously evaluatedin the Safety Analysis Report (SAR) may be increased, or (ii) if a possibility for an accident ormalfunction of a different type than any evaluated previously in the SAR may be created, or (iii)if the margin of safety as defined in the bases for any Technical Specification is reduced.

In accordance with 10CFR50.59, the following evaluation serves to determine whether thismodification constitutes an unreviewed safety question of requires a change to the TechnicalSpecifications:

1. Does the proposed change increase the probability of occurrence of an accident previouslyevaluated in the SAR?



Each instrument inverter provides 120 VAC power to one of the four channels of RPS andESFAS which function to mitigate the consequences of an accident. The malfunction ofthe instrument inverters is not postulated to initiate an accident previously evaluated in theSAR. Since two out of four criteria is used in the logic of all protection systems, amalfunction of a single instrument inverter will not cause spurious actuation of theprotection systems. The modifications being implemented by this EP do not impact theindependence, redundancy or operation of the instrument inverters or its loads. The newAC voltage sensing cards are seismically qualified and mounted to eliminate the possibilityof a common mode failure during a seismic event. Therefore, the probability of occurrenceof an accident previously evaluated in the SAR is not increased.


The instrument inverters supply uninterruptible 120 VAC power to instrumentation andprotection systems required to mitigate the consequences of an accident. The redundantfeatures of the vital 120-VAC system ensure that a single instrument inverter failure will notprevent actuation of the protection systems or cause loss of vital process control andmonitoring systems.

The addition of an alarm for High AC Output Voltage and the adjustment of existing alarmsetpoints will help minimize the possibility of an undetected inverter degraded operatingcondition. The new AC voltage sensing cards are electrically connected in accordance withexisting design criteria and seismically qualified and mounted to prevent adverse affects onthe inverter control circuitry. The disconnection of the trip circuitry to the inverter DC inputbreaker does not alter the failure mode analysis of the instrument inverters. Therefore, theproposed modifications do not compromise the ability of the instrument inverters theysupply to perform their safety function. Accordingly, the consequences of an accidentpreviously evaluated in the SAR have not been increased.


A malfunction of an instrument inverter voltage regulation circuitry resulted in a high ACoutput voltage condition which went undetected for several days. The addition of an alarmfor this condition will not reduce or increase the probability of this malfunction fromoccurring; however, it will ensure that this condition does not go undetected in the future.The new AC voltage sensing cards will be seismically qualified and seismically mountedwithin the existing instrument inverter cabinets. As discussed in the Design Section (2.3.4)of this EP, this modification will have no adverse impact on the seismic qualification of theexisting inverter cabinets.



The removal of the trip circuitry of the DC input breaker does not alter the associatedfailure mechanisms (i.e. Low/High DC Voltage) or failure consequences (i.e. Low ACOutput Voltage). Although these conditions may still cause a loss of the instrumentinverter, the actual failure of the inverter and/or its loads must now occur, not just thedetection of a off-normal operating condition. In addition, the loss of an instrument inverterdue to a transient input/output voltage or malfunction of the voltage sensing cards hasbeen eliminated. Therefore, the proposed change does not increase and actuallydecreases the probability of occurrence of a malfunction of equipment important to safetypreviously evaluated in the SAR.


The instrumentation and control equipment powered by the instrument inverters aredesigned and electrically aligned such that a single inverter failure will not prevent safeshutdown or reduce the ability to mitigate the consequences of an accident. A singleinverter failure will also not cause spurious actuation of protection systems due to the useof four independent process channels and two out of four logic for actuation. Theproposed changes to the alarm and trip circuitry do not affect the redundancy andindependence of the 120 VAC system or alter the inverters'ailure mode analysis.Therefore, the consequences of a malfunction of equipment important to safety previouslyevaluated in the SAR has not been increased.


This modification does not change the design bases, operation, or function of theinstrument inverters or its loads. No new hazards are created that can be postulated tocause an accident different than those previously analyzed in the SAR. There are no newfailure modes introduced and the consequences of an inverter failure have not beenchanged. Therefore, there is no possibility that an accident may be created that is differentfrom one already evaluated in the SAR.


The modifications are limited to the internal alarm and control circuitry of the instrumentinverters. No new hazards are created that can be postulated to cause a malfunction ofequipment important to safety different than those analyzed in the SAR. Therefore, thepossibility of a malfunction of equipment important to safety which is of a different type thanpreviously evaluated in the SAR is not created.


SAFETY EVALUATlON(continued)

7. Does the proposed change reduce the margin of safety as defined in the bases for anyTechnical Specification?

This modification does not adversely affect the operational requirements or reliability of theinstrument inverters. The additional alarm for High AC Output Voltage and the conservativeadjustment of the existing alarm setpoints will provide the operator with improvednotification of an inverter degraded operating condition. The deletion of the trip circuitryof the inverters'C input breaker reduces the possibility of an inadvertent loss of aninstrument inverter. Since there is no impact on the Surveillance Requirements and LimitingConditions for Operation of the 120 VAC System, this modification does not reduce themargin of safety as defined in the bases for any Technical Specification.

Based on the previous discussion, this modification does not impact safe operation of the plant,constitute an unresolved safety issue or require a change to the Technical Specifications.Therefore, this modification does not require prior NRC approval.


ABSTRACT

This Engineering Package (EP) provides the engineering and documentation necessary todisconnect and abandon in place specific heat trace circuits in the Liquid Waste ManagementSystem at St. Lucie Unit 2. The heat tracing circuits being disconnected are associated with theWaste Concentrator Package. The heat tracing is not required due to the boric acid solution inthe Waste Concentrator Package being administratively controlled at a concentration up to 3.5weight percent.'oric acid solution with a concentration up to 3.5 weight percent present in pipelines and components located in the AuxiliaryBuilding do not need heat tracing per FSAR Section9.3.4.3.1 ~ 1.

The Liquid Waste Management System (LWMS) has the potential for personnel radiationexposure. Therefore, this modification is classified as Quality Related.

The safety evaluation of this EP has determined that this PC/M does not constitute an unreviewedsafety question as defined in 10 CFR 50.59 and does not require a change in the Plant TechnicalSpecifications. This PC/M has no adverse impact on plant safety or operation; thus this PC/Mcan be implemented without prior NRC approval.

SAFETY EVALUATION

With respect to Title 10 of the Code of Federal Regulations, Part 10 CFR 50.59 Safety Evaluation,a proposed change shall be deemed to involve an unreviewed safety question: (i) ifthe probabilityof occurrence or the consequences of an accident or malfunction of equipment important to safetypreviously evaluated in the Safety Analysis Report (SAR) may be increased, or (ii) if a possibilityfor an accident or malfunction of a different type than any evaluated previously in the SafetyAnalysis Report may be created, or (iii) if the margin of safety as defined in the bases for anyTechnical Specification is reduced. In accordance with 10CFR50.59, the following evaluationserves to determine whether this modification constitutes an unreviewed safety question:


The malfunction of LWMS is not postulated to initiate an accident previously evaluated inthe SAR. The design objective of the LWMS is to protect plant personnel, the generalpublic and the environment by assuring all releases of radioactive liquids is performed ina controlled manner that meets the requirements of 10CFR20 and 10CFR50, Appendix I.The disconnection of the WCP heat tracing does not compromise the ability of the LWMSto meet this design objective because of the redundant function performed by the BoricAcid Concentrators and administrative controls imposed on the boric acid concentrationof the liquid wastes processed by the WCP. The LWMS will still be capable of effectivelycontrolling, monitoring and processing radioactive liquid wastes in accordance with theexisting design criteria after the WCP heat tracing is disconnected. The design provisions



and controls provided to prevent inadvertent or uncontrolled release of radioactive liquidare not affected. Therefore, the probability of an accident previously evaluated in the FSARhas not been increased.


No credit is taken for operation of the LWMS to mitigate the consequences of any designbasis accident. The system is used to control, monitor and process liquid wastes duringnormal operation of the plant. The disconnection of the WCP heat tracing does not affectthe operation of the effluent line radiation monitor or any isolation valves whichautomatically close on inadvertent release of radioactive liquids. The design provisionsprovided to control the release of radioactive materials due to waste surges or tankoverflows is also not affected. Therefore, the consequences of an accident previouslyevaluated in the FSAR have not been increased.

Does the proposed change increase the probability of an occurrence of a malfunction ofequipment important to safety previously evaluated in the SAR?

The function of the WCP heat tracing was to prevent equipment/system failure due tosolidiTication of the boric acid in the pipe lines, components, or tanks: The administrativecontrols imposed on the boric acid concentration of liquid waste processed by the WCP(i.e <3.5 weight percent) assures that the disconnection of the WCP heat tracing circuitswill not increase the probability of equipment/system failure. Liquid wastes with a boricacid concentration of greater than 3.5 weight percent is routed to the Boric AcidConcentrators for processing. Since the LWMS is operated in a batch mode and isdesigned to handle surges in influent rate, additional waste liquid processing by the BoricAcid Concentrators will not affect overall system performance or equipment designspecifications to be exceeded. Therefore, the probability of a malfunction of equipmentimportant to safety previously evaluated in the FSAR would not be increased.

d


For the worst-case scenario, the malfunction of equipment in the LWMS would cause theinadvertent or uncontrolled release of radioactive materials. The disconnection of the WCPheat tracing circuits does not increase the severity of this worst-case scenario because theamount of radioactivity or the overall effluent volume and flow rate has not been changed.The design features (e.g. radiation monitoring, alarms, fail safe isolation valves) providedto mitigate the consequences of malfunction of equipment are not affected by this PCM.Therefore, the consequences of a malfunction of equipment important to safety previouslyevaluated in the SAR has not been increased.




This modification does not change the design bases, functional requirements, orperformance of the LWMS. No new hazards are created that can be postulated to causean accident different than those previously analyzed in the SAR. There are no new failuremodes introduced and the consequences of the LWMS failure have not been changed.

'herefore, the proposed activity does not create the possibility that an accident may becreated that is different from any already evaluated in the SAR.


The disconnection of the WCP heat tracing circuits does not alter the failure modes ofassociated equipment since administrative controls have been imposed on the boric acidconcentration (< 3.5 weight percent) of the liquid waste processed by the WCP. Theambient temperature of the auxiliary building (location of the WCP) will be sufficient toprevent solidification of boric acid within the WCP, which is the only possible failuremechanism applicable to a loss of heat tracing. No new hazards are created that can bepostulated to cause an accident different than'those previously analyzed in the SAR.Therefore, the possibility of a malfunction of equipment important to safety which is of adifferent type than previously evaluated in the SAR is not created.

Does the proposed change reduce the margin of safety as defined in the basis for anyTechnical Specifications?

The basis for the LWMS is to ensure that releases of radioactive materials in liquid effluentsbe kept As LowAs is Reasonably Achievable (ALARA)in accordance with the requirementsof 10CFR20 and 10CFR50, Appendix I. Appropriate portions of the LWMS must beavailable whenever liquid effluents require treatment prior to release to the environment.The restriction on the boric acid concentration (i.e. < 3.5%) of the liquid wastes processedby the WCP is acceptable since the liquid wastes will be routed to the Boric AcidConcentrators, where the redundant processing is performed. As a result, the availabilityof the LWMS to perform its design basis function is not compromised. The disconnectionof the WCP heat tracing circuits does not affect the effluent monitoring or design featuresfor preventing inadvertent or uncontrolled release of the radioactive materials. Therefore,this modification does not reduce the margin of safety as defined in the bases for theTechnical Specifications.



The foregoing constitutes, per 10 CFR 50.59(b), the written safety evaluation which provides thebases that this change does not involve an unreviewed safety question nor a change to the plant

'echnical Specifications and prior NRC approval for the implementation of this modification is notrequired.


ABSTRACT

NRC Generic Letter 89-10 requires that operating nuclear plants develop and implement aprogram to ensure that switch settings on all safety-related motor-operated valves (MOVs) arecorrectly selected, set and maintained to accommodate the maximum differential pressuresexpected on these valves during all postulated events within the design basis. Item a) of theLetter requires that the design basis for these MOVs be reviewed to determine the maximumdifferential pressure expected during both opening and closing strokes for all postulated events.This has been completed and documented in FPL Calculation PSL-2FJM-91-046, Revision 1, andFPL'Engineering Evaluation JPN-PSL-SEMP-91-048, Revision 0.

Item b) of Generic Letter 89-10 requires that the licensee establish the correct MOVswitch settingsbased on the previously determined maximum differential pressure. AIIswitches, including torqueswitches, torque bypass switches, position limit, position indication, overloads, etc., shall beconsidered. This design package provides the overall switch setting guidelines for fifty-eight (58)motor operated valves, in addition to specific design information, as determined by calculation,necessary to replace actuator spring packs and set both the open and close torque switches tomeet the requirements of Generic Letter 89-10 for the valves identified herein.

Because the motor-operated valves associated with Generic Letter 89-10 are safety-related, ormay affect safety-related systems, this engineering package has been classified as Safety Related.A review of the switch setting changes to be implemented by the PC/M was performed againstthe requirements of 10CFR50.59, and it was concluded that these modifications do not constitutean unreviewed safety question and do not require a change to the plant Technical Specifications.Therefore, prior NRC approval for the implementation of this PC/M is not required.

Supplement 1 of this Engineering Package is issued to reflect revised thrust values for numerousvalves. The revised thrust values resulted from detailed system reviews performed for the valves,or from new or revised vendor information. Two (2) valves were added to the scope of themodification increasing the total to sixty (60) valves. In addition, the methodology for setting theopen torque switch is revised due to a change in the diagnostic equipment which has beenselected for use. This supplement revises the safety evaluation to reflect the revised designinformation. However, the original conclusions of the safety evaluation, that the change does notconstitute an unreviewed safety question nor require a change to the plant TechnicalSpecifications remain unchanged. Therefore, prior NRC approval for the implementation of thisPC/M is not required.


SAFETY EVALUATION

As defined in 10 CFR 50.59, an unreviewed safety question exists; (i) if the probability ofoccurrence or the consequences of an accident or malfunction of equipment important to safety

'reviously evaluated in the Safety Analysis Report (SAR) may be increased; or (ii) if a possibilityof an accident or malfunction of a different type than any previously evaluated in the SAR maybe created; or (iii) if the margin of safety as defined in the basis for any Technical Specificationis reduced.

In accordance with 10 CFR 50.59, the following evaluation serves to determine whether thismodification constitutes an unreviewed safety question:


This modification does not affect any equipment whose malfunction is postulated in theSAR to initiate an accident or prevent an accident from occurring. The modificationsperformed by this Engineering Package enhance the ability of the components to performas intended during emergency and off-normal conditions under maximum differentialpressures.

Replacement of the actuator spring packs and revising the thrust or torque values for theMOV operators only serve to enhance the operational characteristics of the MOVs. Assuch, no new accident initiating events are created. Therefore, the modifications describedin this Engineering Package do not increase the probability of valve failure, and thus theprobability of occurrence of an accident previously described in the SAR is not increasedby.this modification.


This modification does not affect any structures, systems or components that function todeter the release of radioactivity or to provide post-accident shielding. The modificationsperformed by this Engineering Package do affect systems and components that are reliedupon to mitigate accident consequences, and contain radioactive fluids. However themodification performed improves the operational characteristics of the valves and improvesthe equipments ability to function during an accident. Therefore, the consequences of anaccident previously evaluated in the SAR are not increased by this modification.




System operability is not being affected by the modifications to the MOVs identified in thisEngineering Package. Valve operability will be enhanced by the prescribed modifications.In addition, no new failure modes are created as a result of this modification, as thismodification serves to provide additional design documentation, or replace existing parts.

Replacement of the spring packs and revising the thrust or torque values for the MOVoperators only serve to enhance the operational characteristics of the MOVs. As such, nonew accident initiating events are created. Therefore, the probability of occurrence of amalfunction of equipment important to safety previously evaluated in the SAR has notincreased by this modification.


System operation is not affected by this modification. This modification does not interactspatially or functionally with any structure, system or component important to safety otherthan the valves and valve operators themselves. Although actuator and valve loadings mayincrease, the revised loads are within the published ratings for the components.Replacement components have been selected in accordance with the same design criteriaas the original components. The modifications performed by this Engineering Packageenhance the ability of the valves and valve operators to perform as intended duringemergency and off normal conditions under maximum differential pressures. Therefore,the consequences of a malfunction of equipment important to safety previously evaluatedin the SAR is not increased by this modification.


As discussed in Section 3.3, this modification does not change the function or designbases of any structure, system or component important to safety as described in the SAR.This modification provides increased design documentation, makes adjustments tocomponents within their published operating range or makes replacements of equivalentparts. No new failure modes or conditions are created that can be postulated to cause anaccident different than those previously analyzed in the SAR. Therefore, the possibility ofan accident of a different type than any previously evaluated in the SAR is not created bythis modification.



6. Does the proposed change create the possibility of a malfunction of equipment importantto safety of a different type than any previously evaluated in the SAR?

This modification does not interact spatially or functionally with any structure, system orcomponent important to safety other than the valves and valve operators themselves. Thismodification does not alter the function or the design basis of any MOV. As discussed inSection 3.3, no new failure modes are created for the subject MOVs that can be postulated'o cause a malfunction of equipment imp'ortant to safety different than those previouslyanalyzed in the SAR. Therefore, the possibility of a malfunction of equipment important tosafety which is of a different type than any previously evaluated in the SAR is not createdby this modification.


The Technical Specification requirements and Technical Specification Bases are notaffected by this modification. The design bases of the valves and valve operators remainsunchanged. Therefore, the margin of safety as defined in the bases for any TechnicalSpecification is not reduced by this modification.

The foregoing discussions provided in this EP constitutes, per CFR 50.59(b), the written safetyevaluation which provides the bases that these modifications do not impact the safe operation ofthe plant, constitute an unreviewed safety question, or require a change to the plant TechnicalSpecifications. As such, prior NRC approval for the implementation of this PC/M is not required.


ABSTRACT

This Engineering Package (EP) provides for the addition of a removable visual level indicator tothe spent fuel pool (SFP). The level indicator will aid the Operations Department in the

. determination of the pool level when the existing remote alarm circuitry is inoperable. Therefore,this modification will help to ensure that unacceptable pool levels do not occur.

The visual level indicator will be attached, using a concrete expansion anchor, to the SeismicCategory I Fuel Handling Building (FHB) and has been seismically designed to precludeinteraction with spent fuel assemblies stored in the SFP. Accordingly, this PC/M has beenclassified as Quality Related. A safety evaluation has been performed in accordance with 10 CFR50.59 and is documented in this engineering package. This evaluation demonstrates thatimplementation of this modification does not involve an unreviewed safety question and does notrequire a change to the Plant Technical Specifications. In addition, this modification has nodetrimental effects on plant safety or operation. Based upon the above, prior NRC approval isnot required for the implementation of this modification.

SAFETY EVALUATION

As defined in 10CFR50.59, an unreviewed safety question exists; (i) ifthe probability of occurrenceor the consequences of an accident or malfunction of equipment important to safety previouslyevaluated in the Safety Analysis Report (SAR) may be increased; or (ii) if a possibility of anaccident or malfunction of a different type than any previously evaluated in the SAR may becreated; or (iii) if the margin of safety as defined in the basis for any Technical Specification isreduced.



This modification does not affect any equipment whose malfunction is postulated in theSAR to initiate an accident. In"addition, the visual level indicator cannot interact with anyequipment required to prevent an accident from occurring. Therefore, the probability ofoccurrence of an accident previously evaluated in the SAR is not increased.


This modification is designed to ensure that it does not adversely affect any structure,system, or component that functions to mitigate the consequences of an accident, tocontain or detect the release of radioactivity, or to provide post-accident shielding. The



visual level indicator does not perform any nuclear safety function and has been seismicallydesigned to preclude interaction with any safety related item. Therefore, the consequencesof an accident previously evaluated in the SAR will not increase as a result of thismodification.


The visual level indicator is designed to ensure that interaction does not occur with anystructure, system, or component important to safety. The visual level indicator has beenseismically designed to preclude interaction with any safety related items and will notinterfere with the operation of any nuclear safety related systems. Therefore, the probabilityof occurrence of any equipment malfunction important to safety previously evaluated in theSAR will not be increased.


The visual level indicator is designed to ensure that interaction does not occur with anystructure, system or component important to safety. The visual level indicator is seismicallydesigned to preclude interaction with any safety related items. Therefore, theconsequences of a malfunction of equipment important to safety previously evaluated inthe SAR is not changed.

E


This modification does not change the function or design bases of any structure, system,or component important to safety as described in the SAR. Per The Failure Modes andEffects Analysis section of this EP, installation of the visual level indicator does not createany new hazards that can be postulated to cause an accident different than thosepreviously analyzed in the SAR. The visual level indicator does not perform any nuclearsafety function. Therefore, there is no possibility that an accident may be created that isdifferent from any previously evaluated in the SAR.


The visual level indicator does not perform any nuclear safety function and is designed toensure that interaction does not occur with any structure, system, or component importantto safety. No new failure modes which could adversely affect equipment important tosafety are created. The visual level indicator has been seismically designed to preclude



interaction with any safety related items. No new hazards are created that can be postulated tocause a malfunction of equipment important to safety different than those previously analyzed inthe SAR. Therefore, the possibility of a malfunction of equipment important to safety which is ofa different type than previously evaluated in the SAR is not created.


The Technical Specification requirements and bases applicable to this modification arediscussed earlier in this EP and are not affected. Therefore, this modification does notreduce the margin of safety as defined in the bases for the Technical Specifications.


II


ABSTRACT

This Engineering Package (EP) includes the engineering 8 design necessary to delete the MainTurbine Runback feature. The modification will leave the Turbine Runback logic unchanged. Itwill no longer be initiated. Turbine Runback occurs whenever there is a loss of a SteamGenerator Feed Pump (SGFP) above 60% power or the loss of both Heater Drain Pumps (HDP)above 70% power. Deleting the Turbine Runback feature will be accomplished by lifting theTurbine Runback leads in the Turbine Digital Electronic Hydraulic Control (DEH) Cabinet,Sequence of Events Cabinet, Annunciator Logic Cabinet (ALC-I), and at the Reactor Turbine

'eneratorBoard (RTGB) 202. Cables 20712F, 20712J and 20712L will be spared and alldrawings willbe revised accordingly. In addition the Control'Room Turbine Runback AnnunciatorWindow D27 will be rendered inoperative and will be spared.

The function of the Turbine Runback is to run the Turbine/Generator back at a predetermined rateupon loss of a SGFP or both HDP's until Turbine/Generator output decreases to 60% 8c 70%respectively as measured by turbine first stage (impulse) pressure. During a Turbine Runbackthe Main Governor valves throttle the steam flow until the load matches the setpoint of 60% or70% load depending on the initiating event. During this event the Turbine/Generator RPM remainsconstant.

St. Lucie Unit 1 experienced a Reactor trip from 100% power on June 14, 1987 & June 30, 1988,due to a Turbine Runback which was caused by the loss of the 1B SGFP. During both eventsa turbine runback was automatically initiated to approximately 60% power. In less than 30seconds into the transient the Reactor Protection System initiated a Reactor trip on a highpressurizer pressure signal.

The purpose of removing the Turbine Runback feature is to minimize the effects of a partial lossof feedwater transient on the Plant and to provide the plant operators with additional time torestore 100% feedwater flow before a Reactor trip occurs. Although the two previously mentionedtransients that caused a Reactor trip are for Unit 1, removal of the Turbine Runback feature willbe implemented for Unit 2 by this PC/M since both Units respond to this transient in a similarmanner and benefit from its removal as demonstrated by a Thermal Hydraulics Analysis.

A Thermal Hydraulics Analysis of a loss of SGFP transient was performed with and without theTurbine Runback feature. This analysis demonstrates that by removing the Turbine Runbackfeature a Reactor trip could be avoided, provided the plant operators restore 100% feedwater flowwithin 100 seconds into the event. If full feedwater flow is not restored within the 100 seconds,a reactor trip will occur on low steam generator levels. This transient will not challenge thePressurizer Power Operated Relief Valves (PORVs) or the Main Steam Safety Valves (MSSVs)which lifted.



Although the Main Turbine, Turbine Controls and the Turbine Runback feature do not perform asafety function per FSAR Section 7.7, this EP is classified as Quality Related because it requireswork to be performed in the Control Room.

A safety evaluation of this modification has been performed in accordance with 10 CFR 50.59.This evaluation indicates that implementation of this Engineering Package does not involve anunreviewed safety question nor require a change to Plant Technical Specification and has nodetrimental effect on plant safety or operation. Therefore, prior NRC approval for implementationof this modification is not required.

SAFETY EVALUATlON

With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shallbe deemed to involve an unreviewed safety question: (i) if the probability of occurrence or theconsequences of an accident or malfunction of equipment important to safety previously evaluatedin the Safety Analysis Report may be increased, or (ii) if a possibility for an accident or malfunctionof a different type than any evaluated previously in the Safety Analysis Report may be created, or(iii) if the margin of safety as defined in the bases for any Technical Specification is reduced. Themodification included in this engineering package does not involve an unreviewed safety questionbecause of the following reasons:

Does the proposed activity increase the probability of occurrence of an accident previouslyevaluated in SAR?

The proposed activity does not increase the probability of occurrence of an accidentpreviously evaluated because the functionality of the Main Turbine, Turbine Controls orReactor Protection trip signals have not been changed by this modification. Based on this,the probability of occurrence of an analyzed accident remains unchanged.

2. Does the proposed activity increase the consequences of an accident previously evaluatedin the SAR?

The proposed activity does not increase the consequences of an accident due to thedeletion of the Main Turbine Runback feature because the Main Turbine, Turbine Controls,or Runback feature do not serve a Safety Related function. This modification does notaffect any structure, systems or components that function to deter the release ofradioactivity or to provide post-accident shielding.

0Does the proposed activity increase the probability of occurrence of a malfunction ofequipment important to safety previously evaluated in the SAR?



The proposed activity does not increase the probability of a malfunction of equipmentimportant to safety because the Turbine Runback feature is not Safety Related and thedeletion of the Turbine Runback does not affect any Safety Related signals required toinitiate a Reactor trip. This modification does not degrade the reliability or increasechallenges, directly or indirectly for equipment important to safety.


The proposed activity does not increase. the consequences of a malfunction of equipmentimportant to safety because neither the Main Turbine Controls or Turbine Runback featureperform a Nuclear Safety Related function and are not used to mitigate the consequencesof an accident.

Does the proposed activity create the possibility of an accident of a different type than anypreviously evaluated in the SAR?

The proposed activity does not create the possibility of an accident of a different type thanany previously evaluated because during the loss of a SGFP transient without TurbineRunback the heat removal from the Reactor Coolant System (RCS) by the secondary sideof the Plant is maintained and does not challenge the Reactor Protection System.

The RCS pressure and temperature are not increased as they are with the TurbineRunback feature operative. The Turbine Runback feature is not Safety Related and thedeletion of the Turbine Runback does not affect any Safety Related signals required toinitiate a Reactor trip. This modification does not degrade the reliability or increasechallenges, directly or indirectly for equipment important to safety. For these reasons, theproposed activity does not create the possibility of an accident of a different type thanpreviously described in the FSAR.

Does the proposed activity create the possibility of a malfunction of equipment importantto safety of a different type than any previously evaluated in the SAR?

The proposed activity does not create the possibility of a malfunction of equipmentimportant to safety of a different type than any previously evaluated because the TurbineRunback feature is not Safety Related and the deletion of the Turbine Runback does notaffect any Safety Related signals required to initiate a Reactor trip. This modification doesnot degrade the reliability or increase challenges, directly or indirectly for equipmentimportant to safety.

Does the proposed activity reduce the margin of safety as defined in the bases for anyTechnical Specification?



The proposed activity does not reduce the margin of safety as defined in the basis for anyTechnical Specification, because the Turbine Runback feature is not included in the basisof the Technical Specification for the Reactor Coolant System or any TechnicalSpecification.

The foregoing discussions provided in this EP constitutes, per 10 CFR 50.59(b), the written safetyevaluation which provides the bases that these modifications do not impact the safe operation ofthe plant, constitutes an unreviewed safety questions, or require a change to the Plant TechnicalSpecifications. As such, prior NRC approval for the implementation of this PC/M is not required.


ABSTRACT

During operation of the 2A and 2B Emergency Diesel Generators (EDGs), the exposed rotatingshaft couplings between the generator and diesel engines present a personnel hazard. Thisengineering package provides for the installation of sheet metal guards around the two shaftcouplings located on each EDG. Installation of the guards will enhance personnel safety byproviding a barrier between personnel and the rotating shafts.

The components added by this engineering package do not perform any nuclear safety function.However, failure of the shaft guards could result in interaction with the safety related EDGs.Therefore, this engineering package is classified Quality Related.

The modifications provided by this Engineering Package will not adversely affect plant safety oroperation, do not constitute an unreviewed safety question, and do not require a change to theTechnical Specifications. Therefore, prior NRC approval for implementation is not required.

SAFETY EVALUATION

As defined in 10CFR50.59, an unreviewed safety question exists; (i) ifthe probability of occurrenceor the consequences of an accident or malfunction of equipment important to safety previouslyevaluated in the Safety Analysis Report (SAR) may be increased; or (ii) if the possibility for anaccident or malfunction of a different type than any previously evaluated in the SAR may becreated; or (iii) if the margin of safety as defined in the basis for any Technical Specification isreduced.



This modification does not affect any equipment whose malfunction is postulated in theSAR to initiate an accident. In addition, the EDG shaft coupling guards are designed toensure that interaction does not occur with the EDGs or any other equipment required toprevent an accident from occurring. Therefore, the probability of occurrence of an accidentpreviously evaluated in the SAR is not increased.




This modification is designed to ensure that it does not adversely affect any structure,system, or component that functions to mitigate the consequences of an accident, tocontain or detect the release of radioactivity, or to provide post-accident shielding. TheEDG shaft coupling guards do not perform any nuclear safety function and have beenseismically designed to preclude interaction with any safety related items. Therefore, theconsequences of an accident previously evaluated in the SAR will not increase as a resultof this modification.

3. Does the proposed change increase the probability of occurrence of a malfunction ofequipment important to safety previously evaluated in the SAR?

The EDG shaft coupling guards are designed to ensure that interaction does not occur withany structure, system, or component important to safety. The EDG shaft coupling guardsare seismically designed to preclude interaction with any safety related items and will notinterfere with operation of the EDGs. Therefore, the probability of occurrence of anyequipment malfunction important to safety previously evaluated in the SAR will not beincreased.

'oes the proposed change increase the consequences of a malfunction of equipmentimportant to safety previously evaluated in the SAR?

The EDG shaft coupling guards are designed to ensure that interaction does not occur withany structure, system or component important to safety. The EDG shaft coupling guardsare seismically designed to preclude interaction with any safety related items. Therefore,the consequences of a malfunction of equipment important to safety previously evaluatedin the SAR is not changed.


This modification does not change the function or design bases of any structure, system,or component important to safety as described in the SAR. Installation of the EDG shaftcoupling guards does not create any new hazards that can be postulated to cause anaccident different than those previously analyzed in the SAR. The EDG shaft couplingguards do not perform any nuclear safety function. Therefore, there is no possibility thatan accident may be created that is different from any previously evaluated in the SAR.




7.

The EDG shaft coupling guards do not perform any nuclear safety function and aredesigned to ensure that interaction does not occur with any structure, system, orcomponent important to safety. No new failure modes which could adversely affectequipment important to safety are created. The EDG shaft coupling guards are seismicallydesigned to preclude interaction with any safety related items. No new hazards arecreated that can be postulated to cause a malfunction of equipment important to safetydifferent than those previously analyzed in the SAR. Therefore, the possibility of amalfunction of equipment important to safety which is of a different type than previouslyevaluated in the SAR is not created.

Does the proposed change reduce the margin of safety as defined in the basis for anyTechnical Specification?

The Technical Specifications requirements and bases applicable to this modification arediscussed in Section 3.4 of this EP and are not affected. Therefore, this modification doesnot reduce the margin of safety as defined in the bases for the Technical Specifications.



ABSTRACT

The atmospheric dump valves (ADVs) actuator control logic was modified (PCM 069-286) so thatthe closing direction of these valves would be controlled by the torque switch (in lieu of the limitswitch) to ensure positive valve seating. However, difficultyhas been experienced in adjusting thetorque switch of these motor operated valves to achieve the minimum amount of thrust requiredfor closing the valve, and not exceed the vendors'ecommended maximum allowable total thrust.The problem is due to the significant amount of thrust produced after the torque switch is tripped,which is a result of the relatively high stem speed. This additional thrust, or overtravel, can besignificantly reduced in the closing direction by installing a compensating spring pack on theactuator. This field modification willconvert the ADVactuator from the existing-Limitorque SMB-0model to be functionally the same as an SB-0 type actuator. The capacity of the actuator is notadversely affected by this modification, as these Limitorque models have interchangeable partsand have the same thrust and torque ratings. Since the ADVs are controlled by a limit switch inthe open direction, it is not necessary to provide a compensating spring pack in this direction.

This Engineering Package provides for the modification of the St. Lucie Unit 2 ADVactuators, tagnumbers MV-08-18A, -18B, -19A, -19B. The actuators will be modified by installing acompensating spring pack assembly; this spring pack, and its cover, will be installed on top ofthe existing actuator housing. It is actually a Belleville spring mounted above the stem nut. Thestem nut is thereby allowed to "float" upward upon valve seating, rather than being rigidly held inplace. The spring will compress a fixed amount for a given seating load, and will compensate forload variations should the ADV cool or heatup.

Following implementation of this PCM, the torque switch settings will be adjusted to satisfy therequirements of.Generic Letter 89-10; the requirements are provided in PCM 510-291.

Because the ADVs are Safety Related, this engineering package has been classified as SafetyRelated. A review of the modification to be implemented by this PC/M was performed against therequirements of 10CFR50.59, and it was concluded that these modifications do not constitute anunreviewed safety question and do not require a change to the plant Technical Specifications.Therefore, prior NRC approval for the implementation of this PC/M is not required.

SUPPLEMENT NO. 1

Supplement No. 1 to this Engineering Package provides the vendor instructions to modify the ADVactuator, provides additional drawings, adds the TEDB change package and lists additionalEngineering procured parts. Neither the design, the installation, nor the performance of the ADVactuators is adversely affected by this Supplement. Supplement No. 1 does not effect, amend norchange the original Safety Evaluation, the Technical Specifications or the Technical Specificationbases.

e


SAFETY EVALUATION

As defined in 10 CFR 50.59, an unreviewed safety question exists; (i) if the probability ofoccurrence or the consequences of an accident or malfunction of equipment important to safetypreviously evaluated in the Safety Analysis Report (SAR) may be increased; or (ii) if a possibilityof an accident or malfunction of a different type than any previously evaluated in the SAR may becreated; or (iii) if the margin of safety as defined in the basis for any Technical Specification isreduced.

In accordance with 10 CFR 50.59, the following evaluation serves to determine whether thismodification constitutes an unreviewed safety question:

1 ~ Does the proposed change increase the probability of occurrence of an accident previouslyevaluated in the SAR?

This modification affects the St. Lucie Unit 2 ADVs. The malfunction of the ADVs isaddressed in the Safety Analysis provided in chapter 15 of the FSAR. Subsection 15.1.3.1summarizes the opening of a single ADVas an initiating event. Subsection 15.3.5.1.4 alsoconsiders the failure of a single ADVto close following automatic actuation during a reactorcoolant pump seized shaft event with loss of offsite power and technical specification steamgenerator tube leakage. The modifications performed by this Engineering Package serveonly to reduce the total thrust transmitted from the actuator to the valve after the torqueswitch has tripped the motor in the closed direction. This modification does not affect theactuator thrust developed to operate the valve, and therefore, does not alter the ability ofthe valve to operate when required. The modification only reduces the forces imparted onthe ADV, and thus, enhances the ability of the ADVs to perform as intended duringemergency and offnormal conditions under maximum differential pressures.

The installation of compensating spring packs in the ADVactuators only serve to enhance .the operational characteristics of the ADVs. As such, no new accident initiating events arecreated. Therefore, the modifications described in this Engineering Package do notincrease the probability of valve failure, and thus the probability of occurrence of anaccident previously described in the SAR is not increased by this modification.


This modification does not affect any structures, systems or components that function todeter the release of radioactivity or to provide post-accident shielding. The modificationsperformed by this Engineering Package do affect the ADVs, which may be relied upon tocooldown the plant. However, this function is not adversely affected by installation of thecompensating spring packs, since the actuator capacity has not been reduced. Therefore,the consequences of an accident previously evaluated in the SAR are not increased by thismodification.




System operability is not being affected by the modifications to the ADVs identified in thisEngineering Package. Valve operability will not be adversely affected by the prescribedmodifications. In addition, no new failure modes are created as a result of this modification(see Failure Modes and Analysis of this EP, section 3.3). The addition of compensatingspring packs in the ADV actuators will reduce the total thrust on the ADV while notreducing the actuator capacity. As such, no new accident initiating events are created.Therefore, the probability of occurrence of a malfunction of equipment important to safetypreviously evaluated in the SAR is not increased by this modification.


System operation is not affected by this modification. This modification does not interactspatially or functionally with any structure, system or component important to safety otherthan the valves and valve operators themselves. The compensating spring packs havebeen selected in accordance with the same design criteria as the original components.The modifications performed by this Engineering Package do not adversely affect the abilityof the valves and valve operators to perform as intended during emergency and off-normalconditions under design conditions. Therefore, the consequences of a malfunction ofequipment important to safety previously evaluated in the SAR is not increased by thismodification.


As discussed previously, this modification does not change the function or design basesof any structure, system or component important to safety as described in the SAR. Thismodification installs compensating spring packs which in turn increase the reliability of thevalve/actuator combination by reducing stresses imparted on the valve. No new failuremodes or conditions are created that can be postulated to cause an accident different thanthose previously analyzed in the SAR. Therefore, the possibility of an accident of a differenttype than any previously evaluated in the SAR is not created by this modification.




This modification does not interact spatially or fun'ctionally with any structure, system orcomponent important to safety other than the valves and valve operators themselves. Thismodification does not alter the function or the design basis of the ADVs. As discussed inSection 3.3, no new failure modes are created for the subject MOVs that can be postulatedto cause a malfunction of equipment important to safety different than those previouslyanalyzed in the SAR. Therefore, the possibility of a malfunction of equipment important tosafety which is of a different type than any previously evaluated in the SAR is not createdby this modification.


The Technical Specification requirements and Technical Specification Bases are notaffected by this modification. The design bases of the valves and valve operators remainsunchanged. Therefore, the margin of safety as defined in the bases for any TechnicalSpecification is not reduced by this modification. "

The foregoing discussions provided in this EP constitutes, per 10 CFR 50.59(b), the written safetyevaluation which provides the bases that these modifications do not impact the safe operation ofthe plant, constitute an unreviewed safety question, or require a change to the plant TechnicalSpecifications. As such, prior NRC approval for the implementation of this PC/M is not required.


ABSTRACT

Engineering Evaluation (EE) JPN-PSL-SEMP-91-029, Rev. 0," Engineering Evaluation of ShutdownCooling System Transient Response", states air in the Containment Spray (CS) header is causing

. pressure transients in the Shutdown Cooling (SDC) piping when the Low Pressure Safety Injection(LPSI) pumps are operated. As shown in various design documents, the CS header has nomeans of being vented. The EE recommends that vent valves be installed on the CS headerupstream of the containment isolation valves. As valves 1-FCV-07-1A and 1-FCV-07-1B arenormally closed and isolate the containment, the above mentioned vents need to be located athigh'oints of the headers upstream of these isolation valves.

This Engineering Package (EP) provides the specific design information necessary to install one3/4" vent in each CS header immediately up stream of valves 1-FCV-07-1A and 1-FCV-07-1 B. Inaddition this EP revises plant drawings for administrative changes by correcting design andhydrotest conditions which are reflected on drawings JPN-418-291-002, 005 and 006.

The CS system performs a safety related function, as described in FSAR, Section 6.2.2. As such,this EP has been classified as Safety Related. This EP does not have any adverse impact onplant safety and/or operation. Based on a Failure Mode and Effects Analysis and a review of thechanges to be implemented by this EP against the requirements of 10CFR50.59, it was concludedthat these modifications do not constitute an unreviewed safety question and do not require achange to the plant Technical Specifications. Therefore, prior NRC approval for theimplementation of this modification is not required.

SAFETY EVALUATION

With respect to Title 10 of the Codes of Federal Regulations, Part 50.59, a proposed change shallbe deemed to involve an unreviewed safety questions: (i) if the probability of occurrence or theconsequences of an accident or malfunction of equipment important to safety previously evaluatedin the Safety Analysis Report may be increased, or (ii) ifa possibility for an accident or malfunctionof a different type than any evaluated previously in the Final Safety Analysis Report may becreated, or (iii) if the margin of safety as defined in the bases for any technical specification isreduced. The modification here in does not involve an unreviewed safety question because of thefollowing reasons:

Does the proposed activity increase the probability of occurrence of an accident previouslyevaluated in the SAR?

The proposed activity does not increase the probability of occurrence of an accidentpreviously evaluated because the operation and functionality of the CS System has notbeen changed by this modification. The vents are not used during power operation orinvolved in any way with any safety function of this system. Based on this, the probabilityof occurrence of an analyzed accident remains unchanged.




The proposed activity does not increase the consequences of an accident because thesevents meet all regulatory requirements specified in the FSAR. Their operating and pressureretaining characteristics are shown to be acceptable by the Design section of this EP. Inall operational modes they are a passive pressure boundary component. The vents do notincrease the radiological doses of an accident.

3. Does the proposed activity increase the probability of occurrence of a malfunction ofequipment important to safety previously evaluated in the SAR?

The proposed activity does not increase the probability of occurrence of a malfunction ofequipment important to safety because this system's function and performance remainunchanged by the addition of these vents. Furthermore these vents do not directly orindirectly affect equipment important to safety. These vents do not degrade the reliabilityor increase challenges, directly or indirectly, for equipment important to safety.

" Does the proposed activity increase the consequences of a malfunction of equipmentimportant to safety previously evaluated in the SAR?

The proposed activity does not increase the consequences of a malfunction of equipmentimportant to safety because the vents perform no active nuclear safety related orequipment protection function. The vent valves are administratively controlled to preventtheir being left open and are designed to be physically strong enough to preclude beingbroken off. In the extreme unlikely event that the vent fails, the worse-case scenario is theloss of one of the two redundant headers, which has been analyzed in the FSAR.Therefore, their operability will have no effect on the consequences of a malfunction ofequipment important to safety.


The proposed activity does not create the possibility of an accident of a different type thanany previously evaluated because the vents are not accident initiating devices. The ventvalves are only operated when filling the SDC system and serve no active or controllingfunction.




The proposed activity does not create the possibility of a malfunction of equipmentimportant to safety of a different type than any previously evaluated because the methodof operation of the CS system has not changed. The addition of the vents has beenanalyzed to maintain the seismic and pressure boundary integrity of the CS headers.


The proposed activity does not reduce the margin of safety as defined in the bases for anyTechnical Specification because T.S. Section 3/4.6.2 requires that both loops of the CS beoperable, all valves in the CS System be properly aligned and that the pumps and initiationsystem be tested. The Technical Specification bases ensure that adequate containmentcooling and depressurization capability exists during a LOCA and is not affected by this EP.

The foregoing constitutes, per 10 CFR 50.59 (b), the written safety evaluation which provides thebases that this change does not involve an unreviewed safety question nor a change to the PlantTechnical Specification. Prior NRC approval for the implementation of this modification is notrequired.


ABSTRACT

This Engineering Package provides for enhancement modifications to the Containment HydrogenAnalyzer System. The modifications which are designed to help ensure maintainability of theContainment Hydrogen Analyzer System and enhance its performance involve the following:

1) Reconfiguration of gas supply lines and associated components to allow for a continuoussupply of reagent gas (0,).

2) 'onversion of the present 3-wire analyzer cell configuration to a 4-wire analyzer cell designfor the purpose of mitigating temperature oscillations.

3) Addition of a volume chamber, located between regulator R3 and the suction of the samplepump, to reduce sample flow fluctuations.

4) Installation of a structural steel compressed gas cylinder storage rack for the purpose ofproperly storing hydrogen and oxygen supply gas cylinders.

5) Updates to the Vendor Technical Manual (VTM).

a)b)

c)

d)

Incorporate the vendor's latest (May 1991) spare parts replacement schedule.Correct the amp board schematic drawing with regard to the range (0-10% H,)selector switch.Correct the stabilization time period called for during the zero and span adjustmentsof the calibration procedure (due to a change in the catalyst bed) ~

Capture modifications associated with this Engineering Package.

Section 6.2.5 of the St. Lucie 2 FSAR describes combustible gas control in containment.Containment Hydrogen Analyzers (Section 6.2.5.2.1), Containment Hydrogen Recombiners(Section 6.2.5.2.2), and Containment Hydrogen Purge (Section 6.2.5.2.3) are systems that areused to limit the buildup of hydrogen in containment during a LOCA.

The Containment Hydrogen Analyzer System is used during a LOCA to monitor the hydrogenconcentration in containment, such that appropriate operator actions can be taken to ensure thatthe hydrogen flammability limit of 4% is not exceeded. The Containment Hydrogen AnalyzerSystem consists of two redundant trains. Each redundant train is physically separated, operatedindependently, and powered from an independent onsite power source. The piping to and fromthe containment is designed and fabricated in accordance with ASME Section III Class 2 and N-stamped. The hydrogen analyzer package supplied by the vendor is classified as a Class 1Einstrument. Based on the above this Engineering Package is classified as Safety Related.



A safety evaluation of these modifications has been performed in accordance with 10 CFR 50.59.This evaluation concludes that implementation does not involve an unreviewed safety question nora change to Technical Specifications. Additionally, it has no adverse effect on plant safety oroperation. Therefore, prior NRC approval is not required.

SAFETY EVALUATION

With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shallbe deemed to involve an unreviewed safety question: (i) if the probability of occurrence or theconsequences of an accident or malfunction of equipment important to safety previously evaluatedin the Safety Analysis Report may be increased, or (ii) if a possibility for an accident or malfunctionof a different type than any evaluated previously in the Safety Analysis Report may be created, or(iii) if the margin of safety as defined in the bases for any technical specification is reduced. Themodifications included in this Engineering Package do not involve an unreviewed safety questionbecause of the following reasons:

Does the proposed activity increase the probability of occurrence of an accident previouslyevaluated in the Safety Analysis Report (SAR)?

The proposed activity does not increase the probability of occurrence of an accidentbecause the Containment Hydrogen Analyzer System modifications do not introduceaccident initiating devices.

2. Does the proposed activity increase the consequences of an accident previously evaluatedin the SAR?

The proposed activity does not increase the consequences of an accident because theContainment Hydrogen Analyzer System modifications enhance performance of the systemand thus have a positive affect on mitigating the consequences of an accident. Thesemodifications will not cause an increase in radiation dose levels during an accident.

0


1

The propo'sed activity does not increase the probability of occurrence of a malfunction ofequipment important to safety because the Containment Hydrogen Analyzer Systemmodifications are designed to ensure that safety related equipment is not adverselyimpacted. The modifications were evaluated for seismic excitations as required to assurecontinued functioning of safety related equipment.




The proposed activity does not increase the consequences of a malfunction of equipmentimportant to safety because the Containment Hydrogen Analyzer System modifications aredesigned to ensure that safety related equipment is not adversely impacted.

Does the proposed activity create the possibility of an accident of a different type than anypreviously evaluated in the SAR?

The proposed activity does not create the possibility of an accident of a different type thanany previously evaluated because the Containment Hydrogen Analyzer Systemmodifications do not involve any accident initiating devices. This system informs operatorsof an abnormal plant condition and serves no controlling function.


The proposed activity does not create the possibility of a malfunction of equipmentimportant to safety of a different type than any previously evaluated because theContainment Hydrogen Analyzer System modifications are designed to ensure that safetyrelated equipment is not adversely impacted. All new components have been evaluatedfor seismic excitations to assure continued functioning of safety related equipment.

Does the proposed activity reduce the margin of safety as defined in the bases for anytechnical specification?

The proposed activity does not reduce the margin of safety as defined in the bases for anyTechnical Specification because the Containment Hydrogen Analyzer System modificationsdo not affect any safety margins as discussed in the bases of any Technical Specification.The associated Technical Specification basis is to ensure that the equipment and systemsrequired for the detection and control of hydrogen gas are available to maintain thehydrogen concentration within containment below its flammable limit during post-LOCAconditions. No margin of safety is affected by this modification.

The foregoing constitutes, per 10 CFR 50.59(b), that the modifications to the ContainmentHydrogen Analyzer System do not involve an unreviewed safety question nor a change tothe Plant Technical Specifications. Therefore, prior NRC approval for the implementationof these modifications is not required.


ABSTRACT

Since 1985, there have been several failures of the St Lucie Plant Unit 2 reactor coolant pumps(RCP) (2A1, 2B1, 2A2 and 2B2) upper and lower bearing oil reservoir level measurement systemwhich have resulted in several plant shutdowns. The existing RCP oil level measurement system(LT-1156, 1157, 1166, 1167, 1176, 1177, 1186 and 1187) utilizes capacitance probes withamplifiers which have not been effective or reliable in providing an accurate measurement ofreactor coolant pump oil level. This measurement system is being replaced with a "bubbler" typelevel measurement system. This new system willbe designed to interface with the remaining RCPoil level measuring instrumentation loop components (LIA-1156, 1157, 1166, 1167, 1176, 1177,1186 and 1187).

The bubbler system willconsist of a regulator, purgemeter, dip tube, transmitter, and excess flowvalve and will utilize plant instrument air from the Instrument Air supply ring header.

The installation of the bubbler involves instrument and tubing installation and piping and relatedpipe support modifications in the vicinity of safety related equipment in the Containment Building.Furthermore, since the installation interfaces with the RCP Lube Oil System, Appendix "R" FireProtection analysis is a concern. The installation of the bubbler systems is being seismicallymounted to prevent interaction with safety related equipment, and also designed to meet Appendix"R" Fire Protection requirements. Therefore, this PCM is designated as Quality Related.

The safety evaluation has shown that this EP does not constitute an unreviewed safety questionand prior NRC approval is not required for implementation. The implementation of this EP doesnot reduce the margin of safety for any Technical Specifications.

SAFETY EVALUATION

With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shallbe deemed to involve an unreviewed safety question: (i) if the probability of occurrence or theconsequences of an accident or malfunction of equipment important to safety previously evaluatedin the Safety Analysis Report (SAR) may be increased; or (ii) if the possibility for an accident ormalfunction of a different type than any evaluated previously in the Safety Analysis Report maybe created; or (iii) if the margin of safety as defined in the basis for any Technical Specificationsis reduced.

The modification included in this Engineering Package does not involve an unreviewed safetyquestion as demonstrated by the answers to the following questions:

01. Does the Proposed Activityincrease the probability of occurrence of an accident previously

evaluated in the Safety Analysis Report?



The modification does not increase the probability of occurrence of an accident previouslyevaluated. As mentioned in Section 3.2 of this EP, the RCP lube oil system does not havea detailed description in the FSAR, and has no safety related function. However, theinstallation of the bubbler system is being seismically mounted as to protect safety relatedequipment. Therefore, the proposed activity willnot increase the probability of occurrenceof an accident.

2. Does the proposed activity increase the consequences of an accident previously evaluatedin the Safety Analysis Report?

The modification does not increase the consequences of an accident previously evaluatedsince there is no interaction between the RCP oil measurement system and any safetyrelated equipment.

Does the proposed activity increase the probability of occurrence of a malfunction ofequipment important to safety previously evaluated in the Safety Analysis Report?

The modification does not increase the probability of occurrence of a malfunction ofequipment important to safety since there is no interaction with any safety relatedequipment and the modification only involves monitoring instrumentation.

4 Does the proposed activity increase the consequences of a malfunction of equipmentimportant to safety previously evaluated in the Safety Analysis Report?

The modification does not increase the consequences of a malfunction of equipmentimportant to safety because the RCP oil system and the monitoring of its level perform nosafety functions and has been designed not to interact with safety equipment during anaccident.

5. Does the proposed activity create the possibility of an accident of a different type than anypreviously evaluated in the Safety Analysis Report?

6.

The proposed change does not create the possibility of an accident of a different type thanany previously evaluated in the SAR because it does not add or delete interfaces withexisting important to safety structures, systems, or components. The failure modesanalysis performed indicates no new failure modes are created.




The proposed change does not create the possibility of a malfunction of equipmentimportant to safety of a different type than any previously evaluated in the SAR becausethe RCP lube oil system is a non safety related system and the added piping, componentsand modified pipe supports are designed to the same requirements of ANSI B31.1 1973Edition through winter 1973 as the existing RCP lube oil system piping and components.Moreover, Quality Related requirements have been applied to the proposed change by wayof independent verification of design in addition to enhanced quality level for procurementof components. The enhanced quality level provides for certification to the design codesfor the components.

7. Does the proposed activity reduce the margin of safety as defined in the Basis for anyTechnical Specifications?

The modification does not reduce the margin of safety as defined in the basis for anyTechnical Specifications because the Reactor Coolant Pump lube oil system is notspecifically mentioned in the Technical Specifications, therefore the installation of thebubbler level measurement system to the RCP lube oil system will not reduce the marginof safe to the basis for an Technical S ecifications.ty y p

The implementation of this PCM to install a new bubbler system to measure RCP reser voir oil leveldoes not require a change to the Plant Technical Specifications.

The foregoing constitutes, per 10CFR50.59(b), the written safety evaluation which provides thebases that this change does not involve an unreviewed safety question or a change to the PlantTechnical Specifications, and prior NRC approval for the implementation of this PCM is notrequired.


ABSTRACT

This Engineering Package provides for the replacement of two obsolete Fischer and Porter Model51-1401 indicating controllers installed in the St Lucie Plant - Unit No. 2 for component coolingwater temperature local indication and control. In the current plant configuration, temperatureindicating controllers TIC-14-4A and TIC-14-4B throttle temperature control valves TCV-14-4A andTCV-14-4B to regulate Intake Cooling Water flow depending upon Component Cooling Wateroutlet temperature from the Component Cooling Water Heat Exchangers (CCWHX) 2A and 2B,thus moderating CCW temperature. This EP will enhance the existing design by replacing theseobsolete controllers with new, currently available, pneumatic controllers which are qualified SeismicCategory I to provide additional conservatism in the design. Similarly, existing high limit relays andpressure regulators will be replaced with Seismic Category I qualified equivalent components.

The safety evaluation has shown that this EP does not constitute an unreviewed safety questionnor require a change to the Technical Specifications, therefore prior NRC approval is not requiredfor implementation.

Revision 1 was performed to provide clarifications in the safety evaluation.

SAFETY EVALUATION

With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shallbe deemed to involve an unreviewed safety question: (i) if the probability of occurrence or theconsequences of an accident or malfunction of equipment important to safety previously evaluatedin the Safety Analysis Report may be increased, or (ii) if a possibility for an accident or malfunctionof a different type than any evaluated previously in the Safety Analysis Report may be created, or(iii) if the margin of safety as defined in the bases for any technical specification is reduced.

The modification included in this Engineering Package does not involve an unreviewed safetyquestion as demonstrated by the answers to the following questions:

Does the Proposed Activity Increase the Probability of Occurrence of an AccidentPreviously Evaluated in the Safety Analysis Report?

This modification replaces obsolete Component Cooling Water (CCW) temperatureindicating controllers and associated instrumentation. The replacement components arelike-for-like replacements, perform the same functions as the existing equipment.Therefore, the proposed activity does not increase the probability of occurrence of, anaccident previously evaluated in the SAR.



2. Does the Proposed Activity Increase the Consequences of an Accident PreviouslyEvaluated in the Safety Analysis Report?

This modification replaces obsolete Component Cooling Water (CCW) temperatureindicating controllers and associated instrumentation. The replacement components arelike-for-like replacements, perform the same functions as the existing equipment.

The temperature indicating controllers (TIC-14-4A and TIC-14-4B), regulators and relaysinterface with existing instrument air supply, in-line thermowells, valve positioners andinstrument racks and supports. Physical and operational independence of the redundantSafety Channel A and Safety Channel B ICW trains and CCW trains is maintained such thatfailure of any component associated with a given train willhave no impact on the operabilityof the redundant train. Accordingly, the proposed change will not degrade the reliabilityof either the Intake Cooling Water System or the Component Cooling Water System, whichare important to safety. Therefore, the proposed activity does not increase theconsequences of an accident previously evaluated in the Safety Analysis Report.

3. Does the Proposed Activity Increase the Probability of Occurrence of a Malfunction ofEquipment Important to Safety Previously Evaluated in the Safety Analysis Report?

The proposed change does not involve any in-line instrumentation and is separated fromentrained fluids by a Seismic Category I (ASME Class 3) thermowell. Pneumatic pressureregulators and high limit relays are maintained in the design, and helical bourdon tubessimilar to those of the existing indicating controllers are employed in TIC-14-4A andTIQ-14-4B. The mounting of the new instrumentation, tubing and supports has beenseismically designed to provide additional conservatism in the design. Therefore, there isno increase in the probability of occurrence of a malfunction of equipment important tosafety previously evaluated in the Safety Analysis Report.

Does the Proposed Activity Increase the Consequences of a Malfunction of EquipmentImportant to Safety Previously Evaluated in the Safety Analysis Report?

This modification replaces obsolete Component Cooling Water (CCW) temperatureindicating controllers and associated instrumentation. The replacement components arelike-for-like replacements, perform the same functions as the existing equipment. Existingsystem interface points (thermowells, supply air, valve positioners, racks and supports) arecompatible with the new equipment and are not adversely impacted by this modification.Therefore, there is no increase to the consequences of a malfunction of equipmentimportant to safety previously evaluated in the Safety Analysis Report.


SAFETY EVALUATION(continued)r

5. Does the Proposed Activity Create the Possibility of an Accident of a Different Type thanany Previously Evaluated in the Safety Analysis Report?

This modification provides for the replacement of TCV-14-4A and TCV-14-4B pneumaticcontrols with Seismic Category I components. The replacement components arelike-for-like replacements and perform the same functions as the existing equipment. Thereplacement components have been qualified Seismic Category I to provide additionalconservatism in the design. Therefore, the possibility of an accident of a different type thanany previously evaluated in the SAR is not created by the changes proposed herein.

6. Does the Proposed ActivityCreate the Possibility of a Malfunction of Equipment Importantto Safety of a Different Type Than Any Previously Evaluated in the SAR?

The proposed change does not create the possibility of a malfunction of equipmentimportant to safety of a different type than any previously evaluated in the SAR since theexisting CCW operating characteristics are maintained by utilizing the same pneumaticcontrol scheme as currently exists for TCV-14-4A and TCV-14-4B.

7. Does the Proposed Activity Reduce the Margin of Safety as Defined in the Basis for anyTechnical Specification?

The proposed change does not reduce the margin of safety as defined in the basis for anyTechnical Specification because the new controllers, air regulators, relays and associatedhardware satisfy the Plant Technical Specifications requirements for operation andmaintenance of the CCW system and the CCW system operating characteristics aremaintained. Plant Technical Specifications are not impacted by this modification.

The foregoing constitutes, per 10CFR50.59(b), the written safety evaluation which provides thebasis that this change does not involve an unreviewed safety question and does not require achange to the Plant Technical Specifications. Therefore, prior NRC approval for theimplementation of this PCM is not required.


ABSTRACT

This Engineering Package (EP) includes the engineering and design necessary to add dischargeresistors across the shunt field of 125 VDC motor operated valves (MOVs) at St..Lucie Unit 2.The addition of discharge resistors is based on recommendations from the D.C. MOV DesignInadequacies Study, EBASCO report which states that when MOVs are energized andde-energized, this action results in high voltage transients which can cause shunt coil insulationdamage and consequent premature shunt coil failure. This concern is addressed in NRCInformation Notice 88-72. The MOVs that are affected are:

Ta Number MOV Descri tion

MV-08-12MV-08-13MV-08-14MV-08-15MV-08-16MV-08-17MV-09-11MV-09-12

Steam Generator 2B to AFWP 2C TurbineSteam Generator 2A to AFWP 2C TurbineAtmospheric Steam Dump Isolation ValveAtmospheric Steam Dump Isolation ValveAtmospheric Steam Dump Isolation ValveAtmospheric Steam Dump Isolation ValveAFWP 2C Discharge to Steam Generator 2AAFWP 2C Discharge to Steam Generator 2B

This EP involves modification of Nuclear Safety Related MOVs, and is therefore classified asNuclear Safety Related.

The safety evaluation of this EP has determined that this plant change modification (PC/M) doesnot constitute an unreviewed safety question as defined in 10 CFR 50.59 and does not require achange in the Plant Technical Specifications. This PC/M has no adverse impact on plant safetyor operation. Therefore, this PC/M can be implemented without prior NRC approval ~

SAFETY EVALUATION

With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shallbe deemed to involve an unreviewed safety question: (i) if the probability of occurrence or theconsequences of an accident or malfunction of equipment important to safety previously evaluatedin the Safety Analysis Report may be increased, or (ii) ifa possibility for an accident or malfunctionof a different type than any evaluated previously in the Safety Analysis Report may be created, or(iii) if the margin of safety as defined in the bases for any Technical Specification is reduced. Inaccordance with 10 CFR 50.59, the following evaluation serves to determine whether thismodification constitutes an unreviewed safety question:

0Does the proposed activity increase the probability of occurrence of an accident previouslyevaluated in the Safety Analysis Report (SAR)?



The malfunction of MOVs modified by this PC/M does not adversely affect any equipmentwhose malfunction is postulated in the SAR to initiate or exacerbate an accident.Accordingly, a malfunction of the discharge resistor will result in an open circuit across theMOV shunt field winding, which in turn will be electrically equivalent to the existing MOVs.Therefore, neither the function nor the operation of the MOVs are changed due to theaddition of a discharge resistor as per this PC/M. The ADV isolation valves are normallylocked open. Accordingly, the ability to provide feedwater for the removal of decay heatfrom the reactor coolant system during normal and natural circulation system cooldown aswell as, the ability to provide sufficient feedwater capacity to permit plant cooldown,assuming a single active failure and loss of off-site power is not affected by this PC/Msince the discharge resistor electrically drops out of the MOV electrical circuit very quicklyin comparison to the time it takes to open or close a MOV. This modification does notcircumvent the valve safety function. This modification does not affect MOV pressureboundaries. Therefore, the probability of occurrence of an accident previously describedin the SAR is not increased by this modification.


This modification does not affect the function or operation of the MOVs, nor does it affectother systems, and components that are relied upon to mitigate accidents. Therefore, theconsequences of an accident previously evaluated in the SAR is not increased by thismodification.


This modification adds discharge resistors across the shunt field winding of each 125 VDCMOV to protect against potential high voltage transients that may occur duringde-energization of a MOV. However, a malfunction of the discharge resistor will result inan open circuit across the IVIOV shunt field winding, which in turn will be electricallyequivalent to the existing MOVs. The addition of the discharge resistors is considered anenhancement and therefore reliability of the MOV motors will be increased by this change.Therefore, the probability of occurrence of a malfunction of equipment important to safetypreviously evaluated in the SAR is not increased by this PC/M.


The addition of the discharge resistors will not alter the original safety function of theMOVs. The addition will not affect assumed malfunctions of safety equipment describedin the SAR. The discharge resistors have an open circuit failure mode. An open circuit



discharge resistor across a 125 VDC MOV shunt field winding is electrically equivalent tothe existing motor circuit. The addition of discharge resistors does not adversely affect theperformance of the MOV in a mitigating activity. Therefore, the implementation of this EPdoes not increase the consequences of a malfunction of equipment important to safetypreviously evaluated in the SAR.


~

This modification adds protection against high voltage transient that may cause motorinsulation degradation and reducedmotor life. Protection is added by installing a dischargeresistor across each DC MOV shunt field winding. This change does not alter the designbases of the AFWS, the Main Steam System, function or operation of a MOV, or create anynew failure mode hazards or conditions that can be postulated to cause an accidentdifferent than those previously analyzed in the SAR. This change does not result in anychange to component parameters or introduce any system interface. Therefore, theproposed activity does not create the possibility that an accident may be created that isdifferent from those previously evaluated in the SAR.


The modification does not create the possibility of a malfunction of equipment importantto safety of a different type than any previously evaluated. The credible failure mode of theresistor is an open circuit. Therefore, the modified MOV circuit with a failed dischargeresistor is identical to existing MOV circuit. The addition of the resistor does not create anew failure mode of the modified MOVs.

t

Does the proposed activity reduce the margin of safety as defined in the bases for anyTechnical Specification?

e

The modification does not reduce the margin of safety as defined in the bases for anyTechnical Specification. This modification adds a discharge resistor across the 125 VDCMOV shunt field winding. The operation of MOVs modified by this EP is functionallyequivalent to the existing MOVs. The purpose of the discharge resistor is to protect 125VDC MOVs from potential motor damage that can be caused by high voltage transientscreated during de-energization. The 125 VDC system calculations were reviewed todetermine that this increased loading has a negligible affect on the 125 VDC System. Theaddition of a discharge resistor does not adversely affect the MOVs performance. Theexisting margin of safety as defined in the basis for any Technical Specification remainsunchanged after the implementation of this modification.



The forgoing constitutes, per 10 CFR 50.59(b), the written safety evaluation which provides thebases that this change does not involve an Unreviewed Safety Question and prior Nuclear

'egulatory Commission approval for the implementation of this PC/M is not required.


ABSTRACT

This engineering package (EP) provides for the safety evaluation of the St. Lucie Unit 2 Cycle 6reload design. The Cycle 6 energy requirement is 12000 EFPH based on a Cycle 5 length of11500 EFPH. The replacement of approximately one third of the fuel assemblies with fresh onesis required to operate to the end of cycle energy.

The primary change to the core for Cycle 6 is the replacement of 76 irradiated assemblies with76 fresh Region H fuel assemblies. The fuel is arranged in a low leakage pattern with nosignificant differences between the Cycle 6 loading pattern and the Cycle 5 design. Themechanical design of Region H is similar to the mechanical design of Region 8 (Cycle 5).

The reload designs are classified as Safety Related since they must provide for the safe shutdownof the reactor. The design analysis associated with this reload design evaluates plant operatingparameters that are associated with the capability to shutdown the reactor and maintain it in asafely shutdown condition.

Based upon the design analysis by Combustion Engineering, which was independently verifiedby FPL, it was determined that the Cycle 6 operation is bounded by the results in this reference,and can be implemented with no changes required to the existing St. Lucie Unit 2 TechnicalSpecifications. Therefore, prior NRC approval is not required for implementation.

The implementation of this EP will not adversely impact plant safety or operation.

~SI 2

The purpose of this revision is to reflect the following changes and additions. a report whichprovides recommendations to the St. Lucie Unit 2 plant staff concerning excore detector electronicadjustments, was changed. An additional reference, which verifies that the assumptions utilizedin the setpoint analysis are valid for cycle 6 operation was added. The changes requested in aCRN were also addressed.

These revisions do not require revision to the plant Technical Specifications, nor do they affectthe Safety Evaluation provided in Section 3.0 of this EP. Therefore, pursuant to IOCFR50.59 thismodification can be implemented without prior commission approval. These revisions do not haveany adverse impact on plant safety and operation.

SAFETY EVALUATlON

Based on the technical evaluation and the results of the reanalysis included in the Reload SafetyEvaluation report, it is concluded that the Cycle 6 reload design meets all design criteria, isbounded by the results of the referenced analyses, and can be implemented with no changesrequired to the existing St. Lucie Unit 2 Technical Specifications. Therefore, it can be stated that:



The probability of occurrence or the consequences of an accident or malfunction ofequipment important to safety previously evaluated in the safety analysis report is notincreased.

The St. Lucie Unit 2 Cycle 6 reload design does not change the overall configuration of theplant. The mode of operation of the plant remains unchanged. Therefore, the probabilityof occurrence of an accident or malfunction, previously evaluated in the safety analysisreport, is not increased. The RSE report demonstrates that the consequences of anaccident or malfunction have not been increased beyond these evaluated in the previousanalyses since all the transients meet current criteria.

ii~ A possibility for an accident or malfunction of a different type than any evaluated previouslyin the safety analysis is not created.

The St. Lucie Unit 2 Cycle 6 reload design does not change the overall configuration of theplant. The mode of operation of the plant remains unchanged. Therefore, a possibility fora new accident or equipment malfunction has not been created.

The margin of safety as defined in the basis for any Technical Specification is not reduced.

FPL performed an evaluation and review of the St. Lucie Unit 2 Chapter 15 events to verifythat the inputs to the safety analyses and the results are bounding for Cycle 6 applications.Based on this evaluation it was determined that inputs to all Cycle 6 events were boundedexcept the following; the Increased Main Steam Flow, the Pre-Trip Steam Line Break pincensus, the Uncontrolled Control Element Assembly Withdrawal from Subcritical or LowPower Condition, and the Small Break LOCA event. The results of these reanalyses givenin the RSE report show that in each case, the respective reference analysis bound theCycle 6 specific results. Therefore, there is no reduction in the margin of safety relative tothe Technical Specification basis.

As per Federal Regulation 10 CFR 50.59 this activity does not involve an unreviewed safetyquestion and does not require a change to Technical Specifications, therefore, implementation ofthis reload is permissible without prior NRC approval.


ABSTRACT

This Engineering Package (EP) provides for the removal of the identified thermal relief valves,associated piping segments, pipe insulation, heat tracing cables and affected pipe supports fromthe Chemical 8 Volume Control System (CVCS). The affected piping sections extend from thedischarge of the Boric Acid Makeup Tanks (BAMTs) to the Charging Pumps and the VolumeControl Tank (VCT).

The deletion of the aforementioned components is based on Combustion Engineering's evaluationCEN-365(L), titled "Boric Acid Concentration Reduction Effort", which resulted in a reduction ofthe boric acid concentration thus eliminating the need for the heat tracing as well as the thermalrelief valves. The CE report was undertaken to reflect relatively recent advances in themethodology for setting BAMTconcentration and levels vs the past methodology.

The portion of the CVCS System pertaining to this modification as defined by FSAR subsection9.3.4, is classified as Safety Class 2 and Seismic Category I. Therefore, this PCM will be classifiedas Nuclear Safety Related.

The safety evaluation of this EP has shown that the implementation of this PCM does notconstitute an Unreviewed Safety Question and requires no revision to the Unit 2 TechnicalSpecifications. Prior NRC approval is not required for implementation. This PCM has no impacton plant safety and operation, or the Plant Technical Specifications.

SAFETY EVALUATION

With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shallbe deemed to involve an unreviewed safety question: (i) if the probability of occurrence or theconsequences of an accident or malfunction of equipment important to safety previously evaluatedin the Safety Analysis Report may be increased; or (ii) if the possibility for an accident ormalfunction of a different type than any evaluated previously in the Safety Analysis Report maybe created; or (iii) if the margin of safety as defined in the basis for any technical specification isreduced.

This modification removes affected thermal relief valves, their upstream/downstream piping, heattracing cables and the thermal insulation from the CVCS System piping between the dischargeof BAM Tanks to the Charging Pumps and the Volume Control Tank.

The portion of the CVCS System, pertaining to this modification, as defined in FSAR Subsection9.3.4, is classified as Safety Class 2 and Seismic Category I. Therefore, this PCM is classified asNuclear Safety Related.

i) The probability of occurrence or the consequences of an accident or malfunction ofequipment important to safety previously evaluated have not been increased.



The previous related accident evaluated in the FSAR, Table 9.3-9 includes a failure of theboric acid line heat tracing. PCM 283-288 addresses this issue, based on a CombustionEngineering (CE) report. Accordingly, the concentration of the boric acid in the subjectlines has been reduced from 12% by weight to 2.5-3.5% by weight. Per the CE Report andFSAR Section 9.3.4, the ambient temperature in the reactor auxiliary building will besufficient to prevent any precipitation within the Boric Acid Makeup System, therebyjustifying the deletion of the heat tracing on the subject lines.

Since the heat tracing of the subject lines is no longer required, the thermaloverpressurization relief valves, will no longer be required. Thus, the required flow ofborated water for the RCS System will be maintained. Therefore the probability ofoccurrence or consequences of an accident or malfunction does not increase as a resultof this modification since the probability of failure of BA Makeup System (Subsystem ofCVCS System) has not increased and the subject modifications do not adversely impactthe operability of any other safety related functions.

The possibility for an accident or malfunction of a different type other than any evaluatedpreviously in the safety analysis report is not created.

The modification proposed herein does not impact the process flow of the borated waterto the RCS System and does not create the possibility for an accident or malfunction of adifferent type than any previously evaluated in the FSAR. Also, no new active componentsare added by this modification which could adversely impact other safety related equipmentor functions. Therefore, malfunction different than those previously evaluated in the FSARhas not been created. Deletion of the thermal relief valves and the associated piping willnot impact any other safety related equipment or functions.

The margin of safety as defined in the bases for any technical specification has not beenreduced.

This modification has no adverse impact on the operability of the affected Reactivity ControlSystem as addressed in the Section 3/4.1 of the Technical Specification. No changes tothe Technical Specification are required by this modification.

The margin of safety as defined in the Technical Specification is not affected by this EP.

The implementation of this PCM does not require a change to the Plant TechnicalSpecifications.



The foregoing constitutes, per 10CFR50.59(b), the written Safety Evaluation which provides thebases that this change does not involve an unreviewed safety question and prior Nuclear

. Regulatory Commission (NRC) approval for the implementation of this PCM is not required.


ABSTRACT

This Engineering Package provides the engineering and design details required to implement thereplacement of the existing ionization smoke and duct detectors and the existing Main Fire Alarmpanels. The smoke detectors and panels are part of the fire detection system.

The existing detectors are divided into two groups: The originals (installed 9 years ago) which areobsolete; and their replacements (installed as the originals failed) which are no longermanufactured. To ensure the reliability of the fire detection system, new ionization smokedetectors will be installed.

The existing panels are obsolete and spare parts are no longer available. The replacement panelsrepresent the latest evolution in Honeywell's Fire Detection System's hardware and software. Thenew detectors and panels are compatible with the existing plant fire detection system computer.

The fire detection system, which is part of the fire protection system, is non-safety related, but isprovided in areas that contain or present a fire hazard to equipment essential to safe plantshutdown. Therefore, this Engineering Package (EP) is classified as Quality Related.

This EP was reviewed in accordance with 10CFR50.59 and was found not to constitute anunreviewed safety question. The modifications described above have no adverse impact on plantoperations or safety and do not require a change to the plant Technical Specifications. Therefore,prior NRC approval is not required for the implementation of this EP.

SAFETY EVALUATlON

With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shallbe deemed to involve an unreviewed safety question: (i) if the probability of occurrence orconsequences of an accident or malfunction of equipment important to safety previously evaluatedin the Safety Analysis Report may be increased; (ii) if a possibility for an accident or malfunctionof a different type than any evaluated in the Safety Analysis Report may be created; or (iii) if themargin of safety as defined in the bases for any Technical Specification is reduced.

This EP provides the engineering and design details required to replace the existing ionizationsmoke and duct detectors and Main Fire Alarm panels with new equipment. The existingequipment is either obsolete or no longer manufactured. The new detectors and panels arecompatible with the existing plant fire detection system computer.

The implementation of this EP will improve the reliability of the fire detection system by replacingobsolete equipment. This ensures the availability of the individual detectors to detect a fire andthat spare parts will be obtainable in case of equipment failure.



Fire detection systems are provided in areas that contain or present a fire exposure to equipmentessential to safe plant shutdown. Therefore, this EP has been classified as Quality Related.

Based on the preceding, the following conclusions can be made:

i) The probability of occurrence or the consequences of an accident or malfunction ofequipment important to safety previously evaluated in the Safety Analysis Report is not

'ncreased by these modifications. The replacement of the obsolete detectors and panelsenhances the operability of the equipment and the fire detection system. The newdetectors and panels have the same characteristics as the existing equipment. Thepossible failure of this equipment willnot prevent safety related equipment from performingtheir intended functions. The detectors and panels are not required during an accidentcondition. Therefore, the implementation of these modifications cannot increase theprobability of occurrence or the consequences of an accident or malfunction of equipment.

As a result of this modification, there is,no possibility for an accident or malfunction of adifferent type other than any previously evaluated. The detectors and panels are notrequired during an accident condition nor will they prevent safety related equipment fromperforming their intended functions. This modification does not affect any safety relatedequipment.

The margin of safety as defined in the bases for any Technical Specification is not reducedby this modification. The functions of the fire detection system that are controlled by theapplicable Technical Specifications 3/4 3.3.3.7 are maintained by this change.

The implementation of this PC/M does not require a change to plant TechnicalSpecifications.

The foregoing constitutes, per 10CFR50.59(b), the written safety evaluation which provides thebases that this change does not involve an unreviewed safety question and prior NRC approvalfor the implementation of this PC/M is not required.

ACCELERATED DISTRIBUTION TIONtrip solenoid valves, 20-1/OPC and 20-2/OPC, and associated circuits....

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