ML17309A674

Rept of Changes Made to Facility Under Provisions of 10CFR50.59 for Period 901007 to 911006. W/920406 Ltr

ML17309A674
Person / Time
Site:	Saint Lucie
Issue date:	10/06/1991
From:	Sager D FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L-92-85, NUDOCS 9204100272
Download: ML17309A674 (108)

Text

ACCELERATED DICTRIBUTION DEMONS~TION SYSTEM 7 'I REGULATORY INFORMATION DISTR BUTION SYSTEM (RIDE)>>

ACCESSION NBR:9204100272 DOC.DATE: NOTARIZED NO DOCKET FACIL:50-389 St. Lucie Plant, Unit 2, orida Power 6 Light 'Co. 05000389 AUTH. NAME 'UTHOR AFFILIATION SAGER,D.A. , Florida Power & Light Co.

RECIP.NAME RECIPIENT AFFILIATION .

R

SUBJECT:

"St Lucie Plant Unit 2 Rept of Changes Made To Facility Under Provisions of 0 CFR 50.59 for period 901007 to 911006."

D DISTRIBUTION CODE: IE47D COPIES RECEIVED:LTR TITLE: 50.59 Annual Report of Changes, Tests or i ENCL / SIZE:

Experiments Made 3

W/out Approv 8 NOTES /

RECIPIENT COPIES 'RECIPIENT COPIES ID CODE/NAME PD2-2 LA LTTR ENCL 1

ID 'CODE/NAME PD2-2 PD-LTTR ENCL 0 5 5 NORRIS,J 1 0 D

INTERNAL: ACRS 6 6' AEOD/DOA 1 1 D

AEOD/DS P/TPAB NRR/DOEA/OEABll 1 1

1 N

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RGN2 FILE 01 1 1 EXTERNAL: NRC PDR 1 1 NSIC 1 1 D

D NOTE TO ALL "RIDS" RECIPIENTS:

D S

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

TOTAL NUMBER OF COPIES REQUIRED: LTTR 21 ENCL 19

P.O. Box 128, Ft. Pierce, FL 34954-0128 April 6, 1992 L-92-85 10 CFR 50.59 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 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 which were made, and are reportable, pursuant to 10 CFR 50.59. Included with the brief description of each plant change is a summary of the safety evaluation completed by Florida'Power &

Light Company for that plant change. This report includes plant changes completed between October 7, 1990, and October 6, 1991, and correlates with the information included in Revision 7 of the Updated Final Safety Report submitted under a separate cover.

Very truly yours, D. A. ger Vice r sident St. ie Plant DAS:JJB:kw cc: Stewart D. Ebneter, Regional Administrator, Region Senior Resident Inspector, USNRC, St. Lucie'lant II, USNRC Enclosure DAS/PSL 79657-92 U ~r l.i g

an FPL Group compant'Eq7 ff(

RE: St. Lucie Plant Docket No. 50-389 10 CFR 50.59 Report St. Lucie Plant Unit 2 Report of Changes Made To The Facility Under the Provisions of 10 CFR 50.59

, for the period October 7, 1990 to October 6, 1991 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.

~

9204100>72,

Plant Change/Modifications reportable pursuant to 10 CFR 50.59 for St. Lucie Unit 2 Number Su lemen 307-290 0 Annunciator Nuisance Alarms F45, G22, N6 and N23 275-290 0 FIS-14-6 Low Flow Alarm and Atmospheric Dump Valve Annunciator Deletion 268-290 0 BA Makeup Tank Low Level Alarm Select Switch 21 8-290 0 Polar Cran'e Magnatorque Indication 21 5-290 0 RCB Telescoping- JIB Boom Crane Anti-collision Device Modification 1 69-290 0 Modify RPS Cabinet Fan 1 67-290 0 Steam Generator Tube Plugging - CE Design Plugs 130-290 0 Main Feedwater Isolation Valve AFAS Override 080-290 0 Deflector Plate for EDG Crankcase Pressure Sensor 022-290 0 Removal of Startup Neutron Sources 009-290 0 Relocate Metal Building for Craft RCA Access Point 343-289 0 CEDM Generator Rheostat Adjustment 272-289 0 Feedwater Pump Recirculation Valve Control Modification 244-289 0 SGBD Valves Isolation Signal Override Modification 1 59-289 0 Containment Evacuation Alarm 103-289 0-1 ATWS - Diverse Scram System 035-289 1 ICW Piping and Restraints to FIS-21-9B 349-989 0-1 Above, Ground Waste Oil Storage Tank 073-989 0 Meteorological Replacement 355-288 0 Fisher Porter Integrator Obsolescence 230-288 0 Miscellaneous Restraint Modifications 227-288 0 Stabilize Analog Display System - CRT 199-288 0 Loose Parts Monitoring Vibration Amplifier Removal 085-288 0-1 Westinghouse Pressure Transmitter Replacement 399-988 1 Fuel Dispensing Facility 008-988 0-3 Visitors Facility and Training Center 095-287 0 EDG Fan Drive Shaft Guards 089-287 '1 Remote Reactor Vessel Level Indicator 043-287 0 Intake Cooling Water Pump 2A Self Lubrication Modification 109-987 0 Meteorological System Instrumentation 073-987 Replacement'isher 0-1 and Porter Transmitter Replacement 045-986 5 Underwater Intrusion Detection 018-285 2 ICW Pump Lubewater Strainer Replacement 046-285 0-2 DG Air Start System Relief Valve Replacement 107-284 0 DDPS/RTGB 204 Digital Displays 089-284 1 Temporary Replacement of SIGMA Meters

PC/M 307-290 Supplement 0 ABSTRACT This Engineering Package (EP) includes the engineering and design necessary to correct several annunciator nuisance alarms requiring set point modification as well as alarm deletions.

. By implementing this EP, these circuits will be consistent with the NUREG 0700 "Guidelines.

for Control Room Design Review" "Dark annunciator" concept. Under normal operation conditions, no annunciators will be'illuminated.

The Alarm circuits do not perform any Nucleai Safety Related functions. Seismically analyzed equipment/structures will be evaluated for any impact due to instrument removal. Therefore, this EP is classified as Quality Related.

The following annunciators are affected:

1) F45 "PWT/CST DEGASSIFIER LEVEL Hl/VACUUM LO".

2) G22 "CST N2 BLANKET PRESS. HI-HI/LO".

3) N6 "WASTE MGT. HEAT TRACE SYS. 2A/2B LOCAL ALARM".

4) "',

'N23 "CVCS BORIC ACID HEATING SYS. 2A/2B LOCAL ALARM".

A safety evaluation of this modification has been performed in accordance with 10 CFR 50.59. This evaluation iridicates 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, fo' 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 uhreviewed 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:

The probability'of occurrence and the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report are not increased by this modification because it does not affect. the availability, redundancy, capacity, or function of any equipment required to mitigate the effects of an accident. The annunciator circuits are non-nuclear safety related and serve no controlling functions. The deleted and revised annunciators are not used to mitigate the'effects of an accident and are not included in the list of Safety Related Annunciators on PSL-2 FSAR Table 7.5-3. The deletion of the pressure switch (PS-29-4-1) does not affect the seismic integrity of its respective instrumerit rack (IR-10-1A). Thus, the modification will not increase the probability of occurrence or consequences of the respective circuits affecting equipment important to safety.

PC/M 307-290 Supplement 0 SAFETY EVALUATION (Continued)

The possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report will not be created by this modification since no new failure. modes are introduced. The modification involves non-nuclear safety-related annunciator circuits and the failure of any items added by this modification will not impact any nuclear safety-related functions. A review of the failure mode analysis for CVCS (PSL-2 FSAR Table 9.3-9) and AFS (PSL-2 FSAR, Table 10.4.9B-2) has been performed and it has been determined that the results of these analysis have not been impacted by this modification. In addition, the overall seismic integrity of Instrument Rack IR-10-1A,will not be degraded by this modification.

iii) The margin of safety as defined in the basis for any technical specification is not affected by this modification. The modified and deleted alarm circuits perform non-nuclear safety related functions and are not included in the bases of any technical specification.

The annunciator circuits do not perform a safety-related function. The annunciators are not included in the list of Safety Related Annunciators on PSL-2 FSAR, Table 7.5-3. A review of the failure mode analysis for CVCS (PSL-2 FSAR, Table 9.3-9) and AFS (PSL-2 FSAR, Table 10.4.9B-2) has been performed for this modification and it has been determined that no new failure modes have been introduce'd to the plant. The modification required a review to determine the effects on the seismic integrity of 1R-10-1A due to the removal of PS-29-4-1.

Accordingly, this engineering package has been classified as Quality Related.

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 275-290 Supplement 0 ABSTRACT This Engineering Package authorizes the deletion of two control room annunciator functions, which have resulted in numerous nuisance alarms in the past. The CCW low flow alarm from the letdown heat exchanger (FIS-14-6) will be deleted. Secondly, the annunciator window function, which provides indication that the ADV's are in the "manual" position will be deleted.

The St. Luice FSAR section 9.2.2 describes the CCW system and its design bases. FSAR Tables 7.5-1, 9.2-7 and Figure 9.2-2 address the function of FIS-14-6, which provides the CCW low flow alarm. FSAR Table 7.5-3 addresses the ADV's "Manual" position annunciation on annunciator windows LA, LB-12. The modifications authorized by this Engineering Package do not alter the original design bases as described by the FSAR; however, FSAR Tables 7.5-3,.

9.2-7 and Figure 9.2-2 will require revision as outlined in,P'ttachment 7.3.

The safety classification'f this package is designated "safety-related" because the subject alarm circuits are designated "ASA or ASB", which indicates that they are associated with safety. Furthermore, the LA, LB-12 annunciators are listed as safety related annunciator windows by FSAR Table 7.5-3.

The safety evaluation of this modification has determined that this Engineering Package does not affect plant safety or operation and does not involve an unreviewed safety question or a change to the Technical Specifications. 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 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.

The following evaluation 'serves to determine whether this modification constitutes an unreviewed safety question according to 10 CFR 50.59 criteria:

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

The proposed modification provides for deletion of two annunciator functions, ADV's "Manual" position and CCW low flow. Deletion of these annunciator functions does not change the operation of ADV's, CVCS or any other component. Therefore, the probability of occurrence of an accident previously evaluated in the Safety Analysis Report is not increased.

PC/M 275-290 Supplement 0 SAFETY EVALUATION (Continued)

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

The consequences of an accident previously evaluated in the Safety. Analysis Report will not increase because the deletion of subject annunciator functions will not prevent any safety related system, structure or component in mitigating the consequences of an accident.

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

The proposed modification will not affect the function of ADV's or CVCS or any other equipment important to safety. Therefore, the probability of occurrence of a malfunction of equipment important to safety'reviously evaluated in the Safety Analysis Report is not increased.

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

I The proposed change involves deletion of annunciator functions only and it does not change or prevent ADV's and CVCS from performing their intended function.

Therefore, the consequences of a malfunction of equipment important to safety is not increased by this modification from that previously evaluated in the Safety Analysis Report.

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

Technical Specification 3.7.1.7 requires all ADV's to be in the manual position above 15% power. Annunciation of ADV's "Manual" position is not required/desired since it is not an alarm, condition. In case of CCW low flow alarm, there is other instrumentation available (high letdown HX outlet temperature), which provides similar warning of inadequate cooling of letdown flow through the HX.

The proposed modification does not change the operation, function or design bases of ADVs and CVCS as described in the Safety Analysis Report and does not adversely affect any other Safety Related structure, system or component.

Therefore, there is no possibility that an accident may be created that is different from one already evaluated in the Safety Analysis Report.

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?

Deletion of the subject annunciator functions will not introduce any new failure mode to the operation of ADVs, CVCS or any other Safety Related system, structure or component. Therefore, probability of a malfunction of equipment important to safety which is of a different type than previously evaluated in the Safety Analysis Report is not created.

PC/M 275-290 Supplement 0

'AFETY EVALUATION (Continued)

7) Does the proposed change reduce the maigin of safety as defined in the basis for any Technical Specification/

C Technical Specification 3.7.1.7 requires all'ADVs to be in manual control above 15% of power. Technical Specification 3/4.7.1.7 further states that the limitation on maintaining the ADVs in the manual mode of operation is to ensure that the ADVs will be closed in the event of a steam line break. However, this EP only provides for elimination of the ADVs "manual" position annunciation since it's not an alarm condition and it is not required by Technical Specifications.

This modification does not reduce the margin of safety as defined in the .

Technical Specification because the annunciator functions, which are to be deleted, are not required to notify the operator of a condition which may.reduce the margin of safety and are not required to fulfill any Technical Specification requirement/surveillance.

CONCLUSION The modification associated with this EP has no adverse effect on safety related components or systems, and does not constitute an unreviewed safety question or require changes to the Plant Technical Specifications. Therefore, the modification outlined in this Engineering Package can be implemented without prior NRC approval.

PC/M 268-290 Supplement 0 ABSTRACT This Engineering Package provides the necessary details for the permanent modification to the

,BAM Tank Low Level Alarm Select Circuit. The BAM Tank L'ow Level Alarms provide annunciation in the Control Room (Windows N7 and N8) when the associated tank level (2A or 2B) approaches the Tech. spec. 3.1.2.8 limit. As only one tank is required to be maintained above this level, operations normally maintains the other tank at levels below the low level setpoint, for normal plant operation purposes. This keeps the associated annunciator window constantly energized. A select switch is to be added to select the tank to be maintained per Tech. Specs. This switch will defeat the low level alarm of the non-selected tank. The associated low level annunciator. will no longer be energized. By implementing this EP, this circuit will be consistent with the NUREG 0700 "Guidelines for control Room Design Review" "Dark Annunciator" concept. This Engineering Package will provide the circuit modification, and select switch addition necessary to permanently install the BAM Tank Low Level Alarm Defeat Circuit.

The BAM Tank Low Level Alarm Circuit does not perform any Nuclear Safety Related function.

The alarm circuit does provide Control Room Annunciation to Operations informing them that the associated BAM Tank level is approaching Tech. Spec. required limits. The select switch

~ added by this modification shall be seismically mounted. Therefore, this Engineering Package is classified as Quality Related.

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 reduce'd. The modification included in this engineering package does not involve an unreviewed safety question because of the following reasons:

PC/M 268-290 Supplement 0 SAFETY EVALUATION (Continued)

The probability of occurrence and the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report are not increased by this modification because it does not affect the availability, redundancy, capacity, or function of any equipment required to mitigate the effects of an accident. The BAM tank (2A and 28) level indication and alarm circuits are non-nuclear safety related and serve no controlling functions. The modification will meet the seismic requirements of RTGB-205 and will not degrade the seismic qualification of the panel. The modification will meet the. electrical separation requirements for RTG 8-205.

The possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report will not be created by this modification since no new failure modes'are introduced. The modification involves non-nuclear safety-related circuits and the failure of any items added or deleted by this modification will not impact'any nuclear safety-related functions. A review of the failure mode analysis for CVCS (PLS-2 FSAR Table 9.3-9) has been performed and it has been determined that the analysis has not been impacted by this modification. In addition, the overall seismic qualification of RTGB-205 will not be degraded by the addition of this modification.

The margin of safety as defined in the basis for. any technical specification is not affected by this modification. The modified BAM tank level and alarm circuits perform non-nuclear safety related functions and are not included in the bases of any technical specification.

The BAM tank (2A and 28) level indication and alarm circuits do not perform a safety-related function. The alarm circuits do provide Control Room annunciation, informing operations that the respective BAM tank level is approaching technical specification required limits. A review of the failure mode analysis for CVCS (PSL-2 FSAR, Table 9.3-9) has been performed for this modification and it has been determined that no new failure modes have been introduced to the plant. The modification requires. seismic mounting of the select switch within RTGB 205.

Accordingly, this engineering package has been classified-as Quality Related.

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 218-290 Supplement 0-ABSTRACT This Engineering Package provides the necessary details for the permanent modification to the Reactor Building Polar. Crane. The RB Polar Crane is located in the Reactor Containment Building above the Reactor Vessel. This modification will add main hoist magnatorque current, indication remotely at the polar crane operator area. The magnatorque functions as a'n electric load brake for the main hoist. This. indication will provide the polar crane operator with added assurance of the operability of the magnatorque. This Engineering Package will provide the circuit modification, cabling, conduit, instrument panel and indicator necessary to permanently install the magnatorque indication. This indication will be fused at the control panel to minimize the possibility. of a fault in the new circuit affecting the magnatorque control circuit.

. The Reactor Building Polar Crane does not perform any Nuclear Safety-Related function, but is seismically designed. The additional conduit and panel added by this modification must be designed for seismic requirements. Therefore, this Engineering Package is classified as Quality Related..

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 the 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

'h ~

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:

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 'are no't increased by this modification because it does not affect the availability, redundancy, capacity, or function of any equipment required to mitigate the effects of an accident. The RCB Polar Crane is classified as Quality Related and has been designed for seismic loading. The -

Polar Crane is not required to mitigate the effects of an accident. The modification will not increase the probability of occurrence or consequence of the crane affecting equipment important to safety.

PC/M 218-290 Supplement 0 SAFETY EVALUATION (Continued)

The possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report will not be created by this modification since no new failure mode's are introduced. The modification involves non-nuclear safety related structures and failure of any items added or deleted by this modification will not impact any nuclear safety-related functions. In addition, the overall seismic design of the polar crane will not be degraded by the addition of this modification.

I iii) The margin of safety as defined in the basis for any technical specification is not affected by this modification since the components involved in are not included in the bases of any Technical Specifications. this'odification The RCB Polar Crane does not perform a safety-related function however failure of the system-could potentially interact with other safety related equipment. A failure mode analysis review has been performed for this modification and it has been determined that no new failure modes have been introduced to the plant. However, the modification does require seismic considerations. Accordingly, this engineering package has been classified as Quality Related.

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 '215-290 Supplement 0 ABSTRACT I

This Engineering Package provides the necessary details for the permanent modification to the Telescoping Jib Boom Crane located in the Reactor Containment Building above the 2A steam generator by eliminating the photoelectric cells which are used to prevent possible collision during clockwise/counter-clockwise motion of the crane boom. These sensors normally stop boom rotation whenever the infrared beam from any of the six photoelectric sensors is interrupted. Sensor lens occlusion in conjunction with misalignm'ents created by 'crane movements contribute to'erroneous collision detection signals, thus hindering productive use of the crane. To date, Jumper 5 Lifted Lead requests have been dispositioned to allow bypassing the circuit for use of the crane without the photoelectric cells. This Engineering Package will modify the circuitry to permanently disconnect the photoelectric cells.

The'modification will not impact the operability of the anti-collision proximity switch used to prevent collision with the Containment fan cooler ductwork when operating the boom in the forward telescoping direction.

The Telescoping Jib Boom Crane does not perform any Nuclear Safety-Related function but is seismically designed to prevent any adverse effect to any system or component important to safety. Thus the crane is Quality Related. This modification, however, in no way affects the seismic design of the crane. Therefore this Engineering Package is classified as Not Nuclear Safety; A safety evaluation for 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 the. 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 modifications included in this engineering package do not involve an unreviewed safety question because of the following reasons:

PC/M 215-290 Supplement 0 SAFETY EVALUATION (Continued)

The probability of occurrence and 'the consequences of an accident or malfunction of, equipment important to safety previously evaluated in the Safety Analysis Report are not increased by this modification because it does not affect the availability, redundancy; capacity, or function of any equipment required to mitigate the effects of an accident. In accordance with the SAR, the Telescoping Jib Crane has been'designed to meet the requirements of NUREG-0612. In turn, NUREG-0612 specifies that cranes be designed to CMAA-70 which does not require the use of anti-collision devices, other than bridge and trolley stops, to prevent collisions with obstacles during operation.

Thus, the anti-collision device for clockwise and counter-clockwise boom motion was included for equipment protection and personnel-safety and was not intended to reduce the probability of occurrence or the consequences of an analyzed accident. Therefore the removal of the anti-collision devices will not impact the. NUREG-0612 compliance of the crane. Therefore the probability of occurrence or consequences of the modified crane affecting equipment important to safety will remain unchanged by this modification.

The possibility of an accident or malfunction of a different type than any

'valuated previously in the Safety Analysis Report will not be created by this modification because the modification involves non-nuclear safety related

~

equipment and failure of any items due to this modification will not impact any nuclear safety-related functions. In addition, since NUREG-0612 and CMAA-70

=-

do not require the use of the anti-collision device, removal of the photoelectric cells will not cause an accident or malfunction of any structure, system, or

'omponent important to Nuclear Safety.

. iii) The margin of safety as defined in the. bases for any technical specification is not affected by this modification since the components involved in this modification are not included in the bases of any Technical Specifications.

The Telescoping Jib Boom Crane does not perform any Nuclear'Safety-Related function but is seismically designed to prevent any adverse affect to any system or component important to safety. Thus the crane is Quality Related.- A failure mode evaluation has been performed for this modification and it has been determined that no new failure modes have been introduced to the plant. Additionally, this modification in no way affects the seismic design of the crane. Accordingly this engineering package has been classified as Not Nuclear Safety.

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 169-290 Supplement 0 ABSTRACT This Engineering Package (EP) provides the necessary documentation for the Reactor Protective System (RPS) cabinet cooler assemblies modification to reduce the noise emanating from the existing cooler assemblies.

The RPS cabinet assembly is a four (4) bay cabinet consisting of basically four (4) bays within one welded frame. The four (4) cabinet bays are mechanically separated from each other by fireproof mechanical barriers. The RPS cabinet has four (4) cooler assemblies, one for each bay. The cabinet cooler assembly is racg mounted at the bottom of each bay and contains two axial fans to draw in fresh filtered air to the interior of the bay. The noise emanating from these fans is bothersome and distracting to the operator sitting directly in front of the RPS cabinet (HED Serial Number 816)..This modification consists of replacing the existing cooler assemblies with new cooler assemblies designed for mounting in the rear of the cabinet.

The RPS cabinet houses instruments and equipment necessary to monitor selected nuclear steam supply system conditions and to effect reliable and rapid reactor shutdown and is safety related. This Engineering Package is therefore classified as safety related.

A review of the changes to be implemented by this PC/M was performed in accordance with 10 CFR 50.59. As indicated in the Safety Evaluation (Section 3.0), the modifications provided by this Engineering Package do not have an adverse effect on plant safety, security or operation, do not constitute an unreviewed safety question, and do not require changes to the technical specifications. Therefore, prior NRC approval for implementation is not required.

Revision one (1) of this PC/M is issued to remove the Hold Points for approval of the seismic and environmental qualification analysis. Special Instructions for implementation of this modification (Section 9) and Affected Drawing List (section 11) are, also revised to include the new drawings created by this PC/M.

This revision does not affect the PC/M Design Bases or Safety Evaluation and does not change Plant Technical Specifications. Therefore, prior NRC approval for implementation is not required.

SAFETY EVALUATION DESCRIPTION AND PURPOSE This Engineering Package (EP) provides the necessary documentation for the Reactor Protection System (RPS) Cabinet Cooler assemblies modification to reduce the noise emanating from the existing cooler assemblies.

PC/M 169-290 Supplement 0 SAFETY EVALUATION (Continued)

The RPS cabinet assembly is a four (4) bay cabinet consisting of basically four (4) bays within one welded frame. The four (4) cabinet bays are m'echanically separated from each other by fire proof mechanical barriers. The RPS cabinet has four (4) cooler assemblies, one for each bay. The cabinet cooler assembly is rack mounted at the bottom of each bay and contains two axial fans to draw in fresh filtered air to the interior of the bay. The noise emanating from these fans is bothersome and distracting to the operator sitting directly in front of the RPS cabinet. This PC/M will reduce the fan noise by performing the following modifications on each of the four RPS cabinet cooler assemblies:

The panel and grille associated with the existing fans will be replaced with a plain panel containing the two power switches and two indicating lights for their respective fans (Drawing E-83090-489-006, Rev. 2). Terminal Blocks and connectors will be located behind the panel for wiring.

The existing cooler assemblies will be replaced with new cooler assemblies designed for mounting in the rear of the cabinet. The installation utilizes two new mounting arrangements (Drawing E-83090-489-004, Rev. 2 and D-83090-489-005, Rev. 2):

The first arrangement consists of mounting an air inlet plenum to the rear doors which allows the fans to draw in air through the existing door louvers. In addition, this plenum will house the air filter.

The cabinet portion of the assembly will be mounted by means of angles to the- existing cabinet structure. This assembly will house the two fans which will draw air from the plenum into the fan housing and blow the air into the cabinet enclosure. A sound baffle will. be located directly in front of the fan inlets to absorb the noise generated by the fans. The fan housing will be lined with the same sound absorbing material to aid in preventing the noise from entering the control room.

Also this assembly will house the air flow switches located at the fan outlets, to detect the loss of cooling air. Finally all cooler assembly-electrical connections will be made via two MS style connectors. This design will facilitate rapid removal of the cooler assembly as needed during maintenance activities.

ANALYSIS OF THE EFFECTS ON SAFETY Reactor Protective System cabinet houses instruments and equipment necessary to monitor selected nuclear steam supply conditions and to effect reliable and rapid reactor shutdown, and is safety related. The function of cooler assemblies is to maintain cabinet internal temperatures within the limitations of the original RPS Cabinet environmental qualification test, thus ensuring the RPS circuits reliably perform their safety related functions.

PC/M 169-290 Supplement 0 I SAFETY EVALUATION (Continued)

As discussed in detail in the environmental analysis (Reference 6.18) the redesigned cooler assemblies have the same volumetric air flow's the existing cooler assemblies. With the air discharge velocity and orientation of the fa'n discharge housing, the air stream will be utiliz'ed at least,

. as effectively as the existing cooler arrangement. As before, the perforated top covers of the'rack mounted chassis and integral blowers of the power supply assembly are utilized to dissipate heat to the cabinet free volume, where it. is then, removed by the air flow genera'ted by the cooler assembly. In summary the redesigned cooler assemblies will maintain internal cabinet temperatures within the limitations of the original'PS cabinet environmental analysis.

The original cooler assembly design utilized dual redundant fans both of which were powered from the safety grade measurement channel vital bus associated with that bay.'ach fan was sized to provide 100% of the required cooling air flow, and the RPS cabinet environmental qualification test verified this aspect of the design. An air flow switch was installed in the discharge housing of each fan so the operator could verify, via the front panel mounted status lights controlled by the air flow switches, if the fans were operating properly. The redesigned cooler assemblies incorporate these three, features (i.e., dual 100% flow fans with flow switches and status lights). With this design no single failure of the cooler assembly can be'ostulated that will result in a loss of both cooling fans. If the common fan power supply is lost all of the 'electrical equipment in that RPS bay will also be lost. Loss of one RPS channel is bounded by the existing safety analysis. In summary the redesigned cooler assemblies continue to meet the single failure design criteria.

The new cooler assemblies are seismically and environmentally qualified (Reference 6.18). They are safety related and have been purchased QL-1. Therefore electrical separation between the.fan motor power circuits and measurement ch'annel vital bus within each RPS bay is not required. The cooler assemblies are seismically mounted in the RPS cabinet to preclude interaction with other'afety related equipment.

In conclusion, the new cooler assemblies will provide adequate cooling, are as reliable as the existing assemblies, and have been designed to preclude adverse interaction with other equipment.

Therefore this modification has no adverse effects on the safety related functions of the RPS.

FAILURE MODES AND EFFECTS ANALYSIS The functional design of the new cooler assemblies is equivalent in all respects to the existing cooler design. Both designs utilize dual fans each sized to provide 100% of the required air flow; Both designs include air flow switches mounted in the fan discharge housing and front panel mounted fan running status lights. In fact the new cooler assembly design uses the same fan and flow switch model numbers as was used in the existing coolers. The original design, for the fan power supply circuit has also been retained (i.e., both fans are powered from the measurement channel vital bus which also supplies all other electrical loads in that bay)., Due to this design similarity most of the possible failure modes and the effects of those failures will not be changed by this modification.

However for completeness an analysis of any single failure resulting in a.loss of air flow through one fan will be presented. In addition the potential for, and the consequences of, a mechanical failure of the fan discharg'e housing will be discussed.

0 PC/M 169-290 Supplement 0 SAFETY EVALUATION (Continued)

'oss of Air flow through a single fan could be caused by interruption of fan power "or mechanical binding. Once air flow through the fan was less than the setpoint. of the air flow

, switch the red fan running status light mounted on the RPS front panel would extinguish. The operator would be able to observe this failure and tnitiate corrective maintenance. Meanwhile the redundant fan would continue to provide 100% of the necessary cooling capacity. In summary any failure that results in loss of air flow through a single fan will have no adverse effect on the RPS safety functions.

The new cooler assemblies will be mounted in the rear of each RPS bay just above floor level.

In this location a new failure mode is conceivable that could potentially disable both cooling fans. That is, during maintenance work on the RPS cabinets, someone steps on the fan housing and exerts enough force on it that the fan housing integrity is lost and the fans are disabled. As discussed in section (2.3.10) mounting the fan housing on the cabinet's unistrut has been designed for each and fast removalfinstallation. The design has recognized the fact that the modified cooler assembly would reduce the access area in the rear of cabinet bay and has foreseen occasions when the fan housing would be in the way. of performing work in the cabinet. For this reason the design has made it possible to remove the fan housing within a few minutes and reinstall it back when the work is done, thus avoiding stepping on the fan housing. In fact there will be a sign on the fan housing to read "DO NOT STEP ON". In addition, the fan housing is designed to have a strong structure, (i.e., it utilizes 0.1" thick aluminum sheet metal). For all of these reasons this potential failure mode has a very low probability of occurrence. In addition even if loss of both cooling fans in one RPS bay resulted in loss of that RPS channel, this consequence is still bounded by the safety analysis.

EFFECT ON TECHNICAL SPECIFICATIONS As described in detail in section 3.3, the functional design of the cooler assembly has not been changed. The cooler assemblies location and configuration within the Reactor Protective System Cabinet are not described in the Technical Specifications. Therefore, this modification has no effect on the Technical Specifications.

UNREVIEWED SAFETY QUESTIONS DETERMINATION 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 of 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 or require a change to the Technical Specifications:

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

PC/M 169-290 Supplement 0 SAFETY EVALUATION(Continued),

The probability of occurrence of an accident previously evaluated in the FSAR will not increase because this modification does not affect any equipment whose malfunction is postulated in the FSAR to initiate an accident.

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

The Reactor Protective System is'designed to mitigate the consequences of an accident by providing reactor tiip signals when Limiting Safety System Setpoints are exceeded. The function of the RPS Cabinet Cooler assemblies is to maintain cabinet internal temperatures within limitations of the original RPS cabinet environmental qualification test. It has been determined that the modified coolers will perform this function as effectively as the existing cabinet coolers, thus ensuring overall system reliability. This modification has no effect on any other system designed to mitigate the consequences of an accident. Therefore, consequences of an accident previously evaluated in the FSAR will not increase.

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

As discussed in detail in section 3.3, air flow through both cooler assembly fans would have to be lost before any adverse effect on safety related equipment would result. The only new postulated failure mode that could compromise both fans has such a low probability of occurrence that it is not considered to be a credible failure mode. Therefore, the probability of'an occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR will not increase.

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

As detailed in section 3.3 the consequences of the worst case postulated failure are limited to loss of airflow through a single fan which is bounded by the safety analysis. Therefore, this modification does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR.

1 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 operation, function or design bases as described in the FSAR and does not adversely affect any other safety related structure, systems or components. Therefore, this modification does not create the possibility for an accident of a different type than any previously evaluated in the FSAR.

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 169-290 Supplement 0 SAFETY EVALUATION(Continued)

This modification has no adverse effect on either the physical or electrical separation that currently exists between RPS cabinet and all other safety related equipment. Within RPS cabinet, the modified cooler assemblies utilize fans and air flow switches that are the same'make and model as the existing fans and air flow switches. Therefore the possibility of a malfunction of equipment important to safety, of a different type than any previously evaluated in the FSAR is not created.

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

This modification does not change the design bases, functions or operations of any safety related equipment and does not adversely affect any other safety related structures, systems and components. Therefore, this modification does not reduce the margin of safety as defined in the bases for the'Technical Specifications.

PLANT RESTRICTIONS The mode restriction for implementation of the PC/M, shall be in accordance with St. Lucie Unit 2 Technical Specification Table 3.3-1.

CONCLUSIONS This modification has been reviewed against the requirements of 10 CFR 50.59 and has been found to be acceptable, thus prior NRC approval is not required for implementation. This acceptability is based on the fact that operation in accordance with this safety evaluation does, not require a change to the Technical Specifications nor does it constitute an unreviewed safety question.

PC/M 167-290 Supplement 0 ABSTRACT This PC/M documents Engineering review and concurrence for the use of Combustion Engineering expanded type plugs in the St. Lucie Unit 2 steam generators. This PC/M also provides the information necessary to as-build affected documents.

Since the steam generator tubes are nuclear safety related, the tube plugs described herein are also nuclear safety related.

Based upon a failure mode evaluation and 10 CFR 50.59 review, this modification does not involve an unreviewed safety question nor require changes to the technical specifications.

Therefore, prior NRC approval is not required for implementation of this modification. The modification has no adverse affect on plant safety or operability.

SAFETY EVALUATION This modification involves documenting the maintenance practice of plugging steam generator tubes. Steam. generator tubes are nuclear safety related, therefore this engineering package is classified as nuclear safety related. The PC/M provides engineering concurrence for the use of the Combustion Engineering expanded tube plug design (previously utilized on the Unit 2 steam generators), the required 50.59 review of the modification, and the information required for update of affected-documents.

10 CFR 50.59 allows' change to a nuclear facility without prior NRC approval if an unreviewed safety question does not exist and if changes to Technical Specifications are not involved. The followin'g arguments demonstrate that an unreviewed safety question does not exist relative to this modification:

The probability of occurrence of a design basis accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased since this modification does not decrease the design margin of the RCS pressure boundary (the tube plugs meet or exceed all design requirements for ASME Section III, Class 1 components).

1 The consequences of a previously postulated design basis accident or malfunction of equipment important to safety previously evaluated in the FSAR are not made more severe for the same reasons given in (i) and since no

'existing accident mitigation equipment or systems are altered by this modification.

iii) The possibility of an accident of a different type than previously addressed in the FSAR does not exist since no new systems or equipment are introduced-by this modification. Failure of a tube plug would be no more severe than a steam generator tube rupture, a previously evaluated condition. Therefore, no new accidents are created.

PC/M 167-290 Supplement 0 SAFETY EVALUATION iv) The margin of safety as defined in the basis for any technical specification is not reduced since the total number of tubes plugged in the steam gener'ators following this modification is less than assumed in the FSAR Chapter 1.5 analysis.

Since the above arguments demonstrate that an unreviewed safety question does not exist, and since a revision to the Technical Specifications is not required, the addition of the Combustion, Engineering tube plugs to the Unit 2 steam generators does not require prior NRC approval.

PC/M 130-290 Supplement 0 ABSTRACT

'This Engineering Package (EP) provides for the modification of control circuits of valves HCV-09-1A, HCV-09-1B,,HCV-09-2A and HCV-09-28. The purpose of this mo'dification is to provide override capability to allow reopening these containment isolation valves after an Auxiliary Feedwater Actuation Signal (AFAS) closure signal to provide a flow path for adding water to the Steam Generator in a Total Loss of Feedwater (TLOF) event as defined by FPL Emergency Operation Procedure 2-EOP-06, Total Loss of Feedwater. This override will only be used as defined by 2-EOP-06 and after all attempts to restore Auxiliary Feedwater have failed. This modification will not affect the Main Steam Isolation Signal (MSIS) for the feedwater isolation valves.

This EP will involve, for each valve, the addition of auxiliary relays and the rewiring of each control circuit; The existing control switches can be utilized; however, new switch escutcheon plates will be required.

The modification in this EP. is to the Main Feedwater Isolation Valves (MFIVs) which perform a containment isolation function and are defined as Seismic Category I by FSAR Section 10.4.7 and Nuclear Safety Related per FSAR Section 6.2. Due to the involvement with a Nuclear Safety Related function (Containment Isolation), this EP has been classified as Nuclear Safety Related.

The present control circuits will be modified such that the AFAS closure signal may be overridden through the addition of a CLOSE/OVERRIDE position on the valve control switches in place of the existing CLOSE position. In order to utilize the, override feature, with an AFAS signal present, the control switch must first be placed in the CLOSE/OVERRIDE position prior to being placed in the. OPEN position. The first action will arm the override feature and the second will energize the valve coil and open the valve.

Once the closure signals have been overridden, a new AFAS isolation signal will cause the valves to close.

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 require a change to the Technical Specifications and does not reduce the margin of safety for any Technical Specification.

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 unresolved safety question (i) if the probability of occurrence oi consequences of an accident or malfunction of equipment important to safety pieviously evaluated in the Safety Analysis Report may be increased; or (ii) if the possibility of an accident or malfunction of a different type other 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.

PC/M 130-290 Supplement 0 SAFETY EVALUATION (Continued)

This Engineering Package (EP) provides the engineering and design necessary to modify the Main Feedwater Isolation Valve (MFIV) control circuits to (1)'provide Auxiliary Feedwater Actuation Signal (AFAS) override c'apabilities in an off-normal/post-accident scenario, (2)

'e'quire two distinct and deliberate operator actions in order to accomplish the override feature and, (3) provide positive indication, via auditory and visual annunciation on the RTGB-206, that (a) Auxiliary Feedwater Actuation Signal is present and, (b) the valve solenoid is energized (i.e., the MFIV is open with an AFAS present). The control switch escutcheon will be re-engraved to reflect override capabilities.

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Two annunciator windows will be added to provide control room indication of the override status; each steam generator feed header has one annunciator for the two associated isolation valves. They will provide indication that an'AFAS is present and that the subject valve(s) is open. The location of these annunciators will be RTGB-206 windows P-46 and P-56.,

I This modification Ieill provide a flow path for feeding water to the Steam'Generators in the event of a Total Loss of Feedwater (TLOF) event as defined by FPL Emergency Operating Procedure 2-EOP-06, Total Loss of Feedwater; This override will only be used as defined by 2-EOP-06 and after all attempts to restore Auxiliary Feedwater have failed. This modification will not affect the Main Steam Isolation Signal (MSIS) for the feedwater isolation valves.

The'modifications included in this Engineering Package do not involve an, unreviewed safety question because:

The probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased since the use of the override capability will be utilized only in an off-normal/emergency operating condition and with strict adherence to the governing procedure (2-EOP-06, Total Loss of Feedwater). The system's normal operation is unaffected by this modification.

The possibility for an accident or malfunction of a differe'nt type than any previously evaluated in the Safety Analysis Report is not created because components involved with this modification introduce no new type of accident and cannot cause malfunctions of any safety related equipment. No changes have been made to the normal operational design of any control circuits or associated systems which are important to safety. A review of accidents or malfunctions concerning the Main Feedwater System shows,no change to existing analysis due to the implementation of this modification.

iii) This modification does not change the margin of safety as defined in the basis for any Tech'nical Specification since the main feedwater system is not part of any Technical Specification bases. The containment isolation function for the MFIVs as required by Technical Specification 3/4.7.1.6 is not altered.'he AFAS portion of the circuit will still function in accordance with the Technical Specification.

PC/M 130-290 Supplement 0 SAFETY EVALUATION (Continued)

This package is classified as Nuclear Safety Related, since the containment isolation functions of the MFIVs are interlocked with the Engineered Safety Feature Actuation System (ESFAS) and are used to mitigate the consequences of an.ac ant.

Implementation of this Nuclear Safety Related PC/M uoes not require a change to the Plant Technical Specifications.

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

0 PC/M 080-290 Supplement 0 ABSTRACT The St. Lucie Unit 2, 2A2, 12 cylinder diesel engine had a problem with false crankcase over-pressurization alarms. Analysis indicated this problem was attributed to oil splashing against the diaphragm of.the pressure sensor. To resolve this condition, a deflector plate assembly (deflector plate), supplied by EMD, the engine vendor, was installed in front of the crankcase pressure sensor. This modification has corrected the problem and has not adversely affected the engine's operability.

This Engineering Package (EP) performs the same modification to the other Unit 2 diesel engines (2A1, 2B1, and 2B2) as a precautionary measure.

The crankcase pressure sensor does not perform a safety related function. The pressure sensor causes a generator lockout (trip) at high crankcase pressure only during test situations, Test situations for the diesel generators are defined as diesel operation in the absence of a SIAS, CSAS, CIAS or Loss of Offsite Power (LOOP). The pressuie sensor is considered=

quality related because it is attached to the safety related diesel engine. The deflector plate installed by this EP is similarly classified as quality related because it performs no safety related function and is attached to the diesel engine.

The modification to the Unit 2, 2A2, 12 cylinder diesel engine was documented in PC/M 113-286. That PC/M was classified as safety related.

Installation of the deflector plate does not affect plant safety or plant operation and therefore, does not require a change to the Technical Specifications.

Section 3.0 of this EP demonstrates that an unreviewed safety question does not exist and prior NRC approval for implementation is not required.

SAFETY EVALUATION DESCRIPTION AND PURPOSE The proposed modification installs a deflector plate assembly in front of the Unit 2 2A1, 2B1 and 2B2 diesel engine crankcase pressu're sensors. The purpose of the deflector plate is to

. prevent false crankcase over-pressurization alarms caused by oil splashing against the diaphragm of the crankcase pressure sensor.

ANALYSIS OF EFFECTS ON SAFETY Installation of the deflector plate assembly has no effect on safety. The deflector plate prevents false crankcase over-pressurization alarms caused by oil splashing against the diaphragm of the crankcase pressure sensor. The deflector plate is classified as quality related because it is attached to the safety related diesel engine. It does not perform a safety function and does not interfere with the operation of the crankcase pressure sensor, which is also classified as quality related. A high crankcase pressure signal causes a generator lockout when the diesel generator is run in a test situation. During an emergency start situation, the high crankcase pressure signal is blocked (Reference 6.1).

PC/M 080-290 Supplement 0 SAFETY EVALUATION (Continued)

FAILURE MODES AND EFFECTS ANALYSIS The deflector plate assembly, design accounts for the following failure modes:.

Seismic The, deflector plate assembly is designed to meet or exceed the original diesel engine seismic criteria. Morrison-Knudsen Company has supplied a certificate of conformance .

to FPL ensuring the seismic qualification of the part. A copy of the certificate of conformance is included with this EP as Attachment.7.11 ~

~Corr sion The deflector plate assembly and the engine crankcase are both constructed of carbon steel (Refer'ence 6.15, Attachment 7.5). The composition of the deflector plate components are described in "section 2.1.1.2 of this EP. Since the deflector plate material, is similar to other diesel engine components, there will be no galvanic interaction and the plate will be compatible with the engine lubricating oil.,The bolts.

that secure the pressure, sensor, also secure the deflector plate. Because these bolts are compatible with the existing engine configuration, they will be compatible with the deflector plate.

Based on these design'aspects, the addition of the oil deflector plate will not adversely affect any safety related component of the diesel engine.

/

EFFECT ON TECHNICAL SPECIFICATIONS This modification does not affect the basis for any Technical Specification in section 3/4.8, Electrical Power Systems, and therefore does not reduce the margin of safety as defined in the basis for any technical specification.

UNREVIEWED SAFETY QUESTION DETERMINATION 10 CFR 50.59 allows changes to a facility as described in the FSAR, without prior NRC approval, if an unreviewed safety question does not exist and if a change to the Technical Specification is not required. Based on these requirements and information supplied in the design analysis,-the following conclusions can be made with regard to an unreviewed safety question:

1) Does the proposed change increase the probability of occurrence of an accident previously evaluated in the safety analysis report?

The St. Lucie Unit 2 LOOP analysis addresses the failure of one diesel generator. The probability of a diesel engine failure is not increased by this modification because the deflector plate is seismically qualified and constructed

, 'of materials compatible with the crankcase environment.

PC/M 080-290 Supplement 0 SAFETY EVALUATION (Continued)

Based on the above, 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 safety analysis report?

The consequences of a diesel engine failure is not increased as a result of this modification because the deflector plate is installed internal to the engine and does not affect any other accident mitigation system.

Based on the above, the proposed change does not increase the consequences of an accident previously evaluated in the SAR.

3) Does the proposed change increase the probability of occurrence of a malfunction of equipment previously evaluated in the safety analysis report?

The deflector plate assembly provides the same level of dependability (seismic, corrosion) as the existing pressure sensor. Therefore, the probability of a malfunction is equivalent to the existing configuration.

Based on the above, the probability of occurrence of a malfunction of equipment important to safety is not increased.

4) Does the proposed change increase the consequences of a malfunction of equipment evaluated in the safety, analysis report?

The deflector plates are installed internal to the diesel engines and do not affect any equipment evaluated in the safety analysis report except the diesel generators themselves. With regard to the diesel generators, the safety analysis report already addresses the maximum consequences of a diesel engine failure.

Based on the above, the consequences of a malfunction of equipment important to safety are not increased.

5) Does the proposed change create the possibility of an accident of a different type than any evaluated previously in the safety analysis report?

This engineering Package installs a deflector plate in the Unit 2 2A1, 2B1 and 2B2 diesel generators. The modification is internal to the diesel generators and does not affect any equipment evaluated in the safety analysis report except the diesel generators themselves. The only accident of a different type that could be postulated under the circumstances, is a common mode failure of the diesel engines. To ensure that the possibility of a common mode failure does not exist, each failure mode of the deflector plate assemblies has been addressed. The deflector plate assemblies are seismically, qualified and constructed of a material that is compatible with the crackcase environment.

PCjM 080-290 Supplement 0 SAFETY EVALUATION (Continued)

Based on the above, this modification does not create the possibility for an accident of a different type than. any previously evaluated in the SAR.

6) Does the proposed change create the possibility of a malfunction of equipment of a different type than any evaluated previously'he St. Lucie Unit.2 LOOP analysis addresses the failure of one diesel generator. There is no possibility of a failure of a different type because the deflector plates are installed internal to the diesel engines. The deflector plates do not affect any other accident mitigation system.

Based on the above, the modificatio'n does not create the possibility of a malfunction of equipment of a different type than any previously evaluated in the SAR.

7) Does the proposed change reduce the maigin of safety as previously defined in the basis for any Technical Specification)

The Technical Specifications contained in Section 3/4.8, Electrical Power Systems, are not affected by this modification. Therefore, the margin of safety as defined in the basis for these Technical Specifications is not reduced.

The foregoing constitutes, in accordance with 1.0 CFR 50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question or require a change to the Technical Specifications. Prior NRC approval for implementation is not required.

PC/M 022-290 Supplement 0 ABSTRACT This engineering package (EP) provides for the removal of the two startup neutron sources from the St. Lucie Unit 2 core. Startup sources provide neutrons in oider to produce a positive indicatiori on the ex-core'neutron detectors that are used to monitor subcritical neutron multiplication during fuel reload operations. For future c'ycle fuel reload operations, subcritical neutron multiplication is to be monifored using irradiated fuel reloaded from previous cycles as the neutron sources.

The removal of the startup neutron sources from the St. Lucie Unit 2 core is required because the source assemblies are becoming damaged due to wear. The startup neutron sources reside in a vacant control, element assembly (CEA) guide tube of a fuel assembly nearest the ex-core start-up range neutron detectors. The upper end fittings of each source assembly are becoming damaged and it is, difficult to latch the source assemblies with the source lifting tool. As wear on the upper end fittings co'ntinues, it may become impossible.to latch the sources. Currently no vendor exists with the ability to manufacture startup neutron sources of the same design as those currently utilized at St. Lucie Unit 2.'t is therefore necessary to remove the startup neutron sou'rces and use irradiated fuel as the startup neutron source.

The startup neutron sources are to be stored in a permanent location within a spent fuel assembly (CEA guide tube) located in the spent fuel pool. (The location shall be identified by the St. Lucie Reactor Engineering Staff.)

~

During fuel reload operations the startup neutron sources are used in the monitoring of subcritical multiplication of the. loaded fuel assemblies which are safety related. The neutron

~

sources are not classified as Safety Related per FSAR section 4.2, however, the FSAR notes that the sources are manufactured in accordance with 10 CFR 50 Appendix B, therefore this EP has been classified as safety related.

The safety evaluation demonstrates that this EP does not constitute an unreviewed safety

'uestion, nor require a change to the technical specifications. Therefore, prior NRC approval is not required for implementation.

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

'This engineering package addresses the removal of the startup neutron sources from the St.

Lucia Unit 2 core. These sources are used to provide neutrons for positive indication on the ex-core neutron detectors used to monitor subcritical neutron multiplication during fuel reload operations.

Due to wear, the source assemblies are becoming damaged and need to be replaced or permanently removed from the core. Currently, no vendor is capable of manufacturing startup neutron sources of the same design as those currently being used. Therefore, the startup neutron sources need to be permanently removed from the St.,Lucie Unit 2 core.

PC/M 022-290 Supplement 0 SAFETY EVALUATION (Continued)

Reference 6.2 demonstrates that irradiated fuel with a sufficient accumulated exposure can be used as a source of neutrons for positive indication on the ex-core detectors used to monitor subcr'itical neutron multiplication during fuel reload operations. The concentration of isotopes that decay by spontaneous fission increases as a function of accumulated exposure.

If the accumulated exposure is long enough, the source strength from irradiated fuel will be greater than or equal to the design basis value of 5.0E7 neutrons per second calculated in Reference 6.2.

The design modification addressed by this engineering package is the permanent removal of the startup neutron sources from the St. Lucie Unit 2 core. Irradiated fuel reloaded from previous cycles is to be used as the neutron source for monitoring subcritical multiplication during all future fuel reload operations. The startup neutron sources are to be stored in the spent fuel pool.

During fuel reload operations the startup neutron sources are used in the monitoring of subcritical neutron multiplication of the loaded fuel assemblies, and fuel assemblies perform a safety related function. The neutron sources are not classified as Safety Related per FSAR section 4.2, however the FSAR notes that the sources are manufactured in accordance with 10 CFR 50 Appendix B, therefore this engineering package has been classified as safety related.

Title 10 of the Code of Federal Regulations Section 50.59 (Reference 6.3) states that the licensee may make changes to the facility as described in the FSAR without prior Nuclear Regulatory Commission approval unless the proposed change involves a change to the technical specifications or an unreviewed safety question. A proposed change involves an unreviewed safety question if:

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 may be increased, or ii) 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) the margin of safety as defined in the basis for any technical specification is reduced.

The modification proposed in this engineering package neither involves a change to the plant technical specifications nor an unreviewed safety question. Plant operation utilizing irradiated fuel as a neutron source is not a safety concern. The basis for this conclusion are addressed below.

i) The removal of the startup neutron sources and their replacement with irradiated fuel does not increase the probability of occurrence or the consequences of a malfunction or accident.

PC/M 022-290 Supplement 0 SAFETY EVALUATION (Continued)

.'he fuel assemblies that are to reside in the source locations for a given operating cycle, as determined by the reload analysis for that cycle, may not meet the criteria given in Reference 6.2 required for a fuel assembly to be used as a neutron source during refueling. These criteria define a minimum accumulated exposure which a fuel assembly mu'st receive in order to produce a source strength large enough for positive indication on the ex-core neutron detectors. Therefore, temporary placement of irradiated fuel assemblies into the source locations (2- nearest the ex-core wide range detectors) may be required to produce a count rate off the ex-core detectors that is sufficient for monitoring subcritical multiplication. Checking against the F.A.S.T Table (Fuel Assembly Storage Table) gives alternate fuel locations in which fuel assemblies can be temporarily placed during core alterations and maintain the required shutdown margin. When enough'fuel assemblies are-placed in the, core to produce a sufficient count rate off the ex-core neutron detectors, the temporary fuel assemblies may be removed from the source locations and replaced with the fuel assemblies that are to reside in the source location during the operating cycle. The additional transfer of at most two fuel assemblies does not increase the probability of an accident or a malfunction by a measurable quantity.

The removal of the startup sources is needed since the source is becoming damaged

, and it is becoming difficult to transfer the startup sources from one fuel assembly to another during fuel reload operations. Due to wear of the upper'end fitting, it is

'ifficult to latch the source with the source lifting tool. The probability of a malfunction of the source handling tool is reduced since the sources are to be permanently removed from the core and latching of the sources with the lifting tool will no longer be required.

The consequences of a malfunction of the fuel handling tool are bounded by the fuel handling accident of FSAR section 15.7.4.1.2.

No'new types of accidents 'or malfunctions exist as a result of removing the startup neutron sources and utilizing irradiated fuel as the startup neutron source. The operations required to perform this modification are the same as those of existing OP's, with the exception of not reinstalling the sources into the core.

The basis to the refueling instrumentation Technical Specification (TS 3/4.9.2) is that redundant monitoring capability, is available to detect changes in the reactivity condition of the core. The neutron count rate on the ex-core neutron detectors will not change significantly due to this modification and will represent a count rate greater than or equal to that of the design basis. source.

The implementation of this PC/M does not require a change to or impact the plant Technical Specifications.

As per the requirements of 10 CFR 50.59, this change does not involve an unreviewed safety question or a change to the Technical Specifications, therefore, prior NRC approval for the implementation of this PC/M is not required.

PC/M,009-290 Supplement 0 ABSTRACT This Engineering Package provides the necessary details to erect a Craft RCA Access Point just south of the Unit 2 RAB. An existing prefabricated metal building will be constructed on a new foundation and the RCA perimeter fence will be relocated accordingly.

\

The Craft RCA Access'Point does not perform any Nuclear Safety-Related functions: In addition it is not in close proximity to any safety related equipment, nor will it impact any safety-related functions. Accordingly, this Engineering Package has been classified as Non-Nuclear Safety-Related.

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 the 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 equipmerit 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:

The probability of occurrence and the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report are not increased, by this modification because it does not affect the availability,. redundancy, capacity, or function of any equipment required to mitigate the effects of an accident. Furthermore, the Craft RCA Access Point does not perform any function either directly or indirectly related to Power Plant operations. Therefore, there can be no adverse impact on Nuclear Safety.

The possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report will not be created by this modification because the modification involves non-nuclear safety-related structures and failure of any items. added by this modification will not impact any nuclear safety-related functions. In addition, any-conceivable mishap associated with the construction of the Craft RCA Access Point will not cause an accident or malfunction of any structure, system, or component important to Nuclear Safety.

iii) The margin of safety as defined in the bases for any technical specification is not affected by this modification since the components involved in this modification are not included in the bases of any Technical Specifications.

PC/M 009-290 Supplement 0 SAFETY EVALUATION (Continued)

The Craft RCA Access Point does not perform any safety-related functions. A failure mode evaluation has been performed for this modification and it has been determined that no new failure,modes have been introduced to the plant.

Accordingly, this engineering package has been classified as non-nuclear safety related.

The foregoing constitutes, per 10 CFR 50.59, 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 nor does it adversely affect plant safety and operation. Therefore, prior NRC approval for the implementation of this modification is not required.

PC/M 343-289 Supplement 0 ABSTRACT Under normal plant operating conditions, the Control Element Drive Mechanism Control System (CEDMCS) is powered via the Reactor Trip Switchgear (RTS) by the two (2) Control Element Assembly Motor Generator (MG) sets operating in parallel. An operational problem may arise when these two MG sets do not share the load equally (function of voltage adjustment) resulting in the more heavily loaded MG set tripping off line. The lightly loaded unit may not be able to accept the load under this scenario and may also trip with a resulting undervoltage trip on the RTS buses and subsequent reactor trip. This Engineering Package will provide means for fine voltage adjustment rheostat, which will allow adjustment for equal load sharing between the MG sets.

The new rheostat has an overall diameter of 1-9/16" and depth of'1-3/8" with a weight of 0.19 lbs.

FSAR Section =7.7.1.1.1 states that the CEDMCS is not required for safety since the removal of power by the RTS causes all Control Element Assemblies (CEAs) to be inserted by gravity.

Since the MG sets provide power to the CEDMCS, this EP has been classified as Non-Nuclear Safety Related.

The safety evaluation of this package has shown that the implementation of this PC/M does not constitute an unreviewed safety question and requires no revision to the St. Lucie Plant Unit 2 Technical Specifications. Prior NRC approval is not required for implementation. This e PC/M 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 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 have been evaluated under 10 CFR 50.59 and it has been determined that this EP does not involve an unreviewed safety question. The following are the bases for this conclusion.

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. This modification installs a fine voltage adjustment rheostat to.the voltage regulator of each MG set. This modification provides for the capability to adjust the loads between the two MG sets with greater precision, thus preventing operational problems (undervoltage trip on the Reactor Trip Switchgear) that initiate a reactor trip.

PC/M 343-289 Supplement 0 SAFETY EVALUATION The possibility for an accident or malfunction of a different type than any previously evaluated in the Safety Analysis Report is not created. Components involved with this modification do not introduce the potential for a new type of failure (i.e., the new rhe'ostat fails'esulting in an open circuit). The voltage" regulators (i.e., rheostats) and the MG sets are non-safety related and their failure do not cause malfunctions of safety related equipment.

The margin of safety as defined in the bases for any Technical Specification is not reduced since the installation of the rheostats do not affect any Technical Specification, nor do the CEDM MG sets. form the bases of any Technical Specification.

The implementation of this PC/M does not require a change to the plant Technical Specifications, nor'does it create anunreviewed safety question. Prior Nuclear Regulatory Commission approval for the implementation of this PC/M is .

not required.

0 PC/M 272-289 Supplement 0 ABSTRACT This Engineering Package is being performed to enhance'he ability of the Feedwater Pump

-recirculation valve to respond fast enough to compensate for rapid changes in the 15%

bypass valve position at low power levels. This modification utilizes a spare cable between each feedwater pump control switch and a new electrical splice box in the Turbine Building.

A new cable will be added from the new electrical box to the new solenoid valves located at each recirculation valve. This modification will enable plant operating personnel to ensure the complete opening of the recirculation valves when the feedwater pump control switches are placed in the "RECIRC" position for pump starting and low power operation.

The equipment involved in this modification is non-nuclear safety as described in FSAR sections 10.4.7 and 7.7.1.1.4. However, this package is classified as Quality Related due to wiring changes necessary within the RTGBs in close proximity to safety related wiring and components in the Control Room.

The safety evaluation of this package has shown that the implementation of this PC/M does not constitute an unreviewed safety question and prior NRC approval is not required for implementation. Implementation of this Engineering Package does not require any changes to the Plant Technical Specifications and will have no adverse effect on plant safety or operation.

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 defined in the bases for any technical specification is reduced.

This Engineering Package involves connecting each feedwater pump recirculation valve "full open" solenoid valve to its control switch. This will ensure the complete opening of the recirculation valves when the feedwater pump control switch is placed in the "RECIRC" position. This will involve adding cabling, conduit, two electrical splice boxes and two solenoid valves.

The equipment involved in this modification is non-nuclear safety as described in FSAR sections 10.4.6 and 7.7.1.3.1. However, this package is classified as Quality Related due to wiring changes necessary in the RTGB which will be performed in close proximity to Safety Related components and wiring. The sole purpose of the feedwater pump recirculation valve is to protect the feedwater pump from overheating during pump starting and low flow operation. The valve is designed to fail open and protect the feedwater pump.

PC/M 272-289 Supplement 0 SAFETY EVALUATION(Continued)

Presently at low power the recirculation valves modulate to maintain a minimum flow through

,each feedwater pump. As feed flow increases each recirculation valve closes, ensuring sufficient flow to protect the feed water pumps. This modification is being performed to enhance the ability of the recirculation valve to respond fast enough to compensate for rapid changes in the 15% bypass valve position at low power levels. This modification will enable the recirculation valve to be fully opened from the control room. The recirculation valve is designed to fail open to -protect the feedwater pump. Should the feedwater pump recirculation fail open at full power> this could result in a plant trip without prompt operator action. This modification will not change the effects of the failure modes or introduce any new failure modes of the Feedwater Pump Recirculation Valve.

The modifications hav'e been evaluated under 10 CFR 50.59 and it has been determined that this EP does not involve a'n unreviewed safety question. The following are the bases for this

. conclusion.

(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 increased because the feedwater pump recirculation system cannot fail in any manner which would inhibit the fulfillment of any safety related functions. The feedwater pump recirculation valves have no safety related function and are not required by any design bases for the mitigation of a design basis accident.

The possibility for an accident or malfunction of a different type than any previously evaluated in the Safety Analysis Report is not created because components involved with this modification introduce no new type of accidents and cannot cause malfunctions'of any safety related equipment.

(iii) The margin of safety as defined in the bases for any Technical Specification is not reduced since these minor wiring modifications do not affect any Technical Specification.

The implementation of this PC/M does not require a change to the plant Technical Specifications, nor does it create an unreviewed safety question.

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 and prior Nuclear Regulatory Commission approval for the implementation of this PC/M is not required.

j

PC/M 244-289 Supplement 0 ABSTRACT The implementation of this Engineering Package (EP) will provide the capability for the Steam

'enerator Blowdown Containment Isolation Valve Control Switches to override a Containment

, Isolation Signal (CIS) or High Radiation signal and allow reopening of each of the isolation valves (FCV-23-3, FCV-23-5, FCV-23-7, FCV-23-9). This modification will allow blowdown and sampling of the Steam Generators during certain design bases events (i.e. Steam Generator tube rupture).

The modifications in this EP will be made to the Containment Isolation System which is defined as Nuclear Safety Related by FSAR Section 6.2 and the Steam Generator Blowdown system which is classified as Non-Safety Related by FSAR Section 10.4. Due to the involvement with a Safety Related function (Containment Isolation), this EP has been classified as Nuclear Safety Related.,

The present control circuits will be modified such that the containment isolation/high radiation closure signals may be overridden through the addition of a CLOSE/OVERRIDE position on the valve control switches, in place of the existing CLOSE position. In order to utilize the override feature, with a Containment Isolation Signal (CIS) or High Radiation signal present, the control switch must first be placed to the CLOSE/OVERRIDE position prior to being placed to the OPEN position. The first action will arm the override feature and the second will energize the valve coil and open the valve. Once the closure signals have been oyerridden, a new isolation signal will cause the isolation valves to close.

Two spare annunciator windows will be used to provide control room indication of the override status. One annunciator window will be used for the blowdown isolation valve and sample isolation valve for each steam generator.. They will provide indication that a CIS or high radiation signal is present and that the subject valve(s) is open.

Additionally, a low pressure condition in the Steam Generator Blowdown will also cause the isolation valves (FCV-23-2 and FCV-23-5) to close; this feature is unaffected by this modification.

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 require a change.to the Technical Specifications and does not reduce the margin of safety for any Technical Specification. This PC/M 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 unresolved 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; or (ii) if the possibility of 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.

PC/M 244-289 Supplement 0 SAFETY EVALUATION (Continued)

This Engineering. package (EP) provides the engineeririg and design necessary to modify the steam generator blowdown isolation valve control circuits 'to (1) provide containment isolation/high radiation override capabilities in an off normal/post-accident scenario, (2) require two distinct and deliberate operator actions in order to accomplish the override feature'and, (3) provide positive indication, via auditory and visual annunciation on the RTGB-206, that (a) containment isolation" or high radiation signal is present and, (b) the valve solenoid is energized (i.e. the containment isolation valve is open with CIS or high radiation present). The control switch escutcheon will be re-engraved to reflect override capabilities. Two annunciator windows are used'to reflect the override status.

Two existing spare annunciator windows are being utilized to provide control room indication of the overiide status; each steam generator has one annunciator for the associated blowdown isolation valve and sample isolation valve. They will provide indication that a CIS or high radiation signal is present and that the subject valve(s) is open. The location of these annunciators will be RTGB-206 windows P-4, and P-14.

This modification enables blowdown following a tube rupture which will enhance operator control of Steam Generator Level by reducing actions required to use the blowdown system to drain the Steam Generator.

The modifications included in this'Engineering Package do not involve an unreviewed safety question because:

The probability of occurrence or, consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased since the use of the override capability will be utilized only in an off-normal/emergency operating condition and strict adherence to the governing procedure will require alignment of the blowdown system to receive process fluid with potentially high activity levels. The system's normal operation or response to a CIS is unaffected by this modification.

The possibility for an accident or malfunction of a different type than any previously evaluated in the Safety Analysis Report is r)ot created because components involved with this modification introduce no new type of accidents and cannot cause malfunctions of any safety related equipment. No changes have been made to the normal operational design of any control circuits or associated systems which are important to safety.

iii) This modification does not change the margin of safety as defined in the basis for any Technical Specification since the steam generator blowdown system is not part of any Technical Specification bases. The CIS portion of the circuit will still function in accordance with the Technical Specifications.

This package is classified as Nuclear Safety Related since the containment isolation functions of the blowdown valves are interlocked with the Engineered Safety Feature Actuation System (ESFAS) and are used to mitigate the consequences of an accident.

,. PC/M 244;289. Supplement 0

~

SAFETY EVALUATION(Continued)

Implementation of this Nuclear'Safety Related PC/M does not require a change to the Plant Technical Specificatioris.

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

-Plant Technical Specifications; thus, prior NRC approval for the implementation of this PC/M

. is not required.

PC/M 159-289 Supplement 0 ABSTRACT r

Presently, a high radiation signal, in containment results in a containment evacuation alarm being transmitted, first, in the containment. for a predetermined period, and then throughout the plant. This alarm will continue until the high radiation condition subsides. To eliminate this unnecessary continuous high noise source in the control room, the containment evacuation alarm will be modified so that the alarm will continue in the containment but will silence in the general plant including the control room after a thirty (30) second delay time. This 'modification meets the original intent of the design specifications for the plant communications system, which called for the containment evacuation alarm to sound throughout the, plant and then drop out everywhere except in containment after a time delay (Reference 6.12).

This modification also provides the capability for the operator to activate

~

the fire or site evacuation alarm and override the containment evacuation alarm. This modification does not compromise the personnel safety of the general plant population.

The containment evacuation alarm is a part of the non-nuclear safety related plant communications system. For this reason, this Engineering Package (EP) will be classified as Non-Nuclear Safety Related.

The results of the safety evaluation conclude that modifications presented by this Engineering Package do not constitute an unreviewed safety question, do not require any changes to the Plant Technical Specifications and therefore, no prior NRC approval for the implementation of. this PC/M is required.

The implementation of this PC/M will not have an adverse impact on plant safety or operations.

This modification does not involve seismic design, nor does it introduce potential interactions with nuclear safety related components. 4 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.

1105E

PC/M 159-289 Supplement 0

~N SAFETY EVALUATION (Continued)

This Engineering Package (EP) addresses the rewiring of the non-nuclear safety related containment evacuation alarm. The present system provides'ndesirable noise levels in the Control Room. With this modification, a high radiation signal initiates an,, alarm in containment, the Control Room, and to the

'population for a preset time of'"30 seconds. After the 30 seconds has

'general'lant elapsed, the alarm will, sound only in containment. until the high radiation signal is removed. This modification also gives the operator the capability to activate the fire or site'vacuation alarm and override the containment

, evacuation alarm. , This capability - allows for interruption to warn plant personnel 'of new alarms that'an not be sounded with existing design. Once these alarms are, disengaged, the containment evacuation alarm will resume the high radiation signal still exists. This modification does not compromise if the personnel safety of'he general plant population. The containment evacuation alarm is a part of the non-nuclear safety related communications system. This modification does not involve seismic design, nor does introduce potential interactions with n'uclear safety related components.

it For this reason, this EP will be classified as Non-Nuclear Safety Related.

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

The probability of occurrence or the consequence of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased. ,Wiring modifications will be for part of the non-safety related communication system. No safety related equipment, previously evaluated in the safety analysis report is involved with this modification.

As a result of this modification, there is no possibility for an accident or malfunction of a different type than any previously evaluated. The modifications- by -the EP involve the rewiring of the containment evacuation alarm to alter the audible sequencing and provide the capability to activate the fire and site evacuation alarms during the sounding of the containment evacuation alarm.

(iii) The wiring modifications are for the containment evacuation alarm which performs no safety 'related function. The margin of safety, provided by the Technical Specifications is preserved.

The implementation 'f technical specifications.

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

1105 E'

PC/M 103-289 Supplement 1 ABSTRACT This Engineering Package (EP) provides for the installation of a Diverse Scram

'ystem (DSS) in St Lucie Unit 2 to meet the requirements of 10CFR50.62, th'e "ATWS Rule". The DSS will add Reactor Coolant System over-pressure protection circuitry

- designed to function during Anticipated Transient Without Scram (ATWS) events and will, utilize both new and existing equipment. It will include new electronic circuit boards in the Engineered Safety Features Actuation System (ESFAS) cabinets, and new Non-Nuclear Safety Related load contactors installed in the Control Element Assembly (CEA) Motor Generator Sets Control Cabinets at the output of 'the (CEA) drive motor-generator sets. Whenever the motor-generator sets are operating and the ATWS/DSS modules are, energized, the DSS .will be functi:onal and will not require operator attention. The DSS has been designed to, be independent and diverse from the Reactor Protection .System.

Although the regulations in 10CFR50.62 do not requ'ire the DSS to be Nuclear Safety Related, the modifications in the ESFAS cabinets are classified as Nuclear Safety Related based on the existing safety classification of components in those circuits. The .load contactors, however, are Non-Nuclear Safety Related to be consistent with the Non-Nuclear Safety classification of the CEA drive MG sets.

Results of the Safety Evaluation indicate that this EP does not involve any unreviewed safety question nor do these modifications involve a change or addition to the Plant Technical Specifications. Therefore the DSS addition can be implemented without prior USNRC approval. There are no adverse conse'quences to plant safety or operation due to the implementation of this PCM..

Supplement No 1 The purpose of this supplement is to remove all holdpoints "associated with this EP. Environmental and seismic qualification documentation for the ESFAS cabinet modification has been received from Eaton Consolidated .Controls and is acceptable.

The implementation of this supplement will have no impact on plant safety or operation.

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 cr'eated, or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

On July 26, 1984, The Code of Federal Regulations was amended to include Section 10CFR50.62, "Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants" (also known as the ATWS Rule). The ATWS Rule requires specific improvements in the design ,and operation 'of commercial nuclear power facilities to reduce the likelihood'f a failure to shut down the reactor following anticipated transients, and to mitigate the consequences of anticipated transients which occur without a shutdown. The occurrence of an anticipated transient in conjunction with a failure of the Reactor Protection System (RPS) to produce a reactor trip is defined as an ATWS event..

1038I

PC/M 103-289 Supplement 1 SAFETY EVALUATION (Continued)

The combination of an RPS failure and an .anticipated transient is outside the present plant design basis and was analyzed by Combustion Engineering (CE) via CENPD-158. It was determined that a complete loss of feedwater combined with a failure of the reactor to trip 'ould result in a primary coolant system pressure excursion well above reactor vessel service level C limi.ts and therefore potentially challenge the integrity of the reactor coolant pr'essure boundary.

For Combustion Engineering plants the regulations require the implementation of two, methodologies for ensuring that an excessive primary coolant pressure excursion does not occur. These methodologies are called "prevention" and "mitigation".

Prevention takes form as a Div'erse Scram System (DSS) whose purpose is to initiate a shutdown of the reactor by control rod insertion upon conditions indicative of an anticipated transient, independently and diversely from the, RPS. ;,Miti'gation is accomplished by tripping the turbine and initiating, Auxiliary Feedwater to generator inventory and to ensure that a primary coolant heat sink is conserve'team available. A combination of prevention and mitigation will limit the peak reactor coolant sys'em pressure rise to within acceptable values.

The Diverse Scram System (DSS) utilizes existing Nuclear Safety-Related pressurizer pressure instruments and signal converters in the instrument cabinents and takes as inputs, signals from secondary current loops in RTGB-206. These signals are wired to the Engineered Safety Features Actuation System (ESFAS) cabinets where they are processed by DSS'istable. and logic components to provide reactor trip signals.

The trip signals are used to open the= non-safety related control element assembly drive (CEA Drive) motor generator (MG) set output load contactors between the MG set output circuit breakers and the Reactor Trip Switchgear. The consequential loss of voltage on the Reactor Trip 3witchgear. buses causes the reactor to shut down. This system, diverse and independent from the RPS except at the instrument loops, satisfies the ASS Rule requirements for ATWS prevention.

The modifications included in this Engineering Package do not involve an unreviewed safety question because: ~ ~

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased. This is confirmed by the following:

/

To ensure that the existing Reactor Protection System (RPS) is fully capable of functioning as designed to protect against accidents and malfunctions of equipment important to safety that have been evaluated and documented in the FSAR and form the'esign basis for the licensing commitments, the NRC has issued specific guidance for electrical independence for the RPS and the Diverse Scram System (DSS).

1038I

PC/M 103-289 Supplement 1 SAFETY EVALUATION (Continued)

Electrical independence is required from sensor output to the final actuation device at which point non-safety circuits must be isolated from safety related circuits. To achieve this, the, DSS is designed so that inputs to the DSS from the sensors will be electrically isolated from the Reactor Protection System by isolation devices in the instrument cabinets. The DSS logic circuits will be contained in the ESFAS cabinets which are electrically independent from the Reactor Protection System. The DSS outputs will be directed to the CEA drive MG set load contactors, which are not safety-related nor associated with the Reactor Protection System trip functions. DSS outputs will be isolated to prevent adverse electrical interactions between the non-safety actuation devices and the safety related portion of the DSS installed in the ESFAS cabinet.

Physical separation from the RPS is not required, but the separation criteria applied to the existing protection system must not be violated. Since the DSS is not maintained in the same cabinets as the RPS, separation criteria are not affected.

Eaton Consolidated Controls has designed the DSS such that it will not adversely interact with ESFAS, with which it shares the same cabinets. It interfaces with the ESFAS in only three areas: 1) Power supplies The DSS is separately fused so that a DSS fault cannot propagate through the power supplies to any ESFAS functional loops. 2) ATI The DSS is optically isolated from the ATI to ensure independence from other ESFAS funct'ional loops. -3) Input signal (pressurizer pressure) The DSS is designed so that single failure at the bistable input will not affect the ESFAS functions that use the same input. Since the DSS is designed to meet or exceed the requirements of the existing ESFAS design, the DSS will not cause adverse interactions through the interfaces with ESFAS.

No change has been made to the input parameters or the ability of the RPS and ESFAS to perform safe shutdown functions based on these parameters. Therefore, the probability of occurrence or the'onsequences of an accident or malfunction of equipment important to safety previously evaluated is not increased.

(ii) There is no possibility for an accident or malfunction of a different type than any previously evaluated. This .is confirmed by the following:

Design 'of . the DSS does not introduce new components that could violate the RCS pressure boundary, release radioactive material to the environment, or damage the reactor fuel. It does not change the method or ability of the RPS or ESFAS to detect system parameters and perform required safety functions based on the values of those parameters. Malfunction of the DSS cannot cause an event other than a reactor trip. The new DSS will reduce the risk of an over-pressure condition due to an ATWS event and the subsequent stressing of the RCS in excess of ASME Level C pressure of 3200 psia (Reference 6.29, Section 1.3). Therefore the DSS does not increase the possibility for an accident or malfunction of a different type, than any previously evaluated, but actually reduces it.

1038I

PC/M 103-289 Supplement 1 SAFETY EVALUATION (Continued)

(iii) This modification does not reduce the margin of safety as defined in the basis for any technical spegification. This is confirmed by the-following:

This modification does not adversely affect equipment whose operation is defined by the Plant Technical Specifications and does not affect the operation of the RPS in any manner. Therefore, the margin .of safety as defined in the .basis'or any-Technical Specification is not reduced.

The conceptual designs for the Diverse Scram, System and Diverse Turbine Trip have been issued by, FPL documents JPE-M-87-078 and JPN-PSL-SEIJ-89-034, respectively, to the NRC. Rev'iew has been completed on these two documents and it has been concluded by the NRC that the Diverse Scram System and Diverse Turbine Trip designs are acceptable. The NRC documented ' their approval ;with a Safety Evaluation dated September 6, 1989. single concern was noted in the Safety Evaluation . about the use of a single annunciator window to indicate. two different messages (Actuated and Bypassed). The concern is corrected in this EP by providing two windows, one for each message.

Although NRC guidance does,not require the DSS to be classified as safety-related, portions of the DSS and this EP have been class'ified as Nuclear Safety Related due to the DSS integration into safety-relate'd circuits and installation in safety-related ESFAS cabinets. The DSS does not perform a safety-related function. It provides protection against an over-pressure'ondition in the RCS, in which ASME Level C criteria would be exceeded.

Because this EP does not impact in an adverse manner any cables essential to safe reactor shutdown or systems associated with achieving and maintaining safe shutdown conditions, this package has no impact on 10CFR50 Appendix "R"- fire ,protection requirements.

Therefore, the proposed design of this package is in compliance with the applicable codes and St Lucie Unit 2 FSAR requirements for fire protection equipment.

Implementation of this EP does not require a change to the Plant Technical Specifications.

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 nor a change to the Plant Technical Specifications. NRC approval has been obtained for the conceptual design with the Safety Evaluation dated September 6, 1989. Prior NRC approval for the implementation of this PCM is.not required.

1038I

PC/M 035'-289 Supplement 1 ABSTRACT This Engineering Package (EP) provides for the elimination of two 4 inch diameter by 1'2 inch long fiberglass spool pieces located on the Intake Cooling Water (ICW) piping which act as tap connections for the sensing lines for FIS-21-9B.and any necessary rework of associated piping, up to the instrument root valves. A future'upplement will rework the piping from the root valve to the instrument. Until that time, the instrument piping should be considered nonseismically qualified. An evaluation has shown that complete severance of the instrument line does not affect the ability of the. ICW system to perform its safety function and ICW flow can be confirmed by other means.

These changes are required as dispositions to NCR's 2-120 5 2-121, as one fiberglass spool piece is cracked and the other is needed elsewhere in the plant.

Insulating gaskets and bolt sleeves will be used to avoid galvanic interaction between the

'carbon steel ICW piping and aluminum-bronze instrument sense lines.

The modifications considered in the EP are in the ICW System. The ICW System performs a safety related function, therefore, this modification is classified as safety 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.'he implementation of this EP does not require a chan'ge to the plant'Technical Specifications and does not reduce the margin of safety for any Technical Specification.

The implementation of this EP will have no impact on plant safety or operation.

Supplement 1 reroutes piping located adjacent to the instrument root valves as shown on the backfit change sketches. The piping reroute*is required to disposition NCR 2-121.

Although the safety analysis has been revised, the original results of evaluation, as stated above, remain unchanged. The re'routed piping has been seismically qualified and'the supports/restraints for the affected piping have been qualified for the stress analysis loads.

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.

PC/M 035-289 Supplement 1 I

SAFETY EVALUATION (Continued)

The modifications included in this EP will eliminate two 4 inch diameter by 12 inch long fiberglass spool pieces branching off the Intake Cooling Water (ICW) piping which act as a tap,.

connections for the sensing lines for FIS-21-9B, and rerouting of associated piping with consequent relocation of supports/restraints. The new and modified supports/restraints have been demonstrated to be adequate for revised stress analysis loads in accordance with applicable codes and criteria. Insulating gaskets and bolt sleeves will be used to avoid" galvanic interaction between the carbon steel ICW piping and aluminum-bronze instrument sense line.

In the interim period between commencement and completion of this modification, the instrument. root valves shall be normally closed. 'In the unlikely event that the pressure boundary, between the ICW header and the instrument root valves fails, the heat removal

.capabilities of the ICW/CCW Systems will not be adversely affected because:,

1. The break location is downstream of the CCW heat exchanger therefore, total flow through the heat exchanger is not affected.

2. The ICW System is an open system and therefore system fluid inventory is not a concern.

3. Flooding of the CCW Building is not a concern because all equipment is located above elevation 23.66 feet on pedestals (Ref FSAR Section 34.1). The maximum possible flood elevation inside the CCW Building is 23.5 feet. (Ref.

Drawing 2998-G-077, Sh 2 of 3) which is the elevation that the water would spill out of the outside air intakes.

The ICW System performs a Safety Related function, namely transfer of decay heat from the Component Cooling Water system to the ultimate heat sink, (Atlantic Ocean), therefore the modifications included in this Engineering Package (EP) are considered to be safety related.

This EP does not involve an unreyiewed safety question; the following are the bases for this justification:

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 increased since the'changes made by this modification are functionally equivalent and meet all the original design requirements. The rerouted piping and revised supports have'een designed to the same requirements as the original design. Therefore no new failure modes have been created and the probability of occurrence or consequences of a failure has not changed.

PC/M 035-289 Supplement 1 SAFETY'VALUATION(Continued) I The probability for an, accident or malfunction of a different type than any evaluated previously in the safety analysis report has not been created as a result of this modification since the function and design of the ICW system has not been changed. As identified in the failuie modes and effects analysis performed for this modification (Attachment 7.3),-no new failure modes have been created.

iii) This modification does not reduce the margin of safety as defined in the basis for any Technical Specification because if neither, changes the design parameters of the ICW system nor does it change the ICW design flow or functional re q uirements.

The implementation of this PC/M does not require 'a change to or impact the plant 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 sa'fety question or a change to the Technical Specifications, therefore, prior NRC approval for the implementation of this PC/M is not required.

PC/M 349-989 Supplement 1 ABSTRACT This Engineering Package provides the necessary details for the installation of an above ground Waste Oil Tank for the St. Lucie site. The facility will consist of the foundation and spill retainer for a single 8,000 gallon'ank as well as a pump to drain 55 gallon drums into the tank.

The Waste Oil Tank does not perform any Nuclear Safety-Related functions. It will be-

'constructed at the south end of the site just west of the east retention basin and is not in the vicinity of any safety-related equipment or systems, nor will it impact any safety-related functions. Accordingly, this Engineering Package has been classified as Non-Nuclear Safety-Related.

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 the 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.

Su lement No. 1 This supplement is being issued to allow use of expansion anchors in the field routing of electrical equipment. This supplement does not affect, amend, or change the original safety evaluation or the 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 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 tech'nical specification is reduced. The modifications included in this engineering package do not involve an unreviewed safety, question because of the following reasons:

PC/M 349-989 Supplement 1 SAFETY EVALUATION (Continued)

The probability of occurrence and the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report are not increased by this modification because it does not affect the availability,.redundancy,. capacity, or function of any equipment required to mitigate the effects of an accident. An explosion and fire analysis was performed for the Fuel Dispensing Facility Modification (PC/M 399-988).

It demonstrated that there would be no impact to nuclear safety for an explosion of 18,000 gallons of gasoline at a distance of greater than 240 feet .

from any safety-related equipment. Since the facility is greater than 700 feet from any safety-related equipment and the minimum 'allowed distance is proportional to the cube root of the amount of explosive (Ref. 6.16), it is

'oncluded that the storage of 8,000 gallons of waste oil (in addition to the 18,000 gallons of fuel) at the proposed location will have.no adverse effect on nuclear safety. Furthermore, the Waste Oil Storage Tank does not perform any function either directly or indirectly related to Power Plant operations.

Therefore, there can be no adverse impact on Nuclear Safety.

The possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report will not be created by this modification because the modification involves non-nuclear safety-related structures and failure of any items added by this modification will not impact any nuclear safety-related functions. In addition, any mishap at the Waste Oil Storage Tank including fire, explosion, and construction activities will not cause an accident or malfunction of any structure, system, or component important to Nuclear Safety.

, iii) The margin of safety as defined in the bases for any technical specification is not affected by this modification since the components involved in this

. modification are not included in the bases of any Technical Specifications.

The Waste Oil Storage Tank does not perform any safety-related functions. A failure mode evaluation has been performed for this modification and it has been determined that no new failure modes have. been introduced to the plant. An explosion analysis and fire analysis was performed previously for the Fuel Dispensing Facility in accordance with the St. Lucie Unit 1 FSAR, the St. Lucie Unit 2 FSAR, and 10 CFR 50 Appendix R, and it was concluded that there will be no adverse effect on the plant. The analysis was reviewed and it has been determined that the conclusions of the analysis are still valid for the additional 8,000 gallons of flammable liquid stored at this location. Accordingly, this engineering package has been classified as non-nuclear safety related.

The foregoing constitutes, per 10 CFR 50.59, 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 nor does it adversely affect plant safety and operation.

Therefore, prior NRC approval for the implementation of this modification is not required.

PC/N 073-989 Supplement 0 ABSTRACT The requires replacement. 't existing meteorological tower, located north of St Lucie .Units 1 and 2, will be replaced with one of similar construction, a Rohn-80 model designed and erected by Ellis Tower, Inc. Nost "of, the data collecting equipment, processors and transmitters for the tower and for the meteorological station will be replaced with improved equipment. This Engineering Package provides the details for the installation of the new tower including: the location and installation details of the new .tower, details for the relocation or modification of the fence which encircles the tower, and a plan for the removal of the existing tower once the new station is in operation. The removal of the existing tower and the erection of the new tower will be performed by Ellis Tower, Inc. The fencing modifications and the tower erection included in this PC/N will be implemented prior to the implementation of PC/N 109-987 whose scope of work includes the installation of replacement instrumentation and associated electrical work. The removal of the existing tower included in this PC/N will be implemented after the implementation of PC/N 109-987 has been completed.

The components added by this Engineering Package have no nuclear safety related function, and do not have the potential to interact with any nuclear safety related equipment, components or systems. However, the instruments supported by this tower are required to be operable by the Plant Technical Specifications. Therefore, this Engineering Package is classified as Quality Related to obtain Quality Control Verification that instructions are followed during installation of this tower and removal of the existing tower.

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 conse- quences 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 mal'function 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 Engineering Package (EP) provides the requirements for'he installation of a new meteorological tower, its foundation and anchorages, north of the St Lucie Plant.= This tower and its associated equipment perform no safety related function and cannot 'interact with any safety related equipment or system. However, since the tower supports instrumentation required to be operable by the Plant Technical Specifi'cations, this EP has been classified as Quality Related, requiring Quality Control inspection to assure that there will be no interaction with operable equipment during the installation or removal of the towers. This modification does not give rise to an unreviewed safety question. The following are the bases for this conclusion'.

0350c

PC/M 073-989 Supplement 0 SAFETY EVALUATION (Continued)

The probability of occurrence or the consequences of an accident or malfunction of equipment important to 'safety previously evaluated :in the 'afety Analysis Report is not increased by ,these modifications. The Meteorological Tower Replacement Specification requires that the tower be designed such that all of the applicable design criteria are satisfied. Furthermore, failure of the tower or any system associated with 'it cannot affect any safety related equipment or system. Therefore, the implementation of these modifications cannot increase the probability of an accident and cannot increase the con'sequences of any design basis event.

(ii) The possibility'f an accident or malfunction of 'equipment of a different type than any evaluated previously is not created. There is no, possible interaction with any safety related equipment or system. Therefore, a failure of. any safety relate'd component which could cause, contribute to, or become a factor in a new type of accident cannot result from this modification.

(iii) The margin of safety as defined in the bases for any Technical Specification is not reduc'ed by this modification. The meteorological instrumentation is discussed in 'Sections 3.3.3.4 and 4.3.3.4 of the St Lucie Units 1 6 2 Technical Specifications. If failure of-the tower results in the number of operable meteorological channels being less than specified in the Technical Specifications Table 3.3.8, all releases of gaseous radioactive material from the radwaste decay tanks would be required to be suspended until operability is restored. Inoperability for more than 7 days would require special reporting to the NRC. However, the modifications included in this EP do not affect the instrumentation. This instrumentation will be replaced via PCM 109-987. The new tower has been specified to be designed for all applicable loads and installed, and the existing tower removed, such that the existing and replacement instrumentation will not be adversely affected.

The implementation of this PCM 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 or a change in the Technical Specifications and prior NRC approval for the implementation of this PCM is not required.

0350c

-PC/M 355-288 Supplement 0 ABSTRACT This Engineering Package (EP) provides for the replacement of four obsolete Fischer 5 Porter electronic integrators located on RTGB-205 with Rochester integrators and IVO modular counters. The integrators being replaced are FQI-2210X for reactor make-up water, FQI-2210Y for boric acid make-up, FQI-6627 for liquid waste discharge and FQI-6648 for waste gas discharge totalized flow.

The St. Lucie Plant - Unit 2 FSAR Section 9.3.4 discusses the Chemical and Volume Control System (CVCS), Section 11.2 discusses the Liquid Waste System and Section 11.3 discusses the Gaseous Waste System but no specific details have been provided about these integrators.

These integrators perform no Nuclear Safety Related function. This modification does not involve Nuclear Safety Related equipment or function. However, since it involves modifications to the RTGB which contains Safety Related equipment, this EP has been classified 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 This EP provides for the replacement of four obsolete Fischer & Porter electronic integrators located'on RTGB-205 with currently manufactured devices. The existing integrators which have an integral counter on the face and are mounted on the front face of the RTGB-205 will be replaced with panel mounted counters (IVO industries) installed in the same location as the existing integrators and separate integrators (Rochester) mounted within the RTGB-205. This modification meets the original design intent for the measurement of reactor make-up water, boric acid make-up, liquid waste discharge and waste gas discharge totalized flow. The new equipment associated with this modification has been seismically qualified for structural integrity and will be seismically mounted to preclude potential interaction with safety related components of RTGB-205.

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. The modifications included in this EP do not involve an unreviewed safety question because:

0 PC/M 355-288 Supplement 0 SAFETY EVALUATION(continued)

The probability of occurrence or the consequences of an accident or

'malfunction of equipment important to safety previously evaluated is not increased by this modification because it does not affect any equipment required to mitigate the effects of an accident or which can initiate an accident.

Further, the equipment installed is seismically qualified and mounted to preclude any potential interaction with safety related components within the RTGB.

There is no possibility for an accident or malfunction of a different type than any previously evaluated since this modification involves only the replacement of existing components with suitable substitutes and no changes have been made to the operational design of the CVCS and the Waste Management

~

Systems of which they are a part.

This modification does not change the margin of safety as defined in the basis for any Technical Specifications since the integrators being replaced do not form the bases of any Technical Specifications.

RTGB-205 is on the essential equipment list. However, this modification does not affect any Appendix R functions, cables and/or equipment on RTGB-205. Therefore, this EP has no effect on safe shutdown, alternate shutdown,'or 10 CFR 50 appendix R requirements.

Implementation of this Quality Related PC/M does not require a change to the Plant Technical Specifications.

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 nor a change to the Plant Technical Specifications; thus, prior NRC approval for the implementation of this PC/M is not required.

PC/M 230-288 Supplement 0'BSTRACT Engineering Package (EP) provides resolution to support/restraint (S/R)

I'his drawing discrepancies identified in NRC Inspection Report Nos. 50-335/87-26, 50-389/87-25, 50-335/88-28 and 50-389/88-28. This EP additionally involves a modification to a S/R on piping which is Safety Class 3, Quality Group C and Seismic Category I.

Discrepancies identified in the above referenced NRC report and others identified in the course of evaluating the anchor bolt discrepancies have been resolved in this EP by providing a modification drawing for one S/R, and incorporating as-found information into affected drawings and calculations.

This EP does not add or modify any piping nor does it involve any new or revised piping stress analyses.

The implementation of this EP does not require a change to the Plant Technical Specifications. - The safety evaluation demonstrates that this EP does not constitute an unreviewed safety question. Therefore prior NRC approval is not required for the implementation of this EP. The implementation of this EP will have no adverse impact on Plant safety or operation.

SAFETY EUALUATION 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 basis for any Technical Specification is reduced.

The modification included in this EP is to modify S/R Mark No. CC-2061-191 to restore the +Y restraint function to comply with the existing stress analysis of record. The. S/R being modified is located on piping for the Component Cooling Water system, which is Nuclear Safety Related.

There were no modifications performed 'n S/R Mark Nos. BF-4001-190, CS 2012 8015'S 38 Rlly CS (I) 44 R3y CS 2002 1 67'O 17 R4y FS 2138 1'7 and SI-4205-6440B, or the associated piping systems.

The modified and as-found discrepant conditions for these S/Rs were evaluated and the revised calculations have demonstrated the adequacy of these S/Rs for the applicable piping loads in accordance with the existing plant criteria.

Therefore, there is no adverse impact on the ability of these systems to perform their design function and maintain their flow requirements.

0409c

PC/M 230-288 Supplement 0 SAFETY EVALUATION (Continued) t Since the modification included in this Engineering Package involves an S/R for a Nuclear Safety Related piping system, the package is classified as Nuclear Safety Related.. This EP does not involve an unreviewed safety question and the following are the bases for this conclusion:

(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 increased because:

There is no modification performed to the piping. S/R Mark No CC-2061-191 is to be modified to bring it into compliance with the stress analysis of record. All restraint functions and locations of the other S/Rs existing in the associated stress analyses of record remain unchanged.

l

2. The adequacy of the S/R to be modified has been demonstrated for the piping loads, utilizing the existing plant criteria in the St Lucie 2 FSAR, Section 3.9.3.4. and Appendix 3.9B.

3. The actual installed conditions for S/R Mark Nos. BF-4001-190, CS-2012-8015, CS-38-R11, CS-(I)-44-R3,, CS-2002-1-67, D0-17-R4, FS-2138-17 and SI-4205-6440B, which were identified as deviating from the existing design drawings, have been evaluated and determined to be adequate in accordance, with the applicable existing plant design criteria in the St Lucie 2 FSAR, Section 3.9.3.4. and Appendix 3.9B.

As a result of this modification, there is no possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report because:

No piping modifications were performed and no stress analysis calculations were revised.

2. The as-found S/Rs and the modified S/R have been demonstrated to be adequate for the piping loads in accordance . with the applicable codes and criteria.

(iii) The modification and discrepant as-found conditions do not reduce the margin of safety as defined in the bases for any Technical Specification because they neither change the design parameter of the associated piping systems nor do they change the design flow or functional requirements of the systems. The modification does not affect the integrity of the pressure boundary of the subject piping.

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

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 or a change to the Technical Specifications, and prior NRC approval for the implementation of this PCM is not required.

0409c

PC/M 227-288 Supplement 0 ABSTRACT This Engineering Package provides for the installation of Noise Filters in the Analog Display System (ADS) CRT video enables. The ADS CRT has experienced noise bars drifting through the CRT display and periodic loss of synchronism.

The noise bars and loss of .synchronism .are caused by AC current on the video cables'hields due to voltage differences in the grounds of the ADS cabinet and the ADS CRT. The noise filters are ground loop transformers which isolate the ADS cabinet ground from the ADS CRT ground, thus .eliminating the AC currents on the cable shields. Testing performed by St Lucie Plant Instrument and Control Maintenance has confirmed that installation of the noise filters will eliminate the ADS CRT noise bar. and synchronism problems. The noise filters will be installed in the ADS cabinet.

The Analog Display System is classified as Non-Nuclear Safety Related in FSAR sections 7.5 and 7.7. However, this EP is classified as Quality Related since the Analog Display System cabinet is required to maintain its physical integrity during a seismic event and this modification will be performed in the Control Room. The Plant Technical Specifications and the requirement of 10CFR50.59 have been reviewed. This modification goes not require technical specification changes nor d'oes it involve an unreviewed safety question.

Therefore prior NRC approval is not required. This modification has no affect on plant operation or safety.

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 occurence 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 defined in the bases for any technical specification is reduced.

the margin of safety as The modification described in this PC/M installs noise filters in the video cables which connect the Analog Display System output to the ADS CRT display

,on RTGB-204. The ADS monitors the vertical position of the 91 Control Element Assemblies (CEA's), which are graphically displayed on the ADS CRT. A CEA back-up display is also available for the operators. use. FSAR Section 7.5.1 defines the ADS as being Non-Nuclear Safety Related. The noise filters will be mounted in the ADS cabinet. Equipment in the cabinet is not required to function after a seismic event.

The ADS cabinet is seismically qualii ied such that it cannot interact with safety related equipment during a seismic event. The noise filters are not.

seismically mounted within the'DS cabinet. However, 'the ADS cabinet protects all safety related equipment from potential mounting failures of the noise filters during a seismic event by containing the noise filters.

0627I/0066I

PC/M 227-288 Supplement 0 SAFETY EVALUATION (Continued)

These modifications have been evaluated under 10CFR50.59 and 'it has "been determined that this -EP does not involve an unreviewed safety question. The following are the bases for this conclusion:

The probability of occurrence or the consequences of an accident os malfunction of equipment important to safety previously evaluated in the Safety Analysis Report is not increased because the noise filters perform no safety related function and cannot fail in any manner which would inhibit the operation of any safety related equipment.-

(ii)'he -possibility for an accident or malfunction of a different type than any previously evaluated in the Safety Analysis Report is, not created because the noise filters introduce no new type of accidents and cannot cause malfunctions of any safety related equipment.

(iii) This modification does not reduce the margin of safety as defined in the bases for any Technical Specification because it does not affect any of the bases in the Technical Specifications; he implementation of this PC/M does not require a change to the plant technical peci.fi.cati.ons.

he 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 approval for the implementation of this PCM is not required.

0627I/0066I

PC/M 199-288 Supplement 0 ABSTRACT This Engineering Package (EP) covers the modifications required to remove the Model MT250 signal generator/amplifier and electromechanical transducers used to verify the operation of the Loose Parts Monitoring System (LPM) and to calibrate the system. The MT250 signal generator/amplifier is used to drive the electromechanical transducers which provide vibration (acceleration) to the accelerometers of the LPM system through direct contact. However, experience shows that this equipment does not provide traceable or repeatable calibration.

Furthermore, this equipment produces excessive electrical noise in other systems located in the Control Room and is considered a high maintenance item.

Currently to verify operability and/or perform calibration of the LPM system a channel comparison check is performed by the operators at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. A channel

. functional test is performed at least monthly using an oscilloscope. A channel calibration test is performed at least once every 18 months using a'ball and pendulum.

Although the MT250 signal generator/amplifier and electromechanical transducers, along with the balance of the LPM system are classified Non-Nuclear Safety Related, the safety classification of this PC/M is Quality Related since it involves modifications in the Control Room and on the surface of the reactor and steam generators. The safety evaluation has determined that this PC/M does not constitute an unreviewed safety question and does not require a change in the Plant Technical Specifications. This PC/M can be implemented without prior NRC approval ~

This PC/M has no impact on plant safety or operation.

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 EP removes LPM calibration equipment from the LPM cabinet and from the surface of the Steam Generators and Reactor Vessel.

This equipment has been shown to be inaccurate. Currently, other methods are used to calibrate the LPM system. Currently, a channel comparison check is performed by the operators at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. A channel functional test is performed at least monthly using an oscilloscope. A channel calibration test is performed at least once every 18 months using a ball and pendulum. Thus, the Loose Parts Detection Program requirements of NRC Regulatory guide 1.133 are met by the LPM system.

The Loose Parts Monitoring System is not safety related. This package does not involve any Nuclear Safety Related equipment or functions.

PC/M 199-288 Supplement 0 SAFETY EVALUATION (Continued)

This package is classified as Quality Related because it involves modifications in the Control Room and on the surface of the reactor vessel and steam generator. This EP has no affect on the design basis as provided in St Lucie.Unit 2 FSAR Section 4.4.6.

The modifications included in this Engineering Package do not involve an unreviewed safety question because:

The probability of occurrence or, the consequences of an accident or malfunction of equipment important to safety previously evaluated is riot increased since the vibration amplifier and transducers serve no function in the control of plant operations or safe shutdown.

ii) There is no possibility for an accident or malfunction of a different type than any previously evaluated since no changes have been made to the operational design of any control circuits or associated systems which are important to safety.

This modification does n'ot change the margin of safety as defined in the basis for any technical specification since the Vibration Amplifier and transducers serve no function in the control of plant operations or safe shutdown.

Furthermore, their removal does not affect the operability of the LPM system nor does it affect the ability to verify LPM system operability.'ue to the fact that the EP does not involve any cables essential to safe reactor shutdown or systems associated with achieving and maintaining safe shutdown conditions, this package has no impact on 10 CFR 50 Appendix "R" fire protection requirements. Therefore, the proposed design of this package is in compliance with the applicable codes and St Lucie Unit 2 FSAR requirements for fire protection equipment.

Implementation of this'uality Related PC/M does not require a change to the Plant Technical Specifications.

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

PC/M 085-288 Supplement 1 ABSTRACT This Engineering Package (EP) provides for the replacement of three (3) obsolete Westinghouse pressure transmitters PT-22-27, PT-22-28 and PT-22-42 used in the Turbine Digital Electro-Hydraulic (DEH) Control System and the installation of associated instrument power supplies and input signal resistors. These transmitters measure turbine impulse pressure, turbine intermediate stage pressure, and turbine throttle pressure, and input these parameters to the DEH Control System.

r These transmitters perform no Nuclear Safety Related function. Since some of the modifications associated with this EP will be performed in the DEH Control cabinet located in the Control Room, this EP has been classified as Quality Related to preclude potential interaction with safety related components in the vicinity. The dynamic characteristics of the DEH cabinet and consequently its responses during a seismic event are not degraded by this modification.

The safety evaluation of this EP has determined that this PC/M does not constitute an unreviewed safety question and does not require a change in Plant Technical Specifications.

This PC/M can be implemented without prior NRC approval. This PC/M has no adverse impact on plant safety or operation.

Su lement 1 Supplement 1 is issued to install fuses.and fuse blocks for each of the power supplies installed as part of this PC/M. This modification does not involve an unreviewed safety question or a change to any plant Technical Specification. It has no effect on plant safety or operation and does not affect, amend, or change the conclusion of the original 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 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 reduc'ed.

This Engineering Package (EP) provides the engineering and design necessary for the replacement of three (3) obsolete Westinghouse (Hagan) pressure transmitters PT-22-27, PT-22-28 and PT-22-42 used in the Turbine Digital Electro-Hydraulic (DEH) Control System and the installation of associated instrument power supplies, fuses and fuse blocks, and input signal resistors. These transmitters measure turbine impulse pressure, turbine intermediate stage pressure, and turbine throttle pressure, and input these parameters to the DEH Control System.

PC/M 085-288 Supplement 1 SAFETY EVALUATION (Continued)

The existing transmitters are powered by 120 V ac power. Each transmitter outputs a 1 to 5 V dc signal which is the signal range required for DEH Control System input signals. The replacement transmitters, Rosemount model 4'1152 GP, are driven by 24 V dc instrument power and generate a dc current output of 4 to 20 mA. Therefore, three new instrument power supplies (one for each transmitter) will be installed in the DEH Control cabinet to power the new transmitters. Three fuses and three fuse blocks (one for each power supply) will be installed in the DEH Control cabinet. Three (3) ranging resistors (one for each transmitter) will be installed at the DEH Control cabinet input terminal strips to obtain the necessary DEH Control System input signals (i.e., converting the 4 to 20 mA dcsignal into 1 to 5 V dc).

Since some of the modifications associated with this EP will be performed in the DEH Control cabinet located in the Control Room, this package has been classified as Quality Related to preclude potential interaction with safety related'components in the vicinity. The dynamic characteristics of the DEH cabinet and consequently its responses during a seismic event are not degraded by this modification.

The modifications included in this Engineering'Package (EP) do not involve an unreviewed safety question because:

i) The probability of occurrence or the consequences of an" accident or malfunction of equipment important to safety previously evaluated'is not increased by this modification because it does not affect. or change the availability, redundancy,-capacity, or function of any equipment required to mitigate the effects of an accident. Further, the transmitters involved in this EP are not required by any design basis for the mitigation of an accident.

ii) There is no possibility for an accident or malfunction of a different type than any previously evaluated since no changes have been made to the operational design of any control circuits or associated systems which perform or monitor a safety related function. Further, the tr'ansmitters involved in this EP do not perform a safety related function.'ii),

This modification does not change the'margin of-safety as deflated in the basis for any Technical Specification since the new. pressure ti'ansmitters perform a non-nuclear safety related function. Further, the pressure transmitters being

. replaced do not form the bases of any Technical Specifications.

Due to the fact that this EP does not involve any cables essential to safe reactor shutdown or systems associated with achieving and maintaining safe shutdown conditions, this package has no impact on 10 CFR 50 Appendix "R" fire protection requirements. Therefore, the proposed design of this package is in compliance with the applicable codes and St. Lucia Unit 2 FSAR requirements for fire protection 'equipment.

Implementation of this Quality Related PC/M does not require a change to the Plant Technical Specifications.

PC/M 085-288 Supplement 1 SAFETY EVALUATION (Continued)

The foregoing cons'titutes, per 10 CFR 50.59(b), the written safety evaluation which provides the basis that this change does not involve an unreviewed safety question nor a change to the Plant Technical Specifications; thus, prior NRC approval for the implementation of this PC/M-is not required.

PC/M 399-988 Supplement 1 ABSTRACT This Engineering Package provides the necessary details for the installation of a Fuel Dispensing Facility for the St. Lucie site. The facility will consist of the foundation and spill retainer for two tanks (one 8,000 gallon and one 10,000 gallon) as well as the fuel dispensers and necessary appur'tenances.

The Fuel Dispensing Facility does not perform any Nuclear Safety-Related functions. It will be constructed at the south end of the site just west of the east basin and is not in the vicinity of any safety-related equipment or systems, nor will it impact any safety-related

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functions. Accordingly, this Engineering Package has been classified as Non-Nuclear Safety-Related.

A safety evaluation of this modification has been performed in.,accordance with 1.0 CFR 50.59. This evaluation indicates that implementation of this Engineering Package does not involve an unreviewed safety question. Furthermore, the implementation of this modification does not require a change to the Plant Technical Specifications and has no detrimental effect on plant safety,and operation. Therefore, prior NRC approval for implementation of this mo'dification is not required.

h This supplement includes the design'details for adding, an electric sump pump to empty rainwater from the fuel dispensing facility spill retainer. In addition, the original design drawings will be updated to reflect as-built conditions and the PC/M Expiration Date has been extended to December 31; 1990. These changes require the original engineering design bases and design analysis to be amended, but have no effect on the original safety evaluation or the 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 a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report may be create'd, 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:

PC/M 399-988 Supplement 1 SAFETY EVALUATION (Continued)

The probability of occurrence and the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Updated Safety Analysis R'eport are not increased by this modification because it does not affect-the availability, redundancy, capacity, or function of any equipment required to mitigate the effects of an accident. It has been shown that there will be no adverse impact on structures, systems, or components greater than 240 feet from the Fuel Dispensing Facility should an explosion occur there. Since the tanks will be located at least 700 feet from the nearest safety-related structure, system, or component and does not perform any function either directly or indirectly related to Power Plant operations, there will be no adverse impact on Nuclear Safety.

. ii) The possibility of an accident or malfunction of a different type than any evaluated previously in the Final Updated Safety Analysis Report will not be created by this modification because the modification involves rion-nuclear

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safety-related structures arid failure of any items added by this modification will not impact any nuclear safety-related functions. In addition, any mishap at the Fuel Dispensing Facility including fire, explosion, and construction activities will not cause an accident or malfunction of any structure, system, or component important to Nuclear Safety.

iii) The margin of safety as defined in the bases for any technical specification is not affected by this modification since the components involved in this modification are not included in the bases of any Technical Specifications.

The Fuel Dispensing Facility does not perform any safety-related functions. A failure mode evaluation has been performed for this modification and it has been determined that no new failure modes have been introduced to the plant.

An explosion analysis and fire analysis have been performed in accordance with the St. Lucie Unit 1 FSAR, the St. Lucie Unit 2 FSAR, and 10 CFR 50 Appendix R, and it has been concluded that there will be no adverse effect on the plant.

as the design of the facility meets or exceeds the requirements of these documents. Accordingly, this engineering package has been classified as non-nuclear safety-related.

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 008-988 Supplement 3 ABSTRACT This Engineering Package covers modifications to the St. Lucie Plant Simulator Training Facility, including the addition of the Visitors Facility and expansion of the Training Department. The Visitors Facility will be located on the first floor of the building, while an addition to the second floor on the north end of the building will be constructed to accommodate relocation and expansion of the Training Department. The Simulator Training Facility is not located within the perimeter of the plant security fence and does not affect any plant safety-related system. Therefore, this PC/M is classified as Non-Nuclear Safety Related.

Results of the safety evaluation conclude that modifications presented by this Engineering Package do not constitute an unreviewed safety question, do not require any changes to the Plant Technical Specifications, and do not require prior NRC approval for the implementation of this PC/M.

The implementation of this PC/M will not have adverse impact on plant safety or operations.

Imn No.1 This engineering package revision incorporates the following:

~ package expiration data extension.

~ minor administrative changes, including addressing the potential need to revise Security Procedure No. 0006123.

~ drawings revised or added to accommodate the final design of the Visitors Center.

This supplement does not affect, amend, or change the conclusions reached in the original safety evaluation. This supplement will not adversely impact plant safety or operations and does not require a change to any Plant Technical Specification.

Su lemen No. 2 This engineering package revision incorporates drawings revised and added to provide for the enclosure of Room 121 (Covered Shop Area) and for minor electrical modifications to the first floor Visitors Center area.

This supplement does not affect, amend, or change the conclusions reached in the original safety evaluation. This supplement will not adversely impact plant safety or operations and does not require a change to any Plant Technical specification.

Su lemen No. 3 This Engineering Package revision incorporates drawings revised to accommodate the new design of the Visitors Center entrance canopy. This supplement does not affect, amend or change the conclusions- reached in the original safety evaluation. This supplement will not adversely impact plant safety or operations and does not require a change to any Plant Technical Specification.

PC/M 008-988 Supplement 3 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 urireviewed 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)'ifthe margin of safety as defined in the bases for any Technical specification is reduced.

The Simulator Training Facility serves to satisfy NRC'requirements for training. The purpose of the facility is to train the nuclear plant operators in a simulated control room environment,

,and to provide training for other operations and maintenance activities. Expansion of the Training Department on the second floor of the facility will provide much needed space for additional classrooms, study areas, laboratories, and offices.'The Visitors Facility will enable the general public to visit the training facility, gain an understanding of nuclear power generation, and see FPL personnel undergoing training. Enclosure of the covered shop area (Room 121) will prevent equipment stored in this area from experiencing corrosion degradation due to exposure to the salt air. The facility is located on the north side of the site, outside of the security perimeter fence, and is not in the vicinity of any plant safety-related structure or system. The facility does not in any way perform or affect a plant safety-related function.

A failure mode evaluation has been performed for this modification and it has been determined that failure of the modification will not impact any nuclear safety-related functions or equipment. Accordingly, this engineering package is classified as Non-Nuclear Safety Related.

I The utilities for the modification connect to existing building systems, and have been determined not to adversely impact either the existing building systems or the existing plant systems. A supplement to PC/M 013-187 will provide for the Gal-Tronics communication and evacuation alarm system for the Visitors Facility and second floor addition. Administrative controls for evacuation of the facility, as defined by Security Procedure No. 0006123,.are already in effect for the building and will automatically be expanded to include the Visitors Facility and new second floor area. These administrative controls will remain in effect as a compensatory measure until such time as PC/M 013-187 is fully implemented.

The addition of the Visitors Facility and Training Department Expansion does not change any assumptions made or conclusions drawn in the St. Lucie FSAR, and does not adversely affect any site flooding analysis or erosion control plans. The addition of the Visitors Facility and Training Department Expansion does not pose an unreviewed safety question as defined by 10 CFR 50.59. The following are the bases for this conclusion:

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 has not been increased.

Because it has been demonstrated that the Visitors Facility and Training

'Department Expansion has no impact on nuclear safety, all previous safety evaluations contained in the safety analysis report remain valid.

PC/M 008-988 Supplement 3 SAFETY EVALUATION (Continued)

. ii) ~ The possibility of an accident or malfunction different than those previously evaluated in the safety analysis report has not been created.

Because the'Simulator Training Facility, housing the -Visitors Facility and expanded Training Department, is not located in the vicinity of safety-related components or systems, no postulated failures of the building or its functions can affect safety-related activities. Furthermore, no postulated construction accidents can impact nuclear plant safety. Accordingly, no new possible accidents or malfunctions different than those previously evaluated in the safety analysis report have been created.

iii) The margin of safety as defined in the basis for any Technical Specification has not been. reduced.

Because the Visitors Facility and Training Department Expansion has no effect at all on the basis for any technical specification, the margin of safety as defined in the basis for any Technical Specification remains unaffected.

10 CFR 50.59 allows changes to a facility as described in the FSAR if an unreviewed safety question does not exist and if a change to the Technical Specification is not required. As shown in 'the preceding sections, the change proposed by this design package does not involve an unreviewed safety question because each concern posed by 10 CFR 50.59 that pertains to an unreviewed safety question can be positively answered.

'n conclusion, 'I the change proposed by this design package is acceptable from the standpoint of nuclear safety, does not involve an unreviewed safety question, and does not requiie a change to the Plant Technical Specifications. Therefore, this package may be implemented without prior NRC approval under 10 CFR 50.59.

PC/M 095-287 Supplement 0 ABSTRACT The installation of new couplings on the diesel generator fan drive shafts, accomplished under PC/M 019-287, necessitated the removal of the original sheet metal shaft guards to accommodate the larger replacement couplings. Ba'sed on a concern for personnel safety, the installation of new shaft guards is required. This Engineering Package provides the details for the installation of the new shaft guards.

. The components added by this Engineering Package have no safety function. However, since there exists,a potential for the interaction of these modifications with the safety related diesel generators, they are seismically designed and, consequently, this Engineering Package is classified as Quality Related.

As the diesel generator corresponding to the shaft guard being installed will be disabled to minimize the risk to personnel and to safety related components, this modification will be implemented on only one diesel generator at a time. It is recommended to be implemented only during Plant Mode 5 or 6.

These modifications do not involve an unreviewed safety question, have no effect on plant safety or operation, and do not require a change to the plant Technical Specifications.

Therefore, prior NRC approval is not required for the implementation of this Engineering Package.

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 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 Engineering Package (EP) provides the requirements for the installation of sheet metal guards around exposed portions of the diesel generator fan drive shafts for the purpose of

.enhancing personnel protection and preventing dropped tools or other items from striking the shafts. Although the shaft guards perform no safety related function, they and their supports have been seismically designed in order to preclude the possibility of interaction with the diesel generators or other safety related equipment or systems in their vicinity; accordingly, this EP has been classified as Quality Related. This modification does not give rise to an unreviewed safety question. The following are the bases'for this conclusion:

PC/M 095-287 Supplement 0 SAFETY EVALUATION 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 by this modification. The diesel generator fan drive shaft guards and their supports have been designed such that all of the design criteria have been satisfied; therefore, they cannot fail in a 'pplicable

, way which could increase the probability of an accident. The diesels are relied upon to mitigate the consequences of an accident; the shaft guards'will protect the diesels against damage and thus enhance their reliability. Therefore, the implementation of these modifications cannot increase the consequences of any design basis event.

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The possibility of an accident or malfunction of equipment of a different type than any evaluated previously is not created. The effect of the modification on the diesel generators has been evaluated and found not to impact their seismic qualification. There is no impact on or interaction with any other safety related equipment or system. Therefore, a failure of any safety related component which could cause, contribute to, or become a factor in a new type of accident cannot result from this modification.

iii) The margin of safety as defined in the bases for any Technical Specification is not reduced by this modification. For personnel safety considerations, the diesel generator corresponding to the shaft guard being installed will be disabled while the installation is in progress, causing the action statement of Technical Specification 3.8.1.1 to be entered, if the modification is implemented during plant operation. Restoration of the diesel generator operability following the completion of the installation will resolve this condition.

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

r The foregoing constitutes, per 10 CFR 50.59(b), the.w'ritten safety evaluation which provides the bases that this change does not involve an unreviewed safety question or a change in the Technical Specifications and prior NRC approval for the implementation of this PC/M is not required.

PC/M 089-287 Supplement 1 ABSTRACT This Engineering Package (EP) is for the modification of the Remote Reactor Vessel Level Indicator.'his modification will provide more reliable level indication during refueling and reduce personnel radiation exposure since it replaces a temporary system which required more

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attention for operation.

The modifications considered in this EP are on the Reactor Coolant System. The connections are designated as nuclear safety related and seismically qualified because they are within the Reactor Coolant Pressure Boundary and therefore this modification is classified as safety related. The instrument side of the system downstream of the piping isolation. valve is non-safety, seismic design. 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 will have no impact on plant safety or operation.

Su Iement .1 The purpose of this supplement is to implement FPL's response to NRC Generic letter 88-17

. by performing the following:

a) Provide trending by interfacing the refueling level indication loops LI-1117/LI-1117-1 with the Emergency Response Data Acquisition and Display System (ERDADS)/Safety Assessment System (SAS).

b) Reduce the wide range level indicator from 0-720'-'o 0-400" water column.

c) Adding an alarm module (LA-1117) and a selector switch to the Plant Auxiliary Control Board 2 (PACB-1). The alarm annunciates on the PACB-2 Annunciator LC-9 and will alarm, on low level to indicate that there is a possible loss of shutdown cooling. A spare annunciator window will be utilized. The selector switch will allow instrument operation during refueling operation and will not allow nuisance alarms during normal plant operations.

The implementation of Supplement No. 1 will have no impact on plant safety or operation; The original safety evaluation has not been affected.

SAFETY EVALUATION,

/

With respect to Title 10 of the Code of Federal Regulation, 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.

PC/M 089-287 Supplement 1 SAFETY EVALUATION (Continued)

The modifications included in this Engineering Package are for the Reactor Vessel water level indicator installation involving piping, tubing, valves and orifices and differential. pressure transmitters, all connected between the RCS and the Pressurizer.

Su lemen 1 The modifications included in this supplement are for implementing FPL's response to NRC Generic letter 88-17 by adding an audible alarm that will alarm on low level to indicate that there is a possible loss of shutdown cooling and providing trending information of refueling level.

The implementation of Supplement No 1 will have no impact on plant safety. or operation.

Based on the above description, the modification included in this Engineering Package (EP) is considered to be safety related. This.EP does not involve an unreviewed safety question, and the following are bases for this justification:

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 since this modification provides a means whereby an accurate reactor vessel water level can be readily determined during refueling. During power operation this system is isolated from the RCS.

the portions of this modification within the normal RCS pressure boundary have been designed to the original requirements of the RCS pressure boundary.

ii) As a result of this modification, there is no possibility for an accident or malfunction of a different type than any previously evaluated because the modification provides double isolation valving which will isolate the system from the RCS during power operation.

iii) This modification does not reduce the margin of safety as defined in the basis for any Technical Specification because it neither changes the design parameter of the RCS nor does it change the RCS design flow or functional requirements.

The implementation of this PC/M does not require a change to the plant Technical Specification.

The foregoing constitutes, per 10 CFR Part 50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question.

PC/M 043-287 Supplement 0 ABSTRACT Historically, the Intake Cooling Water (ICW) Lube Water System piping has required significantly higher maintenance efforts to ensure lube water system integrity than projected during the system design phase. Various studies covering strainer problems, material selection,.flow considerations and fluid chemistry have not yielded a generic solution that could cost effectively increase system reliability to an acceptable level. One alternative for resolving these problems is to eliminate the need for the external lube water system.

Accordingly, this EP covers those modifications required to convert one ICW pump to a self-Iubricated configuration and terminate the respective external lube water supply to the pump.

To adequately assess long-term reliability of the self-lubricated design, the selected pump will

'perate for some time as the only self-lubricated pump. The decision to convert the remaining two pumps to this design will be contingent on the results of the long-term maintenance requirements of the affected pump.

This modification affects ICW pump 2A and it's associated lube water piping which are part

,. of the ICW System. Since the ICW System is classified as Nuclear Safety Related, Quality Group C, this modification is considered Nuclear Safety Related. Based on the failure modes, evaluation and 10 CFR 50.59 review, it has been demonstrated that failure of one ICW pump does not involve an unreviewed safety question, does not affect plant safety, nor does it adversely affect the operation of any other safety related equipment or functions." In addition, this modification does not require changes to the Technical Specifications. Implementation of this modification does not require prior NRC approval.

SAFETY EVALUATION The Unit 2 ICW System is classified as Nuclear Safety Related. As such, these modifications are considered to be Nuclear Safety Related, Quality Group C and have been designed to the appropriate ASME Code and Seismic Category I criteria as specified by the FSAR.

Based on the failure modes analysis provided in the Design Analysis, failure of a single ICW pump is a previously evaluated event which does not adversely affect plant safety since two separate redundant trains exist and only one train is required for safe shutdown. Also, failure of the coating system will not impact operability of the pump or ICW System.

Title 10 of the Code of Federal Regulations, Part 50.59 allows changes without prior Commission approval provided the proposed change does not involve an unreviewed safety question or require changes to the Technical Specifications. This proposed change does not involve and unreviewed safety question because:

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

The previous accidents evaluated in the FSAR include loss of one ICW pump and/or loss of one vital bus. Both accidents result in one fully functional ICW train. The probability of occurrence or consequences of these malfunctions is not increased by this modification.

PC/M 043-287 Supplement 0 SAFETY EVALUATION (Continued)

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

The modifications proposed herein do not create the possibility. for an accident or malfunction of a different type than any previously evaluated in the FSAR.

Failure of the modified components will not impact any other safety related equipment or functions.

'he k

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iii) 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 ICW system as addressed in the Technical Specifications.. No changes to the Technical Specifications are required and no new technical specifications are required by this modification.

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

In conclusion, this modification is acceptable from the standpoint of nuclear safety since it does not involve an unreviewed safety question as defined in 10 CFR 50;59 and does not require changes to the Technical Specifications. Implementation of this modification does not require prior NRC approval.

PC/M 109-987 Supplement 0 ABSIRACT This Engineering Package (EP) provides for the replacement of the meteorological equipment located in the St Lucie Units 1 and 2 Control Rooms and at the discharge canal. It will be implemented in con'junction with the replacement of .the meteorological tower and instrumentation at the tower station and will provide the plant with state-of-the-art equipment for data processing, display, recording, and transfer and will assist FPL to meet its commitments per Regulatory Guide 1.23, "Onsite Meteorological Programs" and 10CFR100.10, 10CFR100.11 and 10CFR50, Appendix I. In the Unit 1 Control Room, all meteorological display equipment will be replaced with a multipoint recorder which incorporates a digital readout. The Safety Assessment System (SAS) will be used for data display.

Obsolete equipment in the Control Room and at the discharge canal will be removed.

The meteorological system is not Nuclear Safety Related. However, the St Lucie Unit 1 and 2 Plant Technical Specifications (Section 3.3.3.4, both Units) require that the meteorological system be operable at'll times. In addition, the Unit 1 Control Room meteorological monitoring equipment and RTG boards in both Units must be designed to seismic requirements. Therefore, this Engineering Package is classified as guality Related. The safety evaluation has determined that this EP does not involve any unreviewed s'afety

.question or a change to the Plant Technical Specifications. Therefore, this EP can be implemented without prior Commission approval.

Implementation of this PCM will not affect the safety or operation of the plant.

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 t'o 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:

The .probability of occurrence or the consequences of an accident

'or malfunction of equipment important to safety previously evaluated is not increased by this modification because it does not affect or change the availability, redundancy, capacity, or function of any equipment required to prevent or to mitigate the effects of an accident or. of any equipment important to safety.

0715I

PC/M 109-987 Supplement 0 SAFETY EVALUATION (Continued)

(ii) There is no possibility for an accident or malfunction of a different type than any .evalu'ated previously in the Safety Analysis Report since no new failure modes are introduced which would have an effect on existing system capability of equipment important to safety. The existing meteorological parameters being monitored are maintained.

(iii)

N The margin of safety as defined in the. bases for any Technical Specification is not reduced since this modification does not

,reduce the capability of . the, system ,to obtain and store meteorological data necessary to meet the recommendations of Regulatory Guide 1.23, Cnsite Meteorological Programs, February 1972.

This EP does not affect equipment that is identified as Nuclear Safety Related. However, due to its function of providing data to evaluate radiological releases, due to the existing operation limitations in the Technical Specifications (3.3.3.4 for both'nits), and due to seismic and new cable considerations in the .Unit 1, Control Room, this EP is classified as Quality Related.

The implementation of this EP 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 doe's not involve an unreviewed safety question. Prior NRC approval for the implementation of this PCM is not required.

0715I

PC/M 073-987 Supplement 1 ABSTRACT This Engineering Package (EP) covers the replacement of the now obsolete Fischer and Porter transmitters with the currently manufactured equivalent Rosemount transmitters. The transmitters, are providing tank level and process flow monitoring signals in the Makeup Water System and in the Steam Generator Blowdown Treatment Facility.

The existing Fischer and Porter transmitters do not provide any interface with the safety related systems, therefore, this EP is classified Non-Nuclear Safety Related. Since this modification is a one-for-one replacement of the existing Fischer and Porter transmitters with the equivalent Rosemount transmitters, the same classification applies.

The safety evaluation of this EP does not involve an unreviewed safety question, and does not require a change in the Plant Technical Specifications. This EP does not impact plant safety and operation, therefore, NRC approval for these modifications, prior to their implementation, is not required.

Supplement 1 of this EP has been issued to reflect current Rosemount model numbers for transmitters LT-36-1 and LT-36-3. The model number for these devices (as purchased) is 1151LT4EBOB22D and is included in this package with this revision.

This supplement does not effect the safety evaluation; the implementation of this PCM does not affect the Plant Technical Specifications and does not constitute an unreviewed safety question.

This EP has no impact on plant safety or operation.

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 tothesafety consequences of an accident or previously evaluated in the malfunction of equipment important 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:

The probability of occurrence and the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Updated Safety Analysis Report, Chapters 8 & 9, are not increased by this modification because it does not affect the availability, redundancy, capacity, or function of any equipment required to mitigate the effects of an accident.

0338I/0029I

PC/M 073-987 Supplement 1 SAFETY EVALUATION (Continued)

The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Updated Safety Analysis Report, Chapters .8 & 9, will not be created by this modification because the function, of the transmitters has not been altered.

(iii) The margin of safety as defined in the bases for any technical specification is not reduced since the new transmitters are all classified non-nuclear safety related and do not affect any technical specification.

The existing Fischer 6 Porter transmitters do not provide any interface with the safety related systems, therefore, this EP is classified Non-Nuclear Safety Related.

The implementation of this EP does not require a change to the Plant Technical Specifications, nor does it create an unreviewed safety question.

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 Commission approval for the implementation of this PCN is not required.

0338I/0029I

PC/M 045-986 Supplement 5 ABSTRACT This Plant Change/Modification includes the installation of a perimeter security barrier (intake canal crossing), intrusion detection system (perimeter and underwater), surveillance. system (closed circuit television),

security lighting and communications (paging).

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This PCM is not classified as Safety Related since the canal crossing structures and components of the intrusion detection, surveillance, lighting and communication systems do not perform any safety function, and are located away from, and have no effect on, any safety related components. However, this PCM shall be considered quality related and quality related design requirements shall apply because of the following reasons'a) the perimeter security barrier closes the existing gap in the security perimeter fence as requ'ired by 10CFR73.55, b) the nature of the bridge and adjacent walls construction requires QC surveillance to assure proper installation of the concrete piles and correct use 'of concrete materials and mixes, and c) QC inspection/testing of the security system installation is required to assure proper operation and integration with the existing security system.

This PCM does not constitute an unreviewed safety question and enhances the existing plant security system. The installation of the items described above have no impact on plant operation and do not affect any safety related equipment.

SUPPLEMENT 4 This supplement incorporates the details required to install the second sonar system; monostatic microwave and their associated cables and conduits at the Intake Canal south crossing. The original safety evaluation i.s not affected by the modifications detailed in this supplement.

SUPPLEMENT 5 This supplement incorporates the details required to install a new intrusion detection system (perimeter and underwater) and to partially remove the existing system (perimeter and underwater) at the Intake Canal south crossing., The original safety evaluation is not affected by the modifications detailed in this supplement.

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 report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

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PC/M 045-986 Supplement 5 SAFETY EVALUATION (Continued)

This modification is not classified as Safety Related since the underwater intrusion system (canal crossing and associated systems, such's additional lights, paging -stations, fencing and security system hardware) doe's not perform any safety -function, and is located away from, and has no e'ffect on any safety related components. The Ultimate Heat Sink analysis described in FSAR Section 9.2.7 is not affected by this modification since the failure of this crossing during a seismic event will not impede the flow of water, nor limit the intake canal and intake structure bay area from providing the plant with the primary source of shutdown cooling water capacity to dissipate

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reactor decay heat during normal and emergency shutdown conditions.'his modification is on the intake canal area only, therefore the" secondary source of cooling water (Big Mud Creek) is not affected.

The modifi'cations included in this PCM do not involve any unreviewed safety questions bec'ause:

i The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased since the underwater intrusion detection system shall be installed in accordance with the quality related requirements, and this modification will have no effect on equipment performing a safety function.

ii There is no possibility for an accident or malfunction of a different type than any previously evaluated since the underwater

-intrusion detection system has no safety function, no changes have been made ,to any operational design, and the addition of security system hardware (TV cameras for surveillance, microwave 300B, fence protection FPS II, and a barrier net (Safenet) system) enhances. the plant security system by implementing permanent security in the area of the intake .canal crossing and integrating these modifications into the ,existing system. A

'anal crossing fa'ilure during a .seismic event will not provide blockage of the primary water source to the .intake cooling water system.

This canal crossing is not seismically designed and is a multi-span type bridge. Should a seismic event occur the canal crossing 'spans may individually collapse since there are construction joints separating each span. Based on April 1986 soil samples, the canal bottom contains firm soil at approximately elevation -26 ft. A collapse of this structure during a seismic event could hypothetically tip the 16 ft wide bridge walkway and drive it on edge into the canal, thereby leaving the top edge of the walkway no .higher than elevation -10 ft at the center of the canal.

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PC/M 045-986 Supplement 5 SAFETY EVALUATION (Continued)

This hypothetical scenario would still provide constant flow of water to the intake cooling water system since the lowest water elevation in the canal (wit/ two units operating) is elevation -9 ft and the cooling water would still continue to flow over the collapsed canal crossing. In addition to water flowing over the collapsed sections, water can flow in between each collapsed section of the bridge, thereby not blocking the primary water source to the intake cooling water system.

Since the intake structure is provided with a method to prevent floating debris from entering the intake cooling water system, any possible floating debris from the canal crossing would not clog the intake cooling water system. In addition, based on engineering judgement the components associated with this modification would sink to the bottom if a seismic event were to hypoth'etically collapse this structure, therefore no floating debris could float downst:ream to clog the intake cooling water system.

iii This modification does not change the margin of safety as defined in the basis for any technical specification.

The implementation of this PCM does not require a change to the 'plant technical specification.

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, therefore prior Commission approval is not required for implementation of this PCM.

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PC/M 018-285 Supplement 2 ABSTRACT The Leslie Y-strainers that are presently in the ICW pump lube water system have become unserviceable due to excessive system maintenance and must be replaced. New Leslie Y-strainers are unavailable. This PC/M documents the replacement of Leslie Y-strainers with Hayward Y-strainers. This change does not adversely affect plant safety or operation, and does not involve an unreviewed safety question nor a change to the Technical Specifications.

The ICW pump lube water Y-strainers are located within a Quality Group C/Class 3 portion of the piping system. Therefore, this PC/M is classified as Safety Related.

Revision 2 provides for replacement of the 100 mesh, 316 stainless steel screen elements in the ICW lube water Y-strainers with 0.020" perforated, monel screen elements. This change is required to reduce excessive backwashing of the 100 mesh strainers. The safety Evaluation is revised to address the impact of the 0.020" perforated strainer element on the operation of the ICW pumps. This change will not adversely affect operation of the ICW pumps. This change does not adversely affect plant safety or operation, and does not involve an unreviewed safety question nor a change to the Technical Specifications. Revision 2 also updates the format of this PC/M to the current procedure.

SAFETY EVALUATION The new strainers are being replaced in accordance with ASME Section XI and meet all applicable requirements of Section III ND and ANSI B16.34.

IWA-7400 states that piping valves and fittings 1" nominal pipe size and less are exempt from meeting the requirements of Article IWA-7000 with the exceptions that materials and primary stresses shall be consistent with applicable construction code (ANSI B16;34).

The primary function of the lube water system, and thus the strainer, is to provide a continuous supply of water to the ICW pump bearings. The function of the strainer body is to maintain the pressure boundary. This function is assured by meeting the applicable standards for manufacture and installation, which are consistent with the original design. The function of the screen element is to retaid degradation of the. bearings by removing particulate. Replacement of the 100 mesh screen with 0.020" perforated screen may result in minor acceleration in bearing degradation, but adequate testing is performed to identify degradation prior to bearing failure. Therefore, the change in screen mesh size will not adversely affect operation of the ICW pumps.

The proposed modifications do not involve an unreviewed safety question because:

a) The probability of occurrence or the consequences of a design basis accident or malfunction of the ICW system is not increased as the function of the strainers has not changed and the new strainers meet all the requirements of ASME Section XI. The consequences of an accident are not increased since the failure mechanisms of the new strainers are similar to the existing strainers and the potential for failure is considered equal to or less than the existing strainers.

PC/M 018-285 Supplement 2 SAFETY EVALUATION {Continued) b) The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR is not created as no new modes of failure have been created.

c) The margin of safety as defined in the basis for a technical specification is not reduced as the new Y-strainers are equal to or superior to the existing Y-strainers.

PC/M 046-285 Supplement 2

. ABSTRACT This Engineering Package (EP) covers replacement of relief valves on the St. Lucie Unit 2 Diesel Generator Air Start System air receivers. The modification is necessary because the present valve cannot meet FPL QA requirements for spare parts. The modification is classified as nuclear safety related since the Diesel Generator Air Start System is nuclear safety related.

Although the modification alters equipment used to mitigate accidents, it does not constitute an unreviewed safety question since the failure modes analysis and 10CFR50.59 reviews have identified any accidents or malfunctions which have not been addressed by the FSAR.

'ot Additionally, the modification is designed to original DG Air Start System design requirements, and the margin of safety as defined in the basis for any technical specification has not been reduced. Therefore, NRC approval is riot required prior to implementation of this modification.,

Su lemen 1 This supplement incorporated revised drawings for the relief valves procured from Dresser.

The changes to the drawings consisted of correction of vendor tag numbers and changing the material of the bases of the valves from ASME SA 479 type 316 to ASME SA 182 GR. F316.

This supplement also provides instructions for tagging the new relief valves. This supplement does not effect the safety evaluation since the substitute materials meet the design criteria for the valves.

S 1m n 2

, This supplement revises the expiration date for this EP in order to meet the latest guidelines and increases the set pressure for the new relief valves from 205.psig to 220 psig. The set pressure is being increased to ensure leak tightness. Also, a requirement for QC verification of proper valve tagging is added. This supplement does not effect the safety evaluation since the revised set pressure is within the design parameters for the DG Air Start System.

SAFETY EVALUATION This change consists of replacing existing safety related Class 3 relief valves in the DG Air Starting System because the existing valve vendor, Anderson Greenwood Co., cannot meet FPL QA requirements for replacement parts. The replacement valve, Dresser Industries, Inc.

Model 3/4-1975-3, and piping complies with the requirements specified in the St.'Lucie Unit '

FSAR Section 9.5.6 and has been demonstrated so by the preceding design analysis.

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. The modifications included in this EP do not involve an unreviewed safety. question because:

PC/M 046-285 Supplement 2 SAFETY EVALUATION(continued)

'he probability of occurrence or the consequences of an .accident or malfunction of equipment important to safety previously evaluated is not increased by this modification because the proposed designed replacement equipment maintains all of the original design requirements for the DG Air Start System. This ensures that the consequences of analyzed accidents remain unchanged and that the DG Air Start, System performs its designed function.

The possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report has not been created because there are no accidents that are initiated by malfunctions associated with this equipment. Therefore, no new accidents have been created.

iii) The margin of safety as defined in the basis for any Technical Specification has not been reduced.. The safety function that is controlled by the various applicable Technical Specifications, availability of Emergency Diesel Generators, is maintained by this change. The proposed design ensures the Diesel Generators will function as designed during an accident. Thus, the margin of safety is preserved.

10 CFR 50.59 allows changes to a facility as described in the FSAR if an unreviewed safety question does not exist and if a change to the Technical Specification is not required. As shown, the change proposed by this EP does not involve an unreviewed safety question.

Also, no change to the Technical Specifications is required.

In conclusion, the change proposed in this EP does not involve anunreviewed safety question, and does not require any change to Technical Specifications. Therefore, prior NRC approval is not required for implementation of the modification.=

PC/M 107-284 Supplement 0 ABSTRACT This modification will allow the operators to monitor DDPS calometric power and average RCS cold leg terriperature on the RTGB rather than the DDPS operators panel. It will also bring the Unit 2 RTGB jnto the same configuration as Unit 1.

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. The modifications included in this EP do not involve an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not i'ncreased by this modification because all equipment required to implement this change is designated as non safety related. The change will use existing cable.

The. panel meters to be installed will use already existing panel cutouts. A

.failure of any part of the above systems will not affect any safety related equipment due to separation and isolation between the two classifications.

The possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report has n'ot been created because there are no accidents that are initiated by malfunctions associated with this equipment. This PC/M does not change the function of any plant system.

The margin of safety as defined in the basis for any Technical Specification has not been reduced since no safety related equipment is modified, changed, or affected by modifications in this EP. This EP requires no technical specification changes.

The weight of the display assembly is negligible compared to that of the RTGB and will have no affect on the seismic integrity of the board.

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 nor a change to the Plant Technical Specifications; thus, prior NRC approval for the implementation of this PC/M is not required.

PC/M 089-284 Supplement 1 ABSTRACT There has been an excessive rate of failures on International Instruments, Division of Sigma Meter Models 9262, 9263 and 9264 installed at the St. Lucie Unit No. 2.

The problems reported by the l&C maintenance and operations personnel include:

1. Broken gears in the motor drive

2. Thirty (30) day recalibration frequency

3. Pointers remain fixed upon a loss of power to the instruments Instrument failures occur primarily when the plant is shutdown and in those units that monitor plant processes that exceed the zero and'span ranges of the units. Trouble shooting on some of the faulty units has indicated two (2) major modes of failure:

1. Mechanical failure - is almost entirely attributed to worn fiber gear teeth. These teeth are worn by the driving gear when the span and zero ranges are exceeded for long periods of time.

2. Electronic failure - is wider in scope but also limited to a) leaks in power supply capacitors, b) DC output voltage is higher than -11/+11 VDC and c) excessive ripples in DC level power supply among others.

Several meetings have taken place to determine different alternatives to correct the above described problem. One of those alternatives is the replacement of the Sigma Meters servo powered, self balancing type, models 9262, 9263 and 9264 by Sigma Meters Vacuum Fluorescent type models 9281, 9282 and 9285.

PC/M 089-284 provides the'method to temporarily install thirty-six (36) Sigma Meters Vacuum Fluorescent type in the St. Lucie Unit No. 2 control room RTG and HVC Boards. The purpose of this PC/M is to assess the vacuum fluorescent type, prototype meters, for reliability under actual plant conditions and Human Engineering Factors acceptance.

At present, the Vacuum Fluorescent Meter models 9281, 9282 and 9285 are not qualified to the requirements of IEEE 323-1974 and IEEE 344-1975. Therefore the prototype meters will be temporarily monitoring non-nuclear safety related variables.

PC/M 089-284 Supplement ¹1 provides a method to temporarily install five bargraph meters in the St. Lucie Unit 2 RTGB. The purpose of this supplement is to assess the replacement meters for reliability under actual plant conditions and Human Factors Engineering acceptance.

This Supplement will be implemented by FPL IKC Maintenance. Since it is anticipated that this modification will be performed during power operation, some of the alarms intrinsic to the meters may activate during installation and Plant Operation should. be alerted to this condition.

PC/M 089-284 Supplement 1

'AFETY ANALYSIS 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 basis 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:

i) The, probability of occurrence and the consequerices of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report, are not increased by this modification because it does not affect the availability, redundancy, capacity, or function of any equipment required to mitigate the effects of an accident. The meters only monitor non-nuclear safety related variables.

The installation of these meters is intended to assess its reliability under actual plant operation and to determine its acceptance from the human factors point of view (operators response). Even through the VFD meters are not qualified environmentally to!EEE-323-1974 and IEEE-344-.1 975, their use,in the control room is considered acceptable based on the following facts:

a) The design, materials and manufacturing of the cases that house the electronics of VFD meters'model 9281 and 9282 is similar to the Sigma Meter model 9270. The Sigma Meters model 9270 has been qualified by Combustion Engineering and Bechtel for use in the San Onofre Nuclear Generating Stations, Units 1, 2 and 3 in Nuclear Safety Related applications. Also Sigma Model 9270 being used in Arkansas Nuclear Generating Station and has been qualified foi Nuclear Safety Related

,applications by Bechtel Corp.

b) VFD Meter 9285 was seismically tested by Internatio'nal Instruments Division, as part of the performance tests on these prototypes, to the Combustion Engineering System 80 seismic response curve. This test has not been considered a formal test to IEEE-344-1975, however, it has indicated that VFD meters model 9285 will withstand an OBE 9 12.5 G and SSE 5 25 G with a damping factor of 1% maintaining its physical integrity and operability.

By installing the VFD meters in the control room the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not created since this modification is only temporary and involves only replacement of thirty-six (36) non-nuclear safety related meters.

PC/M 089-284 Supplement 1

'AFETY ANALYSIS (Continued) ii) The possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report will not be created by this mo'dification because the modification involves non-nuclear safety related equipment and failure of any items added by this modification will not impact any nuclear safety-related functions.

iii) The margin of safety as defined in the bases for any technical specification is

'ot affected by this modification since the temporary replacement of these non-nuclear safety related meters does not reduce the margin of safety and does not-involve a change nor are any of the variables monitored co'vered by

'echnical Specifications.

The installation of the thirty-six (36) prototype VFD meters does not degrade the operability .

of any system required for the safe shutdown or integration of an accident since they will monito'r only non-nuclear safety related variables and in some cases will provide alarm. None of the meters alarm contact outputs are interlocked with any system, therefore failure of any of the meters will not result in malfunction of any nuclear safety related system.

The foregoing constitutes per 10 CFR 50.59(b), the written evaluation which provides the

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basis that this change does not involve an unreviewed safety question, therefore prior Commission approval is not required for the implementation'of this temporary PC/M. ~

f Eval a ion Su Iemen ¹1

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The temporary replacement of these non-nuclear safety related meters does not constitute a deviation from the original EP's Safety Analysis, which remains valid, and applicable to this

, modification.

A review was performed to determine if the implementation of this supplement would have an adverse effect on the overall seismic qualification of the RTGB. The results of the review indicate that the new'eplacement rrieters will remain intact during seismic event and will not impact the structural integrity of the control boards. Therefore, it was concluded that they are acceptable for use on a temporary basis. Details of the seismic evaluation are provided in Attachment ¹1.

St. Lucie Unit 2 Docket No. 50-389 Generic Letter 89-04 Initial Ten-Year Inservice Inspection Interval Inservice Testin Pro ram Revision 2 ATTACHMENT 1

SUMMARY

OF RELIEF RE UEST CHANGES In response to the NRC Generic Letter 89-04 and the NRC safety evaluation (SE) dated December 5, 1991, FPL has amended the St.

Lucie Unit 2 Pump and Valve Test Program. The majority of the changes are consistent with the requirements of ASME Code or the recommendations of Generic Letter 89-04. Those reliefs that are not in accordance with or were not covered by the code or the generic letter are summarized below:

Relief Re uest No. PR-1 This relief request applies to all pumps in the program. A yearly bearing temperature measurement does not contribute to the monitoring of pump operational readiness during its service life. Concerns regarding extended pump runs while on mini-flow recirculation and ALARA concerns out weigh any benefits obtained from one yearly temperature data point.

Relief Re uest No. PR-2 This relief request applies to only portable instruments used for temperature and speed measurement. Many of the portable instruments used have digital readouts with multiple scales.

These instruments are well within the accuracy requirements of the ASME Code, but cannot meet the upper scale limit of three times the reference values.

Relief Re uest No. PR-3 This relief request applies to all pumps in the program. In IWP-3400, Frequency of Inservice Tests, the code states that a pump need not be stopped for a special test. Along similar lines, a running pump should not have to be stopped in order to obtain its stopped suction pressure.

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Relief Re uest No. PR-7 This relief request applies to 'the diesel -fuel'il transfer pumps. The Unit 2 diesel fuel oil system does not have any installed flow instruments. Fuel oil must be pumped between

-*..storage"tanks in"-order'to measure. pump flow rates. Relief is requested to perform this type of test.. In addition, the relief also describes the modified allowable ranges used to gage the pump operational readiness. The method establishing the new allowable limits was included in this

'f relief request for information only since both the ASME Code

.(IWP-3210) and the ASME Interpretation 'XI-1-79-19 permit the owner to modify the allowable ranges.

Relief Re uest Nos. PR-8 PR-11 PR-15 and PR-16 These relief requests are similar and apply to the boric acid make .up, intake cooling water, containment spray, high

", pressure safety injection, low pressure safety injection, and diesel oil transfer pumps. Many plant systems do not have

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suction gages installed. These reliefs request permission to use tank level as an approximation for suction pressure during testing. The flowrates experienced are low enough that any head loss in the suction piping is not significant.

Relief Re uest Nos-..PR-12 PR-'3 and PR These relief requests are similar and apply to the reactor coolant charging, intake cooling water, and hydrazine pumps.

The ASME Code -requires .the lower frequency .response, of the vibration instruments to be at least one half of the rotational frequency. For the pumps in these relief requests,

.. " ~instruments that" meet- the ~"low" frequency"specification are either impractical or impossible to obtain.

Relief Re uest No. PR-17 This relief request applies to the hydrazine pumps. In IWP-4150, Fluctuations, symmetrical damping devices or averaging techniques are addressed as a means to dampen any fluctuations in measurements. The 2A and 2B hydrazine pumps are low flow, positive displacement, metering pumps. The relief request describes the steps taken to correlate the pump speed (rpm) with pump average flow. The pump flow will be verified quarterly by measuring speed (rpm) and a full flow test will be performed each refueling outage.

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-These relief requests -apply.to.valves in"the safety injection system. These valves are designated as pressure isolation valves. The relief request stipulates that these valves will

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Relief Re uest No. VR-4 This relief request applies to containment isolation valves.

Due to the piping configuration, several containment isolation valves cannot be tested individually. This relief request is that the valves in these penetrations be tested simultaneously in multiple valve arrangements. A maximum leakage rate will be assigned to each combination of valves.

Relief Re uest No. VR-13

,,The"four check valves in this request are the safety injection tank (SIT) discharge check valves. They are normally closed to isolate the SITs from the high pressure safety injection system. During a large break LOCA, these valves would open to allow the SITs to discharge into the reactor coolant system (RCS). The large flow rates expected during this discharge (20,000 gpm) are too large to be reproduced during cold shutdown or refueling conditions. The only other recourse is disassembly for these valves. 'FPL agrees to disassemble these four valves but requests that the test interval be extended to once every 10 years instead of once every 6 years.

Relief Re uest No. VR-14

-The four check,.valves. in this-.request are the safety injection header discharge check valves. They are normally closed to isolate the safety injection system from the high pressure RCS. During a large break LOCA, these valves would open up to allow the SITs to discharge into the RCS. The large flow rates expected during this discharge (20,000 gpm) are too large to be reproduced during cold shutdown or refueling conditions. The only other recourse is disassembly for these valves. FPL agrees to disassemble these four valves but requests that the test interval be extended to once every 10 years instead of once every 6 years.

Relief Re uest No. VR-17 The valves in this relief request are containment isolation valves in the waste management system. Due the piping configuration, this check valve cannot be verified closed

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"except by performing a backflow leak -test. In order to perform the leak test,.the containment penetration isolation

.would-be reduced to only one valve. -Therefore, .the backflow test could only be performed during Modes 5 and 6.

Relief Re uests Nos. VR-18 and VR-25 The check valves in these relief requests are part of the containment isolation for the make up water and the sampling systems. To test them requires that each penetration be reduced to only one valve isolation. Therefore, they cannot be tested while containment isolation is required.

Relief Re uest No. VR-19 This relief request applies to the instrument air system containment penetration. The check valve in this relief request 'cannot be tested without isolating instrument air to over 50 valves, instruments, and controllers inside containment. The test cannot be performed unless a back up compressor is connected to the isolated air header. Testing of this valve would be an unreasonable burden to perform during cold shutdowns.

Relief Re uest No. VR-24 This relief request applies to containment spray system valves. The two check valves covered by this relief request are the hydrazine 'pump discharge check, valves. The only way to full flow test these check valves is to inject the containment spray system. Full flow testing of these

'hydrazine'nto check valves 'is requested to be limited to refueling outages to minimize the amount of hydrazine injected into the

, containment system and the refueling water tank.

Relief Re uest No. VR-26 This relief rec{uest applies to emergency diesel generation air start system valves. The valves in the diesel generator air start system are integral with the system with no valve position indication mechanism. There is no practical method for measuring the stroke times of each individual valve.

Valve failures could be detected by the monthly diesel generator tests and the six month air valve actuation test.

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Relief Re uest No. VR-30 This relief request applies to -'the safety injection system valves. St. Lucie was granted a one time relief by the NRC for the disassembly of V-3103-.during the Unit 2 1992 refueling.

outage...Both check valves, will.,be.disassembled and..inspected.

during the next refueling outage scheduled for the Fall..of.

1993.

Relief Re uest. No. VR-33 This relief request applies to valves in the feedwater system.

These four solenoid valves require flow or a differential-pressure to stroke properly. Pumping from the auxiliary feedwater system into the steam generators during normal operation is not practical nor desirable. Therefore, the valves can only be stroke timed during cold shutdowns. They will be exercised quarterly under no flow conditions. Stroke times'ill be measured quarterly but will not be trended for alert testing. The stroke times will be verified to be less than the maximum allowed stroke time.

Relief Re uest No. VR-34

. This relief request applies to valves in the component cooling water system. These two valves will be stroke tested quarterly and the stroke times will be measured. However, due to the unavoidable variations in the stroke due to the test method, the stroke times will not be trended for alert testing. The stroke'time will be verified to be less. than the maximum allowed stroke time.

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2 Docket No. 50-389

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.,Generic Letter 89-04 Initial Ten-Year Inservice Inspection Interval Inservice Testin Pro ram Revision 2 ATTACHMENT 2 REVISION 2 FIRST TEN YEAR INSERVICE INSPECTION PROGRAM FOR PUMPS AND VALVES

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