ML17309A712

Rept of Changes Made to Facility Under Provisions of 10CFR50.59 for Period 911007-930325. W/930922 Ltr

ML17309A712
Person / Time
Site:	Saint Lucie
Issue date:	03/25/1993
From:	Sager D FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L-93-246, NUDOCS 9309280400
Download: ML17309A712 (123)

Text

ACCELERATED DISTRIBUTION DEMONS TION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9309280400 DOC.DATE: NOTARIZED: NO DOCKET FACIL:50-389 St. Lucie Plant, Unit 2, Florida Power & Light Co. 05000389 AUTH. NAME AUTHOR AFFILIATION SAGER,D.A. Florida Power 6 Light Co.

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

"St Lucie Unit 2 Rept of Changes Made to Facility Provisions of 10CFR50.59 for Period 911'007-930325.

U r W/930922 ltr. D DISTRIBUTION CODE: IE47D COPIES RECEIVED:LTR TITLE: 50.59 Annual Report of Changes, Tests L ENCL / SIZE:

or Experiments Made 7

W/out Approv 8 NOTES: /

RECIPIENT COPIES RECIPIENT COPIES A ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-2 PD 1 0 NORRIS,J 2 2 D INTERNAL: ACRS 6 6 AEOD/DOA 1 1 AEOD/DS P/ROAB 1 '

~AEOD DSP/TPAB 1 1 D NRR/DRCH/HHFB 1 1 REG - E 02 1 1 RGN2 FILE 01 1 1 EXTERNAL: NRC PDR 1 1 NSIC D

D NOTE TO ALL "RIDS" RECIPIENTS PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM P 1-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS 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'-93-246 10 CFR 50.59 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gentlemen:

Re: St. Lucie Unit 2 Docket No. 50-389 Re ort of 10 CFR 50.59 Plant Chan es Pursuant to 10 CFR 50.59 ('b)(2), the enclosed report contains a brief description and summary of the safety evaluation of Plant Changes/Modifications (PCMs) which were made, and are reportable pursuant to 10 CFR 50.59. Included with the brief description of each PCM is a summary of the safety evaluation completed by Florida Power & Light (FPL) for that PCM. This report includes PCMs completed for the period of October 7, 1991 to March 25, 1993, and correlates with the information included in Amendment 8 of the Updated Final Safety Analysis Report submitted under separate cover.

Should there be any questions on this information, please contact us ~

Very truly yours, D. A. ger Vice esident St. Lu ie Plant DAS/CDW/kw Enclosure cc: Stewart D. Ebneter, Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, St. Lucie Plant 1

DAS/PSL 75994-93 A

9309280400 930325 PDR ADQCK 05000389 ~4 R PDR an FPL Group company

RE: St. Lucie Plant Docket No. 50-389 10 CFR 50.59 Report St. Lucie Unit 2 Report of Changes Made to the Facility Under the Provisions of 10 CFR 50.59 for the period October 7, 1991 to March 25, 1993 NOTE: The safety evaluations in this report are chronologically arranged starting with those created more recently. Please note that the level of detail of safety evaluations from earlier years do not reflect current practices.

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Plant Change/Modifications reportable pursuant to 10 CFR 50.59 for St. Lucie Unit 2 Number Su lement Title

'52-292 0 Turbine Trip Controls Modification 146-292 0-1 Fire Barrier Drawing Enhancements 140-292 0 CCW Heat Exchanger Temperature Control Valve Minimum Stop Setpoint 120-292 Fixed Incore Detector Replacement and Movable Incore Detector Assembly Modification 097-992 0 South Craft Lunchroom Facility 067-292 0-3 Fuel Reload for Cycle 7 053-292 0-1 Main Generator Inadvertent Energization Protection 023-292 Condenser Air Evacuation System Header Separation 563-291 Instrument Inverter Trip and Alarm Circuit Improvements 543-291 Deenergizing of Waste Management Heat Trace Circuits 510-291 0-1 NRC Generic Letter 89-10 MOV Thrust Values 500-291 0 Spent Fuel Pool Visual Level Indicator 486-291 0 Removal of Turbine Runback 421-291 0 Emergency Diesel Generator Shaft Coupling Guards 419-291 0-1 Atmospheric Dump Valves Actuator Modification 418-291 0 Containment Spray Vent Valve Installation 309-291 0 Containment Hydrogen Analyzer System Enhancements 247-291 Replacement of RCP Upper & Lower Oil Reservoir Level Measurement System 092-291 0-1 Fisher & Porter Indicating Controllers Replacement 091-291 0 125 VDC Arc Suppression Addition 256-290 1-2 Fuel Reload for Cycle 6 176-290 0 Boric Acid Concentration Reduction Modification 311-289 0 Obsolete Smoke Detectors Replacement

PC/M 152-292 Supplement 0 ABSTRACT This Engineering Package (EP) implements portions of the recommendations of Westinghouse Customer Advisory Letter 92-02, 'Operation, Maintenance, Testing of, and

'ystem Enhancements to Turbine Overspeed Protection System', as addressed in FPL Interoffice Correspondence 'Westinghouse CAL 92-02 Overspeed Protection System'.

The Customer Advisory Letter was issued in response to the destructive turbine-generator overspeed 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 stop trip solenoid valve, 20/AST, emergency trip solenoid valve, 20/ET, overspeed protection trip solenoid valves, 20-1/OPC and 20-2/OPC, and associated circuits. This will provide indication 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 electrical actuation 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 to accomplish manual or automatic turbine trip.

A seal-in circuit will be installed for the coil of the emergency trip solenoid valve, 20/ET.

This will cause the emergency trip solenoid to remain energized following actuation by the turbine trip push button and will ensure that the turbine remains tripped.

Surge suppression diodes will be 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, this EP has been classified as Quality Related.

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

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

PC/M 152-292 Supplement 0 SAFETY EVALUATION As defined in 10 CFR 50.59, an unreviewed safety question exists; (i) if the probability of occurrence 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 previously evaluated in the SAR may be created; or (iii) if the margin of safety as defined in the basis for any Technical Specification is reduced.

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

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

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

Although the proposed changes result in a slight increase in the probability of an initiating event, turbine trip due to circuit failure, the changes also provide compensating effects through the addition of turbine control system test and monitoring functions which enhance the capability to assess the operability of a protective system prior to the system being required to operate, and through modification of control operation to provide assurance that a manual turbine trip will be maintained. In accordance with 'Nuclear Engineering Department Guidance for Performing-10CFR50.59 Safety Evaluations', Revision 0, Section 4.7, the proposed changes therefore do not increase the probability of occurrence of an accident previously evaluated as defined in 10CFR50.59.

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

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

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

PC/M 152-292 Supplement 0 SAFETY EVALUATION (continued)

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

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

No failure of the turbine control system alters the designed operation of any safety related equipment. Therefore, the consequences of malfunction of equipment important to safety previously 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 any previously evaluated in the SAR?

This modification does not change the design bases of any structure, system, or component important to safety. No new failure modes or conditions are created that can be 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 evaluated in the SAR is not created by this modification.

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

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

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

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

PC/M 152-292 Supplement 0 SAFETY EVALUATION (continued)

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

The foregoing discussion constitutes, per 10CFR50.59(b), the written safety evaluation which provides 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 Technical Specifications. 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 2 Fire 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 and resulted in a finding which indicated that current plant configurations were not completely reflected on the fire barrier drawings. Therefore, revisions to enhance these drawings and the creation of additional drawings is necessary.

The,'intent of this EP is to enhance the drawings of existing fire barriers and assign a Fire Area/Fire Zone designation to one stairwell inside the Reactor Auxiliary Building (RAB). A Safe Shutdown Analysis (SSA) has been performed to demonstrate that fire protection equipment required to protect Safety Related equipment, or required to maintain the integrity of a fire barrier necessary to protect Safety Related equipment has not been adversely affected. Accordingly, this EP has been classified as Quality Related. A safety evaluation has beeri performed in accordance with 10 CFR 50.59 and is documented in Section 3.0 of this EP. This evaluation demonstrates that implementation of this modification does not involve an unreviewed safety question and does not require a change to the Plant Technical Specifications. In addition, this modification has no detrimental effects on plant safety or operation. Based upon the above, prior NRC approval is not required for the implementation of this modification.

Su lement 1 Supplement 1 to this EP deletes references to the fire protection related Technical Specifications which 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 shall be deemed to involve an unreviewed safety question: (i) if the probability of occurrence 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 the possibility for an accident or malfunction of 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 basis for any technical specification is reduced.

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

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

PC/M 146-292 Supplement 0-1 Qe SAFETY EVALUATION (continued)

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

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

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

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

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

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

The consequences of a malfunction of equipment important to safety are not affected because no equipment is affected. The EP merely changes the drawings to reflect the present plant configuration. The addition of the stairwells as fire zones will add new responses to the SSA; however, all these cables already appear in the zones adjacent to the 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 than previously evaluated in the SAR?

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

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

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

PC/M 146-292 Supplement 0-1 II SAFETY EVALUATION (continued)

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

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

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

PC/M 140-292 Supplement 0 ABSTRACT This Engineering Package (EP) allows setting the minimum open stop limit on the pneumatic controllers 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 2B respectively by maintaining the CCW outlet temperature of the heat exchanger at its required setpoint. The ICW system serves as a heat sink for the Component Cooling Water System, the Turbine Cooling Water System and the Steam Generator Blowdown System. The ICW System piping 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 Cooling Water (TCW) heat exchangers during normal plant operations. Increased intake cooling water flow to the TCW heat exchangers is desired during peak summer periods due to increased fouling of these heat exchangers and high intake cooling water temperatures. Since the CCW heat exchangers are sized for Design Basis Accident (DBA) conditions they are significantly oversized for 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 operations the CCW outlet temperature setpoint is maintained by the controllers which throttle the valves.

The excess ICW flow will be diverted to the TCW heat exchangers. This modification only affects the valves'inimum open position during normal plant operations and does not prevent the valve from 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 loading sequence. This analysis demonstrates that this modification will not adversely affect the components of the ICW system and is acceptable. In addition, precautionary measures are taken in this EP to preclude valve degradation from potential cavitation due to long term operation at a minimum open position as low as 8 percent.

This modification affects the waterhammer analysis of the Safety Related piping of the ICW system, 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 an adverse effect on plant safety, security or operation, does not constitute an unreviewed safety question and does not require a change to Plant Technical Specifications. Therefore, prior NRC approval for implementation of this modification is not required.

0 PC/M 140-292 Supplement 0 SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question: (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased, or (ii) if a possibility for an accident or malfunction of 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 safety question because of the following reasons:

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

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

e

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

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

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

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

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

The proposed activity does not increase the consequences of a malfunction of equipment important to safety because the ICW and CCW system remain capable of performing their Safety Related function during a DBA. No new failure modes are introduced by this modification.

PC/M 140-292 Supplement 0 SAFETY EVALUATION (continued)

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

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

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

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

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

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

The foregoing discussions provided in Sections 2 and 3 of this EP constitutes, per 10 CFR 50.59(b), the written safety evaluation which provides 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 Technical Specifications. As such, prior NRC approval for the implementation of this PC/M is not required.

PC/M 120-292 Supplement 0 ABSTRACT This Engineering Package (EP) provides the engineering justification and design details necessary to allow use of Fixed Incore Detectors of the "Active Tail" design. This engineering package involves the following tasks:

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

2) 'he Instrument Assembly background detector correction factors and sensitivities will be changed 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 for detector locations where the Active Tail design detector assembly is installed.

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

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

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 involve an unreviewed safety question nor a change to Plant Technical Specifications and has no detrimental effect on plant safety or operation. Therefore, prior NRC approval for implementation of this modification is not required.

PC/M 120-292 Supplement 0 SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question: (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased, or (ii) if a possibility for an accident or malfunction of 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 replacement of the Fixed Incore Detector Assemblies with either Passive Tail or Active Tail design does not involve an unreviewed safety question'because of the following reasons:

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

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

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

The proposed activity does not increase the consequences of an accident since the use of either Passive Tail or Active Tail detector assemblies will not change, degrade or prevent actions described or assumed for any accident as discussed in the SAR. The Incore Instrumentation System does not perform any safety related function and is not required for safe shutdown or to mitigate the consequences of an accident. In addition, the replacement of the Fixed Incore Detector Assemblies does not alter any assumptions previously made in evaluating the radiological consequences of an accident described in the FSAR.

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

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

PC/M 120-292 Supplement 0 SAFETY EVALUATION (continued) and no new components or equipment are introduced that could interact with any equipment important to safety. In addition, the CETs which are housed in the new detector assemblies are not affected by this changes to the detector assemblies. The replacement detector assemblies meet all seismic requirements as well as environmental qualification requirements.

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

The proposed activity does not increase the consequences of a malfunction of equipment important to safety because the use of either Passive Tail design or Active Tail design detector assemblies meets the required design intended for the Incore Instrumentation System. In addition, the Incore Instrumentation System is a non-safety related system and is not used to prevent or mitigate the consequences of an accident. The replacement ICI detectors 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 than previously evaluated in the SAR?

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

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

PC/M 120-292 Supplement 0 po SAFETY EVALUATION (continued)

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

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

The proposed activity does not reduce the margin of safety as defined in the basis for any Technical Specification, since the use of either Passive Tail design or Active Tail design detector assemblies does not change the required design functions of the Incore Instrumentation System or the Core Exit Thermocouples as described in Technical Specifications 3.3.3.2, 3.3.3.6. However, the Active Tail design detector assembly does not use the interconnecting tubing for the Moveable Incore Detection System and the installation of this type of detector requires that the interconnecting tubing be removed.

The removal of the interconnecting tubing will decrease the number of detector locations that the MICDS can position its probe in. This is acceptable since the Fixed Incore Detectors, are used as the primary instrumentation for monitoring neutron flux within the reactor core and the MICDS are used as the secondary instrumentation. The Incore Instrumentation System and CETs will still be capable of complying with all the requirements of the Technical Specifications.

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

PC/M 097-992 Supplement 0 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 Auxiliary Building. 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 existing F5 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 line to the F5 warehouse and fire hydrant 34 will be relocated. The facility will also be connected to the plant paging system (Gai-Tronics) to provide page/party line communications and enable emergency alarms and announcements to be heard within the building.

The Craft Lunchroom Facility does not perform any nuclear safety function as it is a structure which provides a personnel facility outside the power block area. The fire protection system and the paging system, providing audible emergency alarms and announcements, do not perform safety related functions. The section of the fire protection system affected by this modification does not protect safety related equipment, is located outside the Appendix R fire areas and can be isolated from the main fire loop if failure occurs. Therefore, the fire line relocation is "Not Nuclear Safety". Extending the plant paging system (Gai-Tronics) to provide page/party line communications and audible emergency alarms and announcements within the building does not affect any safety related equipment. Therefore, the addition of a page/party line and speaker station in the CLF is "Not Nuclear Safety". A safety evaluation has been performed in accordance with 10 CFR 50.59 and is documented in Section 3.0 of this engineering package. This evaluation demonstrates that implementation of this modification does not involve an unreviewed safety question and does not require a change to the Plant Technical Specifications. In addition, this modification has no detrimental effects on plant safety or operation. Based upon the above, prior NRC 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 of occurrence 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 previously evaluated in the SAR may be created; or (iii) if the margin of safety as defined in the basis for any Technical Specification is reduced.

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

0 PC/M 097-992 Supplement 0 Qe SAFETY EVALUATION (continued)

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

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

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

new 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 detect

'hethe release of radioactivity, or to provide post-accident shielding. The facility and modifications to the fire protection system and the plant paging system (Gai-Tronics) do not 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 not increase as a result of this modification.

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

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

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

The new facility is located such that it does not adversely affect any structure, system, or component that functions to mitigate the consequences of an accident, to contain or detect the release of radioactivity, or to provide post-accident shielding. The modification to the fire protection system is designed to ensure that a high level of fire protection is available for structures, systems, and components important to safety and is in compliance with the applicable codes and FSAR requirements for all fire protection equipment. The extension'o the Gai-Tronics system will have no impact to nuclear safety. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in

PC/M 097-992 Supplement 0 SAFETY EVALUATION (continued) the SAR is not changed.

5. Does the proposed change create the possibility of an accident of a different type than any previously 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 Failure Modes and Effects Analysis section, construction of the Craft Lunchroom Facility and the associated modifications to the fire protection system and the plant paging system do not create any new failure modes that can be postulated to cause an accident different than those previously analyzed in the SAR. Furthermore, postulated construction accidents do not impact nuclear safety. Therefore, there is no possibility that an accident may be created that is different from any previously evaluated in the SAR.

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

The Craft Lunchroom Facility, fire protection system and plant paging system do not perform 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 protect safety related equipment, is located outside the Appendix R fire areas and can be isolated from the main fire loop if failure occurs. No new failure modes are created that can be postulated to cause a malfunction of equipment important to safety different than those previously analyzed in the SAR. Therefore, the possibility of a malfunction of equipment important to safety which is of a different type than previously evaluated in the SAR is not created.

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

The Technical Specification requirements applicable to this modification are discussed in Section 3.4 and are not affected. Therefore, this modification does not reduce the margin of 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.

PC/M 067-292 Supplement 0-3 ABSTRACT This engineering package (EP), prepared in accordance with Ql Supplement 3.1-8, provides the reload 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 nominal Cycle 6 length of 11,400 EFPH. Cycle 6 achieved an EOC exposure of 11,819 EFPH. This increased Cycle 6 exposure will require a coastdown at EOC 7 to meet the Cycle 7 target cycle length.

r The primary design change to the core for Cycle 7 is the replacement of 68 irradiated assemblies with 68 fresh Region J (CE-3) fuel assemblies. The fuel is arranged in a low leakage pattern with no significant differences between the Cycle 7 loading pattern and the Cycle 6 design. The mechanical 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 Engineering Nuclear Power, Inc. (ABB/CE) and independently reviewed by Florida Power and Light Co. It has been determined that the operation of the Cycle 7 reload core does not pose an unreviewed safety question and can be implemented with no changes to the St. Lucie Unit 2 Technical Specifications. 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 at the time qf the initial issue. This data is required to support initial startup, power ascension and beginning 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 pressure from 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 References 63 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 available prior to issuance of previous Supplements.

SAFETY EVALUATION In accordance with 10 CFR 50.59 the following discussion demonstrates that there is no unreviewed safety question associated with this reload:

Does the proposed activity increase the probability of an accident previously evaluated in the 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 of occurrence of an accident previously evaluated in the FSAR is not changed.

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

The Cycle 7 design remains bounded by the FSAR analyses. The consequences of accidents analyzed in the SAR remain unchanged. Therefore, the Cycle 7 reload does not increase the consequences of an accident previously evaluated in the FSAR.

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

Plant refueling is a normal plant operation. Probabilities of equipment failure during refueling have already been incorporated into the plant design basis. Implementation of the St. Lucie Unit 2, Cycle 7 reload does not increase the probability of equipment malfunction.

Therefore, the probability of occurrence of any equipment malfunction important to safety previously evaluated in the FSAR will not increase.

Does the proposed activity increase the consequences of a malfunction of equipment important 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 of equipment important to safety previously evaluated in the FSAR.

PC/M 067-292 Supplement 0-3 SAFETY EVALUATION (continued)

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

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

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

Fuel reload is a normal plant evolution. Possible equipment malfunctions have already been postulated 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 equipment important 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 any Technical Specification?

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

The above discussion illustrates that there is no unreviewed safety issue. Analysis and review has also 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.

PC/M 053-292 Supplement 0-1 ABSTRACT The existing St. Lucie Unit No. 2 turbine generator relay protection scheme does not preclude certain 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 installation of 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 has been classified Quality Related.

The safety evaluation of this EP has shown that the implementation of this PCM does not constitute an unreviewed safety question as defined in IO CFR 50.59 and does not require a change in the Plant Technical Specifications. This PCM has no adverse impact on plant safety or 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 battery loading. The Safety Analysis has been updated to reflect this evaluation but the conclusions remain unaffected.

SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed'safety question: (i) if the probability of occurrence 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 for an accident or malfunction of a different type than an 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.

PC/M 053-292 Supplement 0-1 SAFETY EVALUATION (continued)

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

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

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

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

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

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

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

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

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

PC/M 053-292 Supplement 0-1 SAFETY EVALUATION (continued)

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

The equipment added/modified by this EP is not required during an accident condition nor will it prevent safety related equipment from performing their functions. This modification does not affect any safety related equipment. A failure can only cause turbine trip (modes 1 and 2), which is analyzed in FSAR Section 15.2, or a loss of Off-Site Power, which is analyzed in FSAR Section 15.3. Therefore, the possibility of an accident of a different type than any evaluated previously in the SAR is not created.

6. Does the proposed activity create the possibility of a malfunction of equipment important to 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 previously is not created. The equipment added/ modified by this EP (inadvertent energization protective relaying) is powered from a non-safety section of the safety related 125V dc bus 2AB. The circuit is provided with electrical double isolation (breaker and fuse in series) and is located in the section of the bus physically isolated (steel barrier) from the safety related circuits. Therefore, failure of the subject circuit, i.e., short circuit, will be isolated from the safety bus via an actuation of the protective devices, therefore, no Safety Related equipment will be affected and no nuclear safety related equipment will be prevented from performing their design basis functions.

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

The margin of safety as defined in the basis for any Technical Specification is not reduced by this modification since the equipment added/modified by this EP does not form the basis of any Technical Specification. Also, no Plant Technical Specification system availability 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 the bases that this change does not involve an unreviewed safety question or a change to plant Technical Specifications and prior Nuclear Regulatory Commission approval for the implementation of this EP is not required.

PC/M 023-292 Supplement 0 ABSTRACT This Engineering Package (EP) includes the engineering and design necessary to divide the condenser Air Evacuation System (AES) such that each condenser (2A and 2B) will have independent air removal capability. The purpose of the AES is to remove the non-condensible gases 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 respect to air removal capability, such that the condenser with the higher backpressure provides the majority of the takeoff gases. In this situation, the air ejectors remove more steam (also, less non-condensible gases) and the subcooling required for efficient operation of the ejectors is not achieved. The separation of the two condensers will help to increase the efficiency of the AES by 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 ability to separate the 2A and 2B condenser air evacuation piping. It will also involve the installation of taps for local pressure indication on each of the four AES takeoff lines as well as attendant valves for the steam jet air ejectors to accommodate possible future modifications. The implementation of this EP must occur during an outage since it requires that condenser vacuum be broken to install the various components.

This Engineering Package is classified as Not Nuclear Safety, since the affected portions of the condenser and air evacuation systems as described in the Unit 2 FSAR Sections 10.4.1 and 10.4.2, 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 an unreviewed safety question nor a change to Plant Technical Specifications and has no detrimental effect on plant safety or operation. Therefore, prior NRC approval 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 shall be deemed to involve an unreviewed safety question: (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased, or (ii) if a possibility for an accident or malfunction of 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 safety question because of the following reasons:

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

PC/M 023-292 Supplement 0 SAFETY EVALUATION (continued)

The proposed activity does not increase the probability of occurrence of an accident previously evaluated in the SAR since the portions of the components and systems affected by this modification do not serve a safety function. They are not required for safe shutdown or to mitigate the effects of a LOCA. This modification does not adversely affect any accident initiating components.

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

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

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

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

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

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

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

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

PC/M 023-292 Supplement 0 SAFETY EVALUATION (continued) pipe supports added by this modification will in no way interact with any safety related equipment. The operation of this modified system will not result in any unanalyzed safety related system configurations.

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

The proposed activity does not create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the SAR because this modification does not impact any safety related equipment and does not introduce any new failure modes. The air ejector radiation monitor, which is used to indicate the presence of a 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 any technical specification?

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

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

PC/M 563-291 Supplement 0 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 inverter malfunction resulted in a high AC output voltage condition which went undetected for several days. In order to facilitate the new alarm, an existing alarm for DC Input Breaker Tripped condition will be disconnected. The alarm for DC Input Breaker Tripped can be deleted since this alarm is enveloped by the Low DC Voltage Alarm. In addition to the new alarm circuit being added, this EP is also removing the trip function'of the instrument inverters'C input breaker on Low DC Voltage, High DC Voltage, and Low AC Output Voltage to eliminate automatic power loss to safety related instrumentation during an inverter degraded operating condition. The setpoints for the instrument inverter alarm circuits are being adjusted (or established if not specified) in accordance with equipment protection and/or system operational requirements.

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

The safety evaluation of this EP has determined that this PCM does not constitute an unreviewed safety question as defined in 10 CFR 50.59 and does not require a change in the plant Technical Specifications. This PCM has no adverse impact on plant safety or operation. Therefore, this PCM 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 shall be deemed to involve an unreviewed safety question: (i) if the probability of occurrence 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 for an accident or malfunction 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 this modification constitutes an unreviewed safety question of requires a change to the Technical Specifications:

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

t PC/M 563-291 Supplement 0 SAFETY EVALUATION (continued)

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

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

The instrument inverters supply uninterruptible 120 VAC power to instrumentation and protection systems required to mitigate the consequences of an accident. The redundant features of the vital 120-VAC system ensure that a single instrument inverter failure will not prevent actuation of the protection systems or cause loss of vital process control and monitoring systems.

The addition of an alarm for High AC Output Voltage and the adjustment of existing alarm setpoints will help minimize the possibility of an undetected inverter degraded operating condition. The new AC voltage sensing cards are electrically connected in accordance with existing design criteria and seismically qualified and mounted to prevent adverse affects on the inverter control circuitry. The disconnection of the trip circuitry to the inverter DC input breaker does not alter the failure mode analysis of the instrument inverters. Therefore, the proposed modifications do not compromise the ability of the instrument inverters they supply to perform their safety function. Accordingly, the consequences of an accident previously evaluated in the SAR have not been increased.

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

A malfunction of an instrument inverter voltage regulation circuitry resulted in a high AC output voltage condition which went undetected for several days. The addition of an alarm for this condition will not reduce or increase the probability of this malfunction from occurring; 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 mounted within 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 the existing inverter cabinets.

PC/M 563-291 Supplement 0 SAFETY EVALUATION (continued)

The removal of the trip circuitry of the DC input breaker does not alter the associated failure mechanisms (i.e. Low/High DC Voltage) or failure consequences (i.e. Low AC Output Voltage). Although these conditions may still cause a loss of the instrument inverter, the actual failure of the inverter and/or its loads must now occur, not just the detection of a off-normal operating condition. In addition, the loss of an instrument inverter due to a transient input/output voltage or malfunction of the voltage sensing cards has been eliminated. Therefore, the proposed change does not increase and actually decreases the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR.

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

The instrumentation and control equipment powered by the instrument inverters are designed and electrically aligned such that a single inverter failure will not prevent safe shutdown or reduce the ability to mitigate the consequences of an accident. A single inverter failure will also not cause spurious actuation of protection systems due to the use of four independent process channels and two out of four logic for actuation. The proposed changes to the alarm and trip circuitry do not affect the redundancy and independence of the 120 VAC system or alter the inverters'ailure mode analysis.

Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the SAR has not been increased.

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

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

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

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

PC/M 563-291 Supplement 0 SAFETY EVALUATlON (continued)

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

This modification does not adversely affect the operational requirements or reliability of the instrument inverters. The additional alarm for High AC Output Voltage and the conservative adjustment of the existing alarm setpoints will provide the operator with improved notification of an inverter degraded operating condition. The deletion of the trip circuitry of the inverters'C input breaker reduces the possibility of an inadvertent loss of an instrument inverter. Since there is no impact on the Surveillance Requirements and Limiting Conditions for Operation of the 120 VAC System, this modification does not reduce the margin 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.

PC/M 543-291 Supplement 0 ABSTRACT This Engineering Package (EP) provides the engineering and documentation necessary to disconnect and abandon in place specific heat trace circuits in the Liquid Waste Management System at St. Lucie Unit 2. The heat tracing circuits being disconnected are associated with the Waste Concentrator Package. The heat tracing is not required due to the boric acid solution in the Waste Concentrator Package being administratively controlled at a concentration up to 3.5 weight percent.'oric acid solution with a concentration up to 3.5 weight percent present in pipe lines and components located in the Auxiliary Building do not need heat tracing per FSAR Section 9.3.4.3.1 1.

~

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

The safety evaluation of this EP has determined that this PC/M does not constitute an unreviewed safety question as defined in 10 CFR 50.59 and does not require a change in the Plant Technical Specifications. This PC/M has no adverse impact on plant safety or operation; thus this PC/M can 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) if the probability of occurrence 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 for an accident or malfunction of 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. In accordance with 10CFR50.59, the following evaluation serves to determine whether this modification constitutes an unreviewed safety question:

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

The malfunction of LWMS is not postulated to initiate an accident previously evaluated in the SAR. The design objective of the LWMS is to protect plant personnel, the general public and the environment by assuring all releases of radioactive liquids is performed in a 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 LWMS to meet this design objective because of the redundant function performed by the Boric Acid Concentrators and administrative controls imposed on the boric acid concentration of the liquid wastes processed by the WCP. The LWMS will still be capable of effectively controlling, monitoring and processing radioactive liquid wastes in accordance with the existing design criteria after the WCP heat tracing is disconnected. The design provisions

PC/M 543-291 Supplement 0 SAFETY EVALUATION (continued) and controls provided to prevent inadvertent or uncontrolled release of radioactive liquid are not affected. Therefore, the probability of an accident previously evaluated in the FSAR has not been increased.

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

No credit is taken for operation of the LWMS to mitigate the consequences of any design basis accident. The system is used to control, monitor and process liquid wastes during normal operation of the plant. The disconnection of the WCP heat tracing does not affect the operation of the effluent line radiation monitor or any isolation valves which automatically close on inadvertent release of radioactive liquids. The design provisions provided to control the release of radioactive materials due to waste surges or tank overflows is also not affected. Therefore, the consequences of an accident previously evaluated in the FSAR have not been increased.

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

The function of the WCP heat tracing was to prevent equipment/system failure due to solidiTication of the boric acid in the pipe lines, components, or tanks: The administrative controls 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 circuits will not increase the probability of equipment/system failure. Liquid wastes with a boric acid concentration of greater than 3.5 weight percent is routed to the Boric Acid Concentrators for processing. Since the LWMS is operated in a batch mode and is designed to handle surges in influent rate, additional waste liquid processing by the Boric Acid Concentrators will not affect overall system performance or equipment design specifications to be exceeded. Therefore, the probability of a malfunction of equipment important to safety previously evaluated in the FSAR would not be increased.

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

For the worst-case scenario, the malfunction of equipment in the LWMS would cause the inadvertent or uncontrolled release of radioactive materials. The disconnection of the WCP heat tracing circuits does not increase the severity of this worst-case scenario because the amount 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) provided to mitigate the consequences of malfunction of equipment are not affected by this PCM.

Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the SAR has not been increased.

tl PC/M 543-291 Supplement 0 SAFETY EVALUATION (continued)

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

This modification does not change the design bases, functional requirements, or performance of the LWMS. No new hazards are created that can be postulated to cause an accident different than those previously analyzed in the SAR. There are no new failure modes 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 be created that is different from any already evaluated in the SAR.

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

The disconnection of the WCP heat tracing circuits does not alter the failure modes of associated equipment since administrative controls have been imposed on the boric acid concentration (< 3.5 weight percent) of the liquid waste processed by the WCP. The ambient temperature of the auxiliary building (location of the WCP) will be sufficient to prevent solidification of boric acid within the WCP, which is the only possible failure mechanism applicable to a loss of heat tracing. No new hazards are created that can be postulated 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 a different 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 any Technical Specifications?

The basis for the LWMS is to ensure that releases of radioactive materials in liquid effluents be kept As Low As is Reasonably Achievable (ALARA)in accordance with the requirements of 10CFR20 and 10CFR50, Appendix I. Appropriate portions of the LWMS must be available 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 processed by the WCP is acceptable since the liquid wastes will be routed to the Boric Acid Concentrators, where the redundant processing is performed. As a result, the availability of the LWMS to perform its design basis function is not compromised. The disconnection of the WCP heat tracing circuits does not affect the effluent monitoring or design features for 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 the Technical Specifications.

PC/M 543-291 Supplement 0 SAFETY EVALUATION (continued)

The foregoing constitutes, per 10 CFR 50.59(b), the written safety evaluation which provides the bases 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 not required.

PC/M 510-291 Supplement 0-1 ABSTRACT NRC Generic Letter 89-10 requires that operating nuclear plants develop and implement a program to ensure that switch settings on all safety-related motor-operated valves (MOVs) are correctly selected, set and maintained to accommodate the maximum differential pressures expected on these valves during all postulated events within the design basis. Item a) of the Letter requires that the design basis for these MOVs be reviewed to determine the maximum differential 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, and FPL'Engineering Evaluation JPN-PSL-SEMP-91-048, Revision 0.

Item b) of Generic Letter 89-10 requires that the licensee establish the correct MOV switch settings based on the previously determined maximum differential pressure. AII switches, including torque switches, torque bypass switches, position limit, position indication, overloads, etc., shall be considered. 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 to meet 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, or may 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 against the requirements of 10CFR50.59, and it was concluded that these modifications do not constitute an 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 numerous valves. 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 the modification increasing the total to sixty (60) valves. In addition, the methodology for setting the open torque switch is revised due to a change in the diagnostic equipment which has been selected for use. This supplement revises the safety evaluation to reflect the revised design information. However, the original conclusions of the safety evaluation, that the change does not constitute an unreviewed safety question nor require a change to the plant Technical Specifications remain unchanged. Therefore, prior NRC approval for the implementation of this PC/M is not required.

PC/M 510-291 Supplement 0-1 SAFETY EVALUATION As defined in 10 CFR 50.59, an unreviewed safety question exists; (i) if the probability of occurrence 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 possibility of an accident or malfunction of a different type than any previously evaluated in the SAR may be created; or (iii) if the margin of safety as defined in the basis for any Technical Specification is reduced.

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

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

This modification does not affect any equipment whose malfunction is postulated in the SAR to initiate an accident or prevent an accident from occurring. The modifications performed by this Engineering Package enhance the ability of the components to perform as intended during emergency and off-normal conditions under maximum differential pressures.

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

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

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

PC/M 510-291 Supplement 0-1 SAFETY EVALUATION (continued)

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

System operability is not being affected by the modifications to the MOVs identified in this Engineering 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 this modification serves to provide additional design documentation, or replace existing parts.

Replacement of the spring packs and revising the thrust or torque values for the MOV operators only serve to enhance the operational characteristics of the MOVs. As such, no new accident initiating events are created. Therefore, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR has not increased by this modification.

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

System operation is not affected by this modification. This modification does not interact spatially or functionally with any structure, system or component important to safety other than the valves and valve operators themselves. Although actuator and valve loadings may increase, the revised loads are within the published ratings for the components.

Replacement components have been selected in accordance with the same design criteria as the original components. The modifications performed by this Engineering Package enhance the ability of the valves and valve operators to perform as intended during emergency and off normal conditions under maximum differential pressures. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the SAR is not increased by this modification.

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

As discussed in Section 3.3, this modification does not change the function or design bases of any structure, system or component important to safety as described in the SAR.

This modification provides increased design documentation, makes adjustments to components within their published operating range or makes replacements of equivalent parts. No new failure modes or conditions are created that can be 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 evaluated in the SAR is not created by this modification.

PC/M 510-291 Supplement 0-1 SAFETY EVALUATION (continued)

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

This modification does not interact spatially or functionally with any structure, system or component important to safety other than the valves and valve operators themselves. This modification does not alter the function or the design basis of any MOV. As discussed in Section 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 previously analyzed in the SAR. Therefore, the possibility of a malfunction of equipment important to safety which is of a different type than any previously evaluated in the SAR is not created by this modification.

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

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

The foregoing discussions provided in this EP constitutes, per CFR 50.59(b), the written safety evaluation which provides 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 Technical Specifications. As such, prior NRC approval for the implementation of this PC/M is not required.

PC/M 500-291 Supplement 0 ABSTRACT This Engineering Package (EP) provides for the addition of a removable visual level indicator to the 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 Seismic Category I Fuel Handling Building (FHB) and has been seismically designed to preclude interaction with spent fuel assemblies stored in the SFP. Accordingly, this PC/M has been classified as Quality Related. A safety evaluation has been performed in accordance with 10 CFR 50.59 and is documented in this engineering package. This evaluation demonstrates that implementation of this modification does not involve an unreviewed safety question and does not require a change to the Plant Technical Specifications. In addition, this modification has no detrimental effects on plant safety or operation. Based upon the above, prior NRC approval is not required for the implementation of this modification.

SAFETY EVALUATION As defined in 10CFR50.59, an unreviewed safety question exists; (i) if the probability of occurrence 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 previously evaluated in the SAR may be created; or (iii) if the margin of safety as defined in the basis for any Technical Specification is reduced.

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

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

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

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

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, to contain or detect the release of radioactivity, or to provide post-accident shielding. The

PC/M 500-291 Supplement 0 SAFETY EVALUATION (continued) visual level indicator does not perform any nuclear safety function and has been seismically designed to preclude interaction with any safety related item. Therefore, the consequences of an accident previously evaluated in the SAR will not increase as a result of this modification.

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

The visual level indicator is designed to ensure that interaction does not occur with any structure, system, or component important to safety. The visual level indicator has been seismically designed to preclude interaction with any safety related items and will not interfere with the operation of any nuclear safety related systems. Therefore, the probability of occurrence of any equipment malfunction important to safety previously evaluated in the SAR will not be increased.

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

The visual level indicator is designed to ensure that interaction does not occur with any structure, system or component important to safety. The visual level indicator is seismically designed to preclude interaction with any safety related items. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the SAR is not changed.

E Does the proposed change create the possibility of an accident of a different type than any previously 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. Per The Failure Modes and Effects Analysis section of this EP, installation of the visual level indicator does not create any new hazards that can be postulated to cause an accident different than those previously analyzed in the SAR. The visual level indicator does not perform any nuclear safety function. Therefore, there is no possibility that an accident may be created that is different from any previously evaluated in the SAR.

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

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

PC/M 500-291 Supplement 0 SAFETY EVALUATION (continued) interaction with any safety related items. No new hazards are created that can be postulated to cause a malfunction of equipment important to safety different than those previously analyzed in the SAR. Therefore, the possibility of a malfunction of equipment important to safety which is of a different type than previously evaluated in the SAR is not created.

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

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

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.

II

PC/M 486-291 Supplement 0 ABSTRACT This Engineering Package (EP) includes the engineering 8 design necessary to delete the Main Turbine Runback feature. The modification will leave the Turbine Runback logic unchanged. It will no longer be initiated. Turbine Runback occurs whenever there is a loss of a Steam Generator 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 the Turbine 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 Board (RTGB) 202. Cables 20712F, 20712J and 20712L will be spared and all 'enerator drawings will be revised accordingly. In addition the Control'Room Turbine Runback Annunciator Window 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 rate upon 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 Runback the Main Governor valves throttle the steam flow until the load matches the setpoint of 60% or 70% load depending on the initiating event. During this event the Turbine/Generator RPM remains constant.

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 events a turbine runback was automatically initiated to approximately 60% power. In less than 30 seconds into the transient the Reactor Protection System initiated a Reactor trip on a high pressurizer pressure signal.

The purpose of removing the Turbine Runback feature is to minimize the effects of a partial loss of feedwater transient on the Plant and to provide the plant operators with additional time to restore 100% feedwater flow before a Reactor trip occurs. Although the two previously mentioned transients that caused a Reactor trip are for Unit 1, removal of the Turbine Runback feature will be implemented for Unit 2 by this PC/M since both Units respond to this transient in a similar manner 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 the Turbine Runback feature. This analysis demonstrates that by removing the Turbine Runback feature a Reactor trip could be avoided, provided the plant operators restore 100% feedwater flow within 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 the Pressurizer Power Operated Relief Valves (PORVs) or the Main Steam Safety Valves (MSSVs) which lifted.

PC/M 486-291 Supplement 0 ABSTRACT (continued)

Although the Main Turbine, Turbine Controls and the Turbine Runback feature do not perform a safety function per FSAR Section 7.7, this EP is classified as Quality Related because it requires work 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 an unreviewed safety question nor require a change to Plant Technical Specification and has no detrimental effect on plant safety or operation. Therefore, prior NRC approval for implementation of this modification is not required.

SAFETY EVALUATlON With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question: (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased, or (ii) if a possibility for an accident or malfunction of 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 safety question because of the following reasons:

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

The proposed activity does not increase the probability of occurrence of an accident previously evaluated because the functionality of the Main Turbine, Turbine Controls or Reactor 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 evaluated in the SAR?

The proposed activity does not increase the consequences of an accident due to the deletion 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 not affect any structure, systems or components that function to deter the release of radioactivity or to provide post-accident shielding.

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

PC/M 486-291 Supplement 0 SAFETY EVALUATION (continued)

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

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

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

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

The proposed activity does not create the possibility of an accident of a different type than any previously evaluated because during the loss of a SGFP transient without Turbine Runback the heat removal from the Reactor Coolant System (RCS) by the secondary side of 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 Turbine Runback feature operative. The Turbine Runback feature is not Safety Related and the deletion of the Turbine Runback does not affect any Safety Related signals required to initiate a Reactor trip. This modification does not degrade the reliability or increase challenges, directly or indirectly for equipment important to safety. For these reasons, the proposed activity does not create the possibility of an accident of a different type than previously described in the FSAR.

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

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

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

PC/M 486-291 Supplement 0 SAFETY EVALUATION (continued)

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

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

PC/M 421-291 Supplement 0 ABSTRACT During operation of the 2A and 2B Emergency Diesel Generators (EDGs), the exposed rotating shaft couplings between the generator and diesel engines present a personnel hazard. This engineering package provides for the installation of sheet metal guards around the two shaft couplings located on each EDG. Installation of the guards will enhance personnel safety by providing 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 or operation, do not constitute an unreviewed safety question, and do not require a change to the Technical Specifications. Therefore, prior NRC approval for implementation is not required.

SAFETY EVALUATION As defined in 10CFR50.59, an unreviewed safety question exists; (i) if the probability of occurrence 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 the possibility for an accident or malfunction of a different type than any previously evaluated in the SAR may be created; or (iii) if the margin of safety as defined in the basis for any Technical Specification is reduced.

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

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

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

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

PC/M 421-291 Supplement 0 SAFETY EVALUATION (continued)

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, to contain or detect the release of radioactivity, or to provide post-accident shielding. The EDG shaft coupling guards do not perform any nuclear safety function and have been seismically designed to preclude interaction with any safety related items. Therefore, the consequences of an accident previously evaluated in the SAR will not increase as a result of this modification.

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

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

'oes the proposed change increase the consequences important to safety previously evaluated in the SAR?

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

Does the proposed change create the possibility of an accident of a different type than any previously 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. Installation of the EDG shaft coupling guards does not create any new hazards that can be postulated to cause an accident different than those previously analyzed in the SAR. The EDG shaft coupling guards do not perform any nuclear safety function. Therefore, there is no possibility that an accident may be created that is different from any previously evaluated in the SAR.

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

PC/M 421-291 Supplement 0 SAFETY EVALUATION (continued)

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

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

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

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.

PC/M 419-291 Supplement 0-1 ABSTRACT The atmospheric dump valves (ADVs) actuator control logic was modified (PCM 069-286) so that the closing direction of these valves would be controlled by the torque switch (in lieu of the limit switch) to ensure positive valve seating. However, difficulty has been experienced in adjusting the torque switch of these motor operated valves to achieve the minimum amount of thrust required for 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 be significantly reduced in the closing direction by installing a compensating spring pack on the actuator. This field modification will convert the ADV actuator from the existing-Limitorque SMB-0 model to be functionally the same as an SB-0 type actuator. The capacity of the actuator is not adversely affected by this modification, as these Limitorque models have interchangeable parts and have the same thrust and torque ratings. Since the ADVs are controlled by a limit switch in the 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 ADV actuators, tag numbers MV-08-18A, -18B, -19A, -19B. The actuators will be modified by installing a compensating spring pack assembly; this spring pack, and its cover, will be installed on top of the existing actuator housing. It is actually a Belleville spring mounted above the stem nut. The stem nut is thereby allowed to "float" upward upon valve seating, rather than being rigidly held in place. The spring will compress a fixed amount for a given seating load, and will compensate for load variations should the ADV cool or heatup.

Following implementation of this PCM, the torque switch settings will be adjusted to satisfy the requirements 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 Safety Related. A review of the modification to be implemented by this PC/M was performed against the requirements of 10CFR50.59, and it was concluded that these modifications do not constitute an 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 NO. 1 Supplement No. 1 to this Engineering Package provides the vendor instructions to modify the ADV actuator, provides additional drawings, adds the TEDB change package and lists additional Engineering procured parts. Neither the design, the installation, nor the performance of the ADV actuators is adversely affected by this Supplement. Supplement No. 1 does not effect, amend nor change the original Safety Evaluation, the Technical Specifications or the Technical Specification bases.

e

PC/M 419-291 Supplement 0-1 SAFETY EVALUATION As defined in 10 CFR 50.59, an unreviewed safety question exists; (i) if the probability of occurrence 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 previously evaluated in the SAR may be created; or (iii) if the margin of safety as defined in the basis for any Technical Specification is reduced.

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

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

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

The installation of compensating spring packs in the ADV actuators only serve to enhance .

the operational characteristics of the ADVs. As such, no new accident initiating events are created. Therefore, the modifications described in this Engineering Package do not increase the probability of valve failure, and thus the probability of occurrence of an accident previously described in the SAR is not increased by this modification.

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

This modification does not affect any structures, systems or components that function to deter the release of radioactivity or to provide post-accident shielding. The modifications performed by this Engineering Package do affect the ADVs, which may be relied upon to cooldown the plant. However, this function is not adversely affected by installation of the compensating 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 this modification.

PC/M 419-291 Supplement 0-1 SAFETY EVALUATION (continued)

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

System operability is not being affected by the modifications to the ADVs identified in this Engineering Package. Valve operability will not be adversely affected by the prescribed modifications. 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 compensating spring packs in the ADV actuators will reduce the total thrust on the ADV while not reducing the actuator capacity. As such, no new accident initiating events are created.

Therefore, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR is not increased by this modification.

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

System operation is not affected by this modification. This modification does not interact spatially or functionally with any structure, system or component important to safety other than the valves and valve operators themselves. The compensating spring packs have been selected in accordance with the same design criteria as the original components.

The modifications performed by this Engineering Package do not adversely affect the ability of the valves and valve operators to perform as intended during emergency and off-normal conditions under design conditions. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the SAR is not increased by this modification.

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

As discussed previously, this modification does not change the function or design bases of any structure, system or component important to safety as described in the SAR. This modification installs compensating spring packs which in turn increase the reliability of the valve/actuator combination by reducing stresses imparted on the valve. No new failure modes or conditions are created that can be 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 evaluated in the SAR is not created by this modification.

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

PC/M 419-291 Supplement 0-1 SAFETY EVALUATION (continued)

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

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

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

The foregoing discussions provided in this EP constitutes, per 10 CFR 50.59(b), the written safety evaluation which provides 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 Technical Specifications. As such, prior NRC approval for the implementation of this PC/M is not required.

PC/M 418-291 Supplement 0 ABSTRACT Engineering Evaluation (EE) JPN-PSL-SEMP-91-029, Rev. 0," Engineering Evaluation of Shutdown Cooling 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 no means of being vented. The EE recommends that vent valves be installed on the CS header upstream of the containment isolation valves. As valves 1-FCV-07-1A and 1-FCV-07-1B are normally closed and isolate the containment, the above mentioned vents need to be located at high'oints of the headers upstream of these isolation valves.

This Engineering Package (EP) provides the specific design information necessary to install one 3/4" vent in each CS header immediately up stream of valves 1-FCV-07-1A and 1-FCV-07-1 B. In addition this EP revises plant drawings for administrative changes by correcting design and hydrotest 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 on plant safety and/or operation. Based on a Failure Mode and Effects Analysis and a review of the changes to be implemented by this EP against the requirements of 10CFR50.59, it was concluded that these modifications do not constitute an unreviewed safety question and do not require a change to the plant Technical Specifications. Therefore, prior NRC approval for the implementation 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 shall be deemed to involve an unreviewed safety questions: (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased, or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the Final 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 here in does not involve an unreviewed safety question because of the following reasons:

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

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

PC/M 418-291 Supplement 0 SAFETY EVALUATION (continued)

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

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

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

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

" Does the proposed activity increase the consequences important to safety previously evaluated in the SAR?

of a malfunction of equipment The proposed activity does not increase the consequences of a malfunction of equipment important to safety because the vents perform no active nuclear safety related or equipment protection function. The vent valves are administratively controlled to prevent their being left open and are designed to be physically strong enough to preclude being broken off. In the extreme unlikely event that the vent fails, the worse-case scenario is the loss 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 of equipment important to safety.

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

The proposed activity does not create the possibility of an accident of a different type than any previously evaluated because the vents are not accident initiating devices. The vent valves are only operated when filling the SDC system and serve no active or controlling function.

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

PC/M 418-291 Supplement 0 SAFETY EVALUATION (continued)

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

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

The proposed activity does not reduce the margin of safety as defined in the bases for any Technical Specification because T.S. Section 3/4.6.2 requires that both loops of the CS be operable, all valves in the CS System be properly aligned and that the pumps and initiation system be tested. The Technical Specification bases ensure that adequate containment cooling 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 the bases that this change does not involve an unreviewed safety question nor a change to the Plant Technical Specification. Prior NRC approval for the implementation of this modification is not required.

PC/M 309-291 Supplement 0 ABSTRACT This Engineering Package provides for enhancement modifications to the Containment Hydrogen Analyzer System. The modifications which are designed to help ensure maintainability of the Containment Hydrogen Analyzer System and enhance its performance involve the following:

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

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

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

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

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

a) Incorporate the vendor's latest (May 1991) spare parts replacement schedule.

b) Correct the amp board schematic drawing with regard to the range (0-10% H,)

selector switch.

c) Correct the stabilization time period called for during the zero and span adjustments of the calibration procedure (due to a change in the catalyst bed) ~

d) 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 are used to limit the buildup of hydrogen in containment during a LOCA.

The Containment Hydrogen Analyzer System is used during a LOCA to monitor the hydrogen concentration in containment, such that appropriate operator actions can be taken to ensure that the hydrogen flammability limit of 4% is not exceeded. The Containment Hydrogen Analyzer System consists of two redundant trains. Each redundant train is physically separated, operated independently, and powered from an independent onsite power source. The piping to and from the 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 1E instrument. Based on the above this Engineering Package is classified as Safety Related.

PC/M 309-291 Supplement 0 ABSTRACT (continued)

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 nor a change to Technical Specifications. Additionally, it has no adverse effect on plant safety or operation. 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 shall be deemed to involve an unreviewed safety question: (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased, or (ii) if a possibility for an accident or malfunction of 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 modifications included in this Engineering Package do not involve an unreviewed safety question because of the following reasons:

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

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

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

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

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

1 The propo'sed activity does not increase the probability of occurrence of a malfunction of equipment important to safety because the Containment Hydrogen Analyzer System modifications are designed to ensure that safety related equipment is not adversely impacted. The modifications were evaluated for seismic excitations as required to assure continued functioning of safety related equipment.

0

PC/M 309-291 Supplement 0 SAFETY EVALUATION (continued)

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

The proposed activity does not increase the consequences of a malfunction of equipment important to safety because the Containment Hydrogen Analyzer System modifications are designed 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 any previously evaluated in the SAR?

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

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

The proposed activity does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated because the Containment Hydrogen Analyzer System modifications are designed to ensure that safety related equipment is not adversely impacted. All new components have been evaluated for 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 any technical specification?

The proposed activity does not reduce the margin of safety as defined in the bases for any Technical Specification because the Containment Hydrogen Analyzer System modifications do 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 systems required for the detection and control of hydrogen gas are available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions. No margin of safety is affected by this modification.

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

PC/M 247-291 Supplement 0 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 system which 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 with amplifiers which have not been effective or reliable in providing an accurate measurement of reactor coolant pump oil level. This measurement system is being replaced with a "bubbler" type level measurement system. This new system will be designed to interface with the remaining RCP oil level measuring instrumentation loop components (LIA-1156, 1157, 1166, 1167, 1176, 1177, 1186 and 1187).

The bubbler system will consist of a regulator, purgemeter, dip tube, transmitter, and excess flow valve 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 related pipe 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" Fire Protection analysis is a concern. The installation of the bubbler systems is being seismically mounted 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 question and prior NRC approval is not required for implementation. The implementation of this EP does not 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 shall be deemed to involve an unreviewed safety question: (i) if the probability of occurrence 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 the possibility for an accident or malfunction of 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 basis for any Technical Specifications is reduced.

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

1. Does the Proposed Activityincrease the probability of occurrence of an accident previously evaluated in the Safety Analysis Report?

0

PC/M 247-291 Supplement 0 SAFETY EVALUATION (continued)

The modification does not increase the probability of occurrence of an accident previously evaluated. As mentioned in Section 3.2 of this EP, the RCP lube oil system does not have a detailed description in the FSAR, and has no safety related function. However, the installation of the bubbler system is being seismically mounted as to protect safety related equipment. Therefore, the proposed activity will not increase the probability of occurrence of an accident.

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

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

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

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

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

The modification does not increase the consequences of a malfunction of equipment important to safety because the RCP oil system and the monitoring of its level perform no safety functions and has been designed not to interact with safety equipment during an accident.

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

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

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

PC/M 247-291 Supplement 0 SAFETY EVALUATION (continued)

The proposed change does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR because the RCP lube oil system is a non safety related system and the added piping, components and modified pipe supports are designed to the same requirements of ANSI B31.1 1973 Edition 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 way of independent verification of design in addition to enhanced quality level for procurement of components. The enhanced quality level provides for certification to the design codes for the components.

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

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

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

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

PC/M 092-291 Supplement 1 ABSTRACT This Engineering Package provides for the replacement of two obsolete Fischer and Porter Model 51-1401 indicating controllers installed in the St Lucie Plant - Unit No. 2 for component cooling water temperature local indication and control. In the current plant configuration, temperature indicating controllers TIC-14-4A and TIC-14-4B throttle temperature control valves TCV-14-4A and TCV-14-4B to regulate Intake Cooling Water flow depending upon Component Cooling Water outlet 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 these obsolete controllers with new, currently available, pneumatic controllers which are qualified Seismic Category I to provide additional conservatism in the design. Similarly, existing high limit relays and pressure 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 question nor require a change to the Technical Specifications, therefore prior NRC approval is not required for 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 shall be deemed to involve an unreviewed safety question: (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased, or (ii) if a possibility for an accident or malfunction of 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 safety question as demonstrated by the answers to the following questions:

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

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

Therefore, the proposed activity does not increase the probability of occurrence of, an accident previously evaluated in the SAR.

l' PC/M 092-291 Supplement 1 SAFETY EVALUATION (continued)

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

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

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

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

The proposed change does not involve any in-line instrumentation and is separated from entrained fluids by a Seismic Category I (ASME Class 3) thermowell. Pneumatic pressure regulators and high limit relays are maintained in the design, and helical bourdon tubes similar to those of the existing indicating controllers are employed in TIC-14-4A and TIQ-14-4B. The mounting of the new instrumentation, tubing and supports has been seismically designed to provide additional conservatism in the design. Therefore, there is no increase in the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the Safety Analysis Report.

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

This modification replaces obsolete Component Cooling Water (CCW) temperature indicating controllers and associated instrumentation. The replacement components are like-for-like replacements, perform the same functions as the existing equipment. Existing system interface points (thermowells, supply air, valve positioners, racks and supports) are compatible with the new equipment and are not adversely impacted by this modification.

Therefore, there is no increase to the consequences of a malfunction of equipment important to safety previously evaluated in the Safety Analysis Report.

PC/M 092-291 Supplement 1 SAFETY EVALUATION (continued) r

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

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

6. Does the Proposed Activity Create the Possibility of a Malfunction of Equipment Important to Safety of a Different Type Than Any Previously Evaluated in the SAR?

The proposed change does not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR since the existing CCW operating characteristics are maintained by utilizing the same pneumatic control 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 any Technical Specification?

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

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

PC/M 091-291 Supplement 0 ABSTRACT This Engineering Package (EP) includes the engineering and design necessary to add discharge resistors 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 Design Inadequacies Study, EBASCO report which states that when MOVs are energized and de-energized, this action results in high voltage transients which can cause shunt coil insulation damage and consequent premature shunt coil failure. This concern is addressed in NRC Information Notice 88-72. The MOVs that are affected are:

Ta Number MOV Descri tion MV-08-12 Steam Generator 2B to AFWP 2C Turbine MV-08-13 Steam Generator 2A to AFWP 2C Turbine MV-08-14 Atmospheric Steam Dump Isolation Valve MV-08-15 Atmospheric Steam Dump Isolation Valve MV-08-16 Atmospheric Steam Dump Isolation Valve MV-08-17 Atmospheric Steam Dump Isolation Valve MV-09-11 AFWP 2C Discharge to Steam Generator 2A MV-09-12 AFWP 2C Discharge to Steam Generator 2B This EP involves modification of Nuclear Safety Related MOVs, and is therefore classified as Nuclear Safety Related.

The safety evaluation of this EP has determined that this plant change modification (PC/M) does not constitute an unreviewed safety question as defined in 10 CFR 50.59 and does not require a change in the Plant Technical Specifications. This PC/M has no adverse impact on plant safety or 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 shall be deemed to involve an unreviewed safety question: (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased, or (ii) if a possibility for an accident or malfunction of 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. In accordance with 10 CFR 50.59, the following evaluation serves to determine whether this modification constitutes an unreviewed safety question:

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

0

PC/M 091-291 Supplement 0 SAFETY EVALUATION (continued)

The malfunction of MOVs modified by this PC/M does not adversely affect any equipment whose 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 the MOV 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 the addition of a discharge resistor as per this PC/M. The ADV isolation valves are normally locked open. Accordingly, the ability to provide feedwater for the removal of decay heat from the reactor coolant system during normal and natural circulation system cooldown as well 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/M since the discharge resistor electrically drops out of the MOV electrical circuit very quickly in comparison to the time it takes to open or close a MOV. This modification does not circumvent the valve safety function. This modification does not affect MOV pressure boundaries. Therefore, the probability of occurrence of an accident previously described in the SAR is not increased by this modification.

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

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

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

This modification adds discharge resistors across the shunt field winding of each 125 VDC MOV to protect against potential high voltage transients that may occur during de-energization of a MOV. However, a malfunction of the discharge resistor will result in an open circuit across the IVIOV shunt field winding, which in turn will be electrically equivalent to the existing MOVs. The addition of the discharge resistors is considered an enhancement 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 safety previously evaluated in the SAR is not increased by this PC/M.

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

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

PC/M 091-291 Supplement 0 SAFETY EVALUATION (continued) discharge resistor across a 125 VDC MOV shunt field winding is electrically equivalent to the existing motor circuit. The addition of discharge resistors does not adversely affect the performance of the MOV in a mitigating activity. Therefore, the implementation of this EP does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR.

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

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

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

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

t Does the proposed activity reduce the margin of safety as defined in the bases for any Technical Specification?

The modification does not reduce the margin of safety as defined in the bases for any Technical Specification. This modification adds a discharge resistor across the 125 VDC MOV shunt field winding. The operation of MOVs modified by this EP is functionally equivalent to the existing MOVs. The purpose of the discharge resistor is to protect 125 VDC MOVs from potential motor damage that can be caused by high voltage transients created during de-energization. The 125 VDC system calculations were reviewed to determine that this increased loading has a negligible affect on the 125 VDC System. The addition of a discharge resistor does not adversely affect the MOVs performance. The existing margin of safety as defined in the basis for any Technical Specification remains e unchanged after the implementation of this modification.

PC/M 091-291 Supplement 0 SAFETY EVALUATION (continued)

The forgoing constitutes, per 10 CFR 50.59(b), the written safety evaluation which provides the bases 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.

PC/M 256-290 Supplement 1-2 ABSTRACT This engineering package (EP) provides for the safety evaluation of the St. Lucie Unit 2 Cycle 6 reload design. The Cycle 6 energy requirement is 12000 EFPH based on a Cycle 5 length of 11500 EFPH. The replacement of approximately one third of the fuel assemblies with fresh ones is 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 with 76 fresh Region H fuel assemblies. The fuel is arranged in a low leakage pattern with no significant differences between the Cycle 6 loading pattern and the Cycle 5 design. The mechanical 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 shutdown of the reactor. The design analysis associated with this reload design evaluates plant operating parameters that are associated with the capability to shutdown the reactor and maintain it in a safely shutdown condition.

Based upon the design analysis by Combustion Engineering, which was independently verified by 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 Technical Specifications. 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 which provides recommendations to the St. Lucie Unit 2 plant staff concerning excore detector electronic adjustments, was changed. An additional reference, which verifies that the assumptions utilized in the setpoint analysis are valid for cycle 6 operation was added. The changes requested in a CRN were also addressed.

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

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

PC/M 256-290 Supplement 1-2 SAFETY EVALUATION (continued)

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased.

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

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

The St. Lucie Unit 2 Cycle 6 reload design does not change the overall configuration of the plant. The mode of operation of the plant remains unchanged. Therefore, a possibility for a 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 verify that 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 bounded except the following; the Increased Main Steam Flow, the Pre-Trip Steam Line Break pin census, the Uncontrolled Control Element Assembly Withdrawal from Subcritical or Low Power Condition, and the Small Break LOCA event. The results of these reanalyses given in the RSE report show that in each case, the respective reference analysis bound the Cycle 6 specific results. Therefore, there is no reduction in the margin of safety relative to the Technical Specification basis.

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

PC/M 176-290 Supplement 0 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 from the Chemical 8 Volume Control System (CVCS). The affected piping sections extend from the discharge of the Boric Acid Makeup Tanks (BAMTs) to the Charging Pumps and the Volume Control Tank (VCT).

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

The portion of the CVCS System pertaining to this modification as defined by FSAR subsection 9.3.4, is classified as Safety Class 2 and Seismic Category I. Therefore, this PCM will be classified as Nuclear Safety Related.

The safety evaluation of this EP has shown that the implementation of this PCM does not constitute an Unreviewed Safety Question and requires no revision to the Unit 2 Technical Specifications. Prior NRC approval is not required for implementation. This PCM has no impact on 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 shall be deemed to involve an unreviewed safety question: (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased; or (ii) if the possibility for an accident or malfunction of 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 basis for any technical specification is reduced.

This modification removes affected thermal relief valves, their upstream/downstream piping, heat tracing cables and the thermal insulation from the CVCS System piping between the discharge of 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 Subsection 9.3.4, is classified as Safety Class 2 and Seismic Category I. Therefore, this PCM is classified as Nuclear Safety Related.

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

PC/M 176-290 Supplement 0 SAFETY EVALUATION (continued)

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

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

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

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

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

This modification has no adverse impact on the operability of the affected Reactivity Control System as addressed in the Section 3/4.1 of the Technical Specification. No changes to the 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 Technical Specifications.

PC/M 176-290 Supplement 0 SAFETY EVALUATION (continued)

The foregoing constitutes, per 10CFR50.59(b), the written Safety Evaluation which provides the bases 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.

PC/M 311-289 Supplement 0 ABSTRACT This Engineering Package provides the engineering and design details required to implement the replacement of the existing ionization smoke and duct detectors and the existing Main Fire Alarm panels. 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 are obsolete; and their replacements (installed as the originals failed) which are no longer manufactured. To ensure the reliability of the fire detection system, new ionization smoke detectors will be installed.

The existing panels are obsolete and spare parts are no longer available. The replacement panels represent the latest evolution in Honeywell's Fire Detection System's hardware and software. The new 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 is provided in areas that contain or present a fire hazard to equipment essential to safe plant shutdown. 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 an unreviewed safety question. The modifications described above have no adverse impact on plant operations 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 shall be deemed to involve an unreviewed safety question: (i) if the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased; (ii) if a possibility for an accident or malfunction of a different type than any evaluated 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.

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

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

PC/M 311-289 Supplement 0 SAFETY EVALUATION (continued)

Fire detection systems are provided in areas that contain or present a fire exposure to equipment essential 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 of equipment important to safety previously evaluated in the Safety Analysis Report is not

'ncreased by these modifications. The replacement of the obsolete detectors and panels enhances the operability of the equipment and the fire detection system. The new detectors and panels have the same characteristics as the existing equipment. The possible failure of this equipment will not prevent safety related equipment from performing their intended functions. The detectors and panels are not required during an accident condition. Therefore, the implementation of these modifications cannot increase the probability 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 a different type other than any previously evaluated. The detectors and panels are not required during an accident condition nor will they prevent safety related equipment from performing their intended functions. This modification does not affect any safety related equipment.

The margin of safety as defined in the bases for any Technical Specification is not reduced by this modification. The functions of the fire detection system that are controlled by the applicable 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 Technical Specifications.

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