ML17308A499

Rept of 10CFR50.59 Plant Changes, Covering 890123-900122. W/900713 Ltr

ML17308A499
Person / Time
Site:	Saint Lucie
Issue date:	01/22/1990
From:	Sager D FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L-90-215, NUDOCS 9007200199
Download: ML17308A499 (89)

Text

'<CCELERATED DIS- IBUT ON DEMONST ION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9007200199- DOC.DATE: 90/01/22 NOTARIZED: NO DOCKET. 4 FACIL:50-335 St. Lucie Plant, Unit 1, Florida Power & Light Co. 05000335 AUTH. NAME AUTHOR AFFILIATION SAGER,D.A. Florida Power 6 Light Co.

RECIP.NAME RECIPIENT AFFILIATION

%go~~

SUBJECT:

"Rept of 10CFR50.59 Plant Changes," covering 890123-900123.

W/900713 ltr.

DISTRIBUTION CODE: IE47D COPIES RECEIVED:LTR ENCL SIZE: D TITLE: 50.59 Annual Report of Changes, Tests or Experime ts Made W/out Approv NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-2 LA 1 .0 PD2-2 PD 5 5 NORRIS,J 1 0 D INTERNAL: ACRS 6 6 AEOD/DOA 1 1 D AEOD/DS P/TPAB 1 1 RRQDLPQ+LH~1 1 1 NRR/DOEA/OEAB1 1 1 1 G FILE (f2 1 1 RGN2 FILE 01 1 1 EXTERNAL: LPDR 1 1 NRC PDR 1 1 NSIC 1 1 D

D D

NOTE TO ALL "RIDS" RECIPIENTS:

I'LEASE HELP US TO REDUCE AVASTE! CONTACT THE DOCUMENT CONTROL DESK, ROON I PI-i7 (EXT. 20079) TO ELli~llNATEYOUR YAillEFROibl DISTRIBUTION LI(TS FOR DOCUb,!ENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 22 ENCL 20

r vy ~~) i'~~'1 ~ > t'v P,U t<t ~140(e, Juju f5':ash I I 33408 0420

'JUL 1 3 1990 L-90-215 10 CFR 50.59 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gentlemen:

Re: St. Lucie Unit 1 Docket No. 50-335 Re ort of 10 CFR 50.59 Plant Chan es Pursuant to 10 CFR 50.59(b)(2), the enclosed report contains a brief description and summary of the safety evaluation of Plant Changes/Modifications (PCMs) which were made, and are reportable, pursuant to 10 CFR 50.59. Included with the brief description of each PCM is a summary of the safety evaluation completed by Florida Power 6 Light -Company for that PCM. This report includes PCMs completed between January 23, 1989, and January 22, 1990, and correlates with the information included in Revision 9 of the Updated Final Safety Analysis Report submitted under separate cover.

Very truly yours, D. A. ' ger Site President St. L ie Plant DAS/DMB/gp Enclosure cc: Stewart D. Ebneter, Regional Administrator, Region Senior Resident Inspector, USNRC, St. Lucie Plant II, USNRC

ENCLOSURE ST. LUCIE UNIT 1 REPORT OF PLANT CHANGES MADE PURSUANT TO 10 CFR 50.59 JANUARY 23 I 1989 TO JANUARY 22 ~ 1990 90072007.99

PLANT CHANGE/MOD REPORTABLE PURSUANT TO 10CFR50.59 FOR ST LUCIE UNIT 1 FSAR AMENDMENT 9 NIJHBER SUPPLEMENT TITLE 125-181 BREATHINQ AIR SYSTEM 137-181 CHARGING PUMP MODIFICATIONS 281-183 WATER TREATMENT PLANT CONDUCTIVITY MOD 027-185 CONCENTRATOR HI&I LEVEL FEED PUMP 3RIP MOD 082-185 PROTECTIVE COATING FOR INTERNALS OF ICW PUMP COLUMNS 154-185 RCP MOTOR MODIFICATIONS 158-185 STEAM GENERATOR TUBE PULLING & TUBESHEET PLUGGING 018-186 0-4 SPENT FUEL POOL RERACK 050-186 2-3 INSTRUMENT AIR UPGRADE 074-186 HEATER DRAIN PUMP DEMINERALIZED WATER SUPPLY 149-186, PRESSURIZER MANWAY COVER STUD TENSIONER ATTACHMENT HOLES 152-186 DIESEL GENERATOR OIL LEVEL ALARM 154-186 CCW PUMPS SUCTION PRESSURE GAUGES 001-187 Z&E BULLETIN 853 MOV SWITCH SETTINGS 076-187 ERDADS/SAS UPGRADE 006-188 RCP SEAL COOLER HX LEAK DET 025-188 SFP TEMP JIB CRANE 033-188 INST CHGS FOR HUMAN FACTORS 069-188 DOUBLE INSUL EXC BEARIlS PED 269-188 GUIDE TUBE PLUGGING DEVICE REMOVAL 018-189 COND PUMP CURR BAL RELAY 139-189 STM GEN TUBE PLUG REPL 239-984 0- SODIUM HYPOCHLORITE INJECTION PUMPS 199-985 WATER TREATMENT PLANT REGENERATION WASTE NEUTRALIZATION TANK 053-987 CONDENSATE POLISHER CROSS-TIE INS 0198L/

PCM 125>>181 PAGE 1 OP 2 BREATHING AIR SYSTEM FUNCTIONAL AND DESIGN RE UIREMENTS Function The function of the Breathing Air system is to provide clean, cool, dry air to personnel inside containment. This air wQI provide a source for proper and comfortable respiration through a filter and mask system while personnel are working hside containment.

1.2 Design Requirements The major design requirement of this sytem is to provide a source of breathable air which has a temperature no greater than 120oF, low water content and low carbon monoxide. An additional requirement is to provide this breathable air during all modes of plant operation..

SAFETY ANALYSIS A. Containment Isolation Valve la) With respect to the probability of occurrence of an accident previously evaluated in the FSAR:

The service air system is not utilized in the determination of accident probability and therefore this modification would not increase the probability of any accidents previously evaluated in the FSAR.

b) With respect to the consequences of an accident previously evaluated in the PSAR:

The automatic containment isolation valves were installed in accordance with the containment Isolation criteria specified in the St. Lucie Unit $ 1 PSAR. Therefore, no accident discussed in the PSAR would be made more severe.

c) With respect to the probability of malfunction of equipment important to safety previously evaluated in the PSAR!

The equipment prescribed in this package is of similar design or better than that already installed in the plant so the probabQity of an equipment malfunction important to safety would remain unchanged.

d) With respect to the consequences of malfunction of equipment important to safety previously evaluated in the PSAR:,

For the reasons stated in 1b) above the seriousness of equipment malfunction for equipment discussed in the above accident analysis of the FSAR would not be increased by the addition of the new valve.

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PCM 125-181 PAGE 2 OF 2 2a) With respect to the possiblity of an accident of a different type than any analyzed in the FSAR:

This modification does not decrease the design margins of the service air system, change the. operating conditions or functions or affect other safety related equipment. Therefore this change does not create the possibility of an accident of a different type than that considered in the FSAR.

b) With respect to the possibQity of malfunction of a different type than any analyzed in the FSAR:

For the reasons discussed in 2a) above the addition of the containment isolation valves does not create the possibility of malfunction of equipment not presently considered in the FSAR.

3) With respect to the margin of safety as defined in the basis for any Technical Specificaton:

The containment isolation valves will be tested periodically in accordance with the requirements of Appendix J to 10 CFR 50 and the St. Iucie Unit 81 FSAR. The valves meet the criteria spelled out for containment isolation and therefore would not decrease any margin of safety discussed in the Technical Specification.

B. Aftercooler and Associated Pi in and Valves The change described herein of installing a breathing air aftercooler may be classified as a nonnuclear safety related change based on the following:

1. The service air system and the turbine plant cooling water system are both non-nuclear safety related systems. Therefore, a change to any component in these systems will also be a non-nuclear safety related change.

2. The applicable piping code utilized for this change is ANSI B31.1, code for non-nuclear safety related piping.

3. Failure of equipment involved in this design change will not affect any safety related structures, systems or components, because no safety related equipment is adjacent to the service air system.

PCM 137-181 PAGE 1 OR 2 CHARGING PUMP MODIRICATIONS ABSTRACT This supplement provides new PC/M drawings, revises one and deletes one drawing. This supplement also removes changes made by supplement 7 which was cancelled by the plant. The safety classification does not change, the safety evaluation remains valid and the Technical Specifications are not affected as a result of this revision.

I. SYSTEM DESCRIPTION 1.0 Performance R'e uireme ts The Charging Pump Modification (CPM) shall perform the following:

l. Improve circulation of water to plunger packing

2. Provide ceramic coated plungers

3. Provide a vent for the pump cylinder block.

This will increase the reliability of the plunger packing.

2.0 ~0 The CPM will provide one seal water line to each plunger chamber from the seal water tank instead of the present one line per pump.

Ceramic,plungers will be provided, installed and then evaluated for adequate performance. The vent will be installed in the pump discharge blind flange to vent the pump cylinder block before starting the pump.

SAFETY ANALYSIS 1.1 4'ith respecc to the probability of occurrence of an accident previously evaluated in the FSAR:

Acc'Vents previously evalpated in the FSAR which involve portions of che CVCS included in this PC/M are "Charging Pumps", which is described in Table 9.3-25 of the FSAR and a Chemical and Volume Control System ~ialfunccion" whici> is described in Section 15.2.4 of the FSAR.

The possibility of occurrence of these accidents will not increase since the modification is of equal or improved design, materials of construction, fabrica-tion, installation and testing standards.

1.2 4'ich respect to the consequences of an acc'dent previouslv evaluaced in the FSAR:

The compensacing provision for the "Charging Pumps" are:

Low level in pressurizer will start second and third pump. This modification will not change these compensa=

ting provisions: therefore, this accident will not be made more severe.

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r'm i3I-ill PAGE 2 OF 2 This modification will noc change the assumptions or corrective actions used in the analysis of a chemical and volume control system malfunction: therefore, this accident will not be made more severe.

1.3 arith respect to the probability of malfunction of equip-mene important to safety previously evaluated in the FSAR:

As stated by the pump manufacturer, this modification will not prevent the charging pump from withstanding the loadin"s Eor which ir. was designed. For this reason. and the reasons staced in la and b above, the probability of malfunction of equipment important to safety previously evaluated in the FSAR would not increase.

1.4 lilith respect to the consequences of malfunction of equip-ment important to safety previously evaluated in the FSAR:

For th>> ress >>>s state<i ln la, h 'liid c above, the prohabl I-ity of:n'll.f<<>>ccion of <<q<<ipmu>>t important to safL'ty pre:i )<<sly evaluated l.n the l'SAR wo<<ld nut increase.

2.1 With respect to the possibility oE an accident of a different type rhan any analyzed in the FSAR:

As stated by rhe pump supplier, this modification will noc decrease.

the design mar~ins of the charging oumps. This modification will also noc chan'e the operating functions or conditions, or. effect other safety-related equipment. Therefore, this change would not create the possiblity of an accident not considered in the FSAR.

2.2 With respect to the possibilitv oE malfunction of 8 different type than any analyzed in the FSAR:

For the reasons discussed in 2a, the modification would noc creace the possibility of malfunction of equipment not considered in the FSAR.

2.3 lkich respect co che marjin of safety as defined in the basis for any Technical Specifications: '

For the reasons discussed in la, lc and 2a, the modification will not decrease any margin of safety discussed in the Techni-cal Specificacions.

CO'.lCLUSIOX:

The Charging Pump Modification does not involve an unreviewed safety question.

PCM 281-183 PACE I OF 1 WATER TREATMENT PLANT. CONDUCTIVITY MOD INTRODUCTION The function of the make-up system is to supply treated, demineralized water of the required quality for make-up use to various systems of the plant. 'Water for the make up system is pumped by two city water transfer pumps thru the water treatment system. This PC/M covers the necessary control circuit modifications to the Water Treatment Pump (WTP) to isolate the final effluent valve and trip the Treated Water Storage Tank (TWST) transfer pumps.

SAFETY 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 occurence or the conse-of an accident or malfunction of equipment important to quences 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 ana'is report may be'reated; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

The make up water system serves no safety function, since it is not required to achieve safe shutdcwn or mitigate the consequences of a .

LOCA.

Since the Water Treatment Plant 'controls and instrumentation are non-safety related, the modification under'this PC/M does not create any new accidents or malfunctions. Similarly, no previously evaluated equipment malfunction is increased nor does any technical specification need to be modified. Therefore, no unreviewed safety questions con-cerning this modification exist and prior commission approval is not required.

The following FSAR Sections/Tables/Figures are referenced:

Make-up Water System -Section 9.2.5 Design Data for MWS Components -Table 9.2-13 Flow Diagram Make>>up System -Figure 9.2-5 The following Technical Specification Section/Tables/Figures are referenced:

Secondary Water Chemistry -Section 3.7.1.6 019 8L/

PCM 027-185 PAGE 1 OF I CONCENTRATOR HI-HI LEVEL FEED PUMP TRIP MOD INTRODUCTION At present, the Concentrator Feed Pumps vill not trip on hi-hi level vhen both concentrators A 6 B are on line. This PC/M vill modify the present design to provide a Hi-Hi level trip signal from either Concentrator A or B. This modification is required to prevent boron carry"over into the vapor separator, vapor condenser and distillate system ~

SAFETY ANALYSIS With respect to Title lO of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unrevieved 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.

This modification is for the disconnectXbn'end reconnection of viring betveen non-safety/non seismic relay zozgacce This change vill permit the tripping of the feed pumps from eitter Concentrator h or B Hl-Hl level. Werefore this modification vill not increase the probability of the occurrence of any accidents vhether previously evaluated or of a different type than previou~ evaluated and vill not reduce the safety of the plant. I The foregoing constitutes, per 10CFR50.59(b), the vritten safety evaluation vhich provides the basis that this change does not involve an unreviewed safety question, therefore prior Commission approval is not required for'mplementation of this PCM.

This PCM does not reduce the margin of safety as defined in the basis of any technical specification, nor does it require a revision of a technical specification.

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PCM 082-185 PAGE 1 OF 1 PROTECTXVE COATIN" FOR IKZHNALS OF ICV PUMP COLUMNS DESCRIPTIOH Protective coating will be provided on all internal surfaces of the spare ICW pump columns which are subJect to corrosfon by the seawater environment. The application of this coating will greatly minimize the corrosion thereby preserving the fntegrity *of the pumps.. The coating material will be Corroglass or engrg. approved equal applied in strict accordance with the manufacturer's recaanendations.

Q~ EVALUATION This modification has been reviewed with respect to Title 10 of the Code of Federal Reguiat io'ns, Part 50.59 which states that a proposed change sha I 1 be deemed to involve an unrev I ewed saf ety question I f: ( I ) the probability of occurrence or the consequence of an accident or malfunction of equipment Important to safety previously evaluated,ln the safety analysis report may be increased; or (ii) a possibility for an accident or malfunction of a different type than any evaluated previously fn the safety analysfs report may be created; or (ill) the margin of safety as defined in the basis for any technical specifcation Is reduced.

(I)- The means of bonding the, glass flakes with the polyester resin to form Corroglass and application with strict adherence to manufacturer's recommendation renders the coating superfor to granular-filled coatfng materials. The coatfng's resistance to delamination combined with the minimal amount of coatfng available for delamination, as compared with the amount of exfsting coating upstream of the ICW pumps, Is not deemed to increase 'he probability of an accident prevfously evaluated.

(ii)- Protective coatings have been applied to pumps and pump columns as well as the intake bay w'alls, with documented success, at St. Lucie and other FPSL locations. Therefore, the posslbilfty to create malfunctions not previously eva'luated in the Final Safety Analysis Report has not been fnltiated.

(iii)- The ability of the ICW'ystem to provide a heat sink to essential systems during accident condftfons has been assured through redundant headers to fulfill the requirements of the Component Cooling HXs.

Also, strainer by-pass line or cross-connect valves from alternate cooling water supply may be opened downstream of strainer. This modification is not deemed to reduce the margin of safety as defined in the basis for any technfcal specfficatlons.

The above evaluations provide the bases pursuant to 10CRF 50.59'or the determination that this modiffcation does not fnvolve an unrevlewed safety question.

0198L/

PCM 154-185 PAGE 1 OF 2 RCP MOTOR MODIFICATIONS ABSTRACT This Engineering Package covers modifications to the Unit l Reactor Coolant Pump (RCP) Motors to increase the motors'eliability by making them less susceptible to (l) oil seal damage caused by excess shaft vibration and (2) anti-reverse rotation pin damage. The major features of this packge include modifications to the "stovepipe" oil seals, oil slinger, and anti-reverse rotation pins. These modifications are classified as Quality Related and do not constitute an unreviewed safety question.

l l"Gal JR-LOD PAGE 2 OF 2 0 SAFETY EVALUATION The Unit 1 FSAR Section 5.5.5 discusses the design bases of the RCP's and RCP motors. The RCP motors function to provide motive force for the pumps during normal plant operation and provide inertia to improve coastdown characteristics during a loss of pump power condition. The motors also contain an anti-reverse rotation device to preclude reverse rotation caused by backf low through the pump impel er. 1 Potent I a 1 causes of backf low considered are: ( 1~ Reversed power leads (2) Loss of power to one RCP, with others operating (3) and RCP suction line break; LOCA.

The greatest challenge to the anti-reverse rotation device (ARRD) is the reverse torque due to a large break LOCA. Combustion Engineering (in enclosure 1 to F-CE-8746) has analyzed the ARRD in regards to the reverse torque from LOCA loads. Their analysis shows that a minimum of four

~

equally spaced ARRD pins per motor of the correct mate al (SA193) are adequate to withstand the maximum specified reverse torque.

The modifications performed to the ARRD pins, (chamfering at edge) has not reduced the number of pins, and therefore the ARRD's ability to function as-required has not been altered.

The Unit 1 FSAR Chapter 15 discussps'three accidents which are affected by loss of one oI more RCP motors: <1)Section 15.2.5 (Loss of Coolant Flow Accident), (2>Segtion 15.2.9 (Loss of Offsite Power to the Station Auxiliaries) and <3) Section 15.3.4 (Seized Rotor Event). A loss of normal coolant flow may result either from a loss of electrical power to one 'or more of the RCP's or from a mechanical failure, such as a pump shaft seizure. Simultaneous mechanical faiure of two or more pumps is not considered credible. Under normal operating conditions, no single failure will result in a complete (four pump) loss of flow accident. Under complete loss of offsite power conditions, the pump-motor-flywheel-comblnation provides the source of coastdown energy; The RCP motor manufacturer, Siemens-Allis, has evaluated the proposed modifications and has concluded that the capability of the motor to perform its function will not be affected. Since the normal operation of the RCP motor is not affected by these modifications, the probability of occurrence of a design basis accident or malfunction of equipment important to safety as previously evaluated in the FSAR wi 11 not be increased. The modifications do not change the postualated failure modes of the RCP motors which would result in loss of reactor coolant flow. - The consequences of a design basis accident or malfunction of equipment important to safety as previously evaluated in the FSAR will therefore not be increased. The possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR will not be created.

The Unit 1 Technical Specifications section 3/4.4 discusses the operability requirements of the Reactor Coolant System. The proposed modifications do not alter this Technical Speclflcation or its bases. This PC/H, therefore, does not reduce the margin of safety as defined in the basis for a Technical Specification.

ln conclusion, .the changes proposed in this design package are acceptable from the standpoint of nuclear safety; do not involve an unreviewed safety question; do not require NRC approval prloi to implementation.

PCM 158-18S PAGE 1 OP 4 STEAN GENERATOR TUBE PULLING & TUBESHEET PIUGGING Introduction Zt is desired to perform a metallurgical exam an parl~s of selected ~em generator U-tubm fran the St. tucie Unit 1A Steam Generator.

Gris modification wiU. reanove previously installed steam generator U-tube plugs fran selected tubes on the reactor coolant inlet plenum (hat leg side), ranave portions of the steam generator U-tubes designated far metaDurgical emanation, install steam generator U-tube stakes as respired and inst~ welded tubesheet plugs.

this mcxiificatian has no effect an the previously existing number of. steam genexatar U-tubes available for heat transfer since the only U-tubes MIng mcx3ified axe those which were previously plugged and hence unavailable for heat transfer service+

'Ibis nedificatim involves five basic cperatians an each designated steam genexator U-tube:

(a) removal of the plug in the hot leg side of the U-tube (b) cuttizg the U-tube fnxn the plenum cutting aut the U~ to tuR~m4 weld

'c)

(d) pulling the cut pcaMan of the U-tube aut of the tubesheet (e) installatica of a welded plug/stake assembly (if recgdrai) into the tubesheet.

~ U"tubm designated for cutting by Reference (C) are as follows:

Raw 14 Line 32 Raw 7 Line 153 Raw 7 Line 129 Line 53 019 8L/

PAGE 2 OF 4 Alternate U-tubes also designated hy Reference (C)care as follcms:

Raw 15 Line 149 Rear 25 Line 49 Raw 99 Line 55 Raw 94 Line 100 Raw 65 Line 65 Rcv and line designatians are M accordance with Reference (D) .

The general saunce of cgeratians is as follows:

(1) Remove the steam generator U~ plugs fran the hot leg side of the four designated ~axe.

(2) Cut the four U-tubes at an elevation of ~aximately 175 3/8" above. the prhaary side surface of the steam generator tulxel;Ieet. ~ location of this cut is to be gust below the fourth full egg crate. By ma3dng the cut within ~aximately 2 inches of this egg crate, subsequent stakizg of the remaining portian of the aff~ U-tube wiU. not be necessary unless the full portion of U-tube intended to be removed M not successfully valved.

(3) &knave the caaplete cut length of the U-tubes located row 7, line 129 and re+ 35, line 53.

(4)

E~ZA3 the iIlitial78 to 80" of the U-tube located at re 14, line 32 and the initial 39" to 41" of the U tube located at nw 7, line 153.

(5) Perform a video eMmination of all four affected U-tube locatians.

(6) Remove the reining length of the U-tube located at mv 14, line 32 and the next 39" to 41" of the U-tube located at row 7, line 153.

4 I 0

i'CH 15a-i85 PAGE 3 OF 4 (7) Perform a video maminatian of the U-tube locatians of step (6).

(8) Remove the rmaining length of the U-tube located at row 7, line 153.

(9) Perform a ~ho examination of the U-tube located at re 7, line 153.

(10) Reaxnw a sludge sample if sludge is observed.

(11) Install tube.'~ plugs in the four designated locatians.

(12) Transfer the cut partians of the steam generatar tubes to the designated facility for metallurgical examinatian.

Safe With respect to Title 10 of the Gx1e of Federal Regulatians, Part 50.59, a prcposed change shall be deeined to involve an unreviewed safety question:

accident ar malfunctian of equipnent important to safety previously evaluated in the safety analysis report may be increased p 01 (ii) if a possibiLLty for an accident ar malfunctian 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 f'r any technical specification is reduced.

For the follcmiing reasans, C-E ~udes this change does nat involve any unresolved safety questians as defined in items (i), (ii) or (iii) abave:

P~;.'~8-iba PAGE.4 OF 4

1. Ghe modificatians descry+.d in this PC/M do not zemcve any steam genexator U-tubes fran service which were gzevimsly in sexvice. Ghe tu)~M.et plugs, after welding, have a small, but cantxolled, pxa4zusian fran the tubesh~ face into the reactor coolant fleer path. ~ fleer and other perfonnance characteristics will not be affected.

modifications described in the PC/M have no effect an the steam generator's abi1ity to perfaxm its designed purpose.

2. Ghe md'.catians described in this PC/M do not affect any te~cal specificatian or the basis for any technical specification.

3. 'y cutting the designated steam generator U-tubes within a distance of agpvmhaately two inches below a full egg crate thee wiU. be no significant impact an the secandaxy side performance or intaepCty of the stean genexator.

4. Any cuts made outside the limitations impcmed by (3) abave, or any portions of a cut U-tube which cannot be

'fully withdzawn wi11 necessitate insertion of stakes fran U-tube. ~

the welded plug into the remaining portian of the affected action will ensure that no significant impact an the secandaxy side perfoxmance ar integrity of the steam genitor cxoaua as a result of making the cut away fram an egg crate ar leaving @casions of a cut ~ube installed.

5. 'Ihe structure. adequacy of the steam genexatoxs, as required by the ASME Boiler and Pnmmme Vessel Ocde, Section ZZZ, 1971 editian, and addenda thxough Summer, 1972, is not affected by the mcjdificatians de+~ed in this PC/M.

Ghe implementation of the PC/M does not requixw a cthe to the plant technical specificatians.

She foregoing constitutes, per 10 CFR 10.59(b), the written safety evaluatian which pxavides the basis that this change does not involve an unxeviewed safety question and prior Cmanissian approval for the implaaentatian of this PC/M is not required.

I PC@ 018-186 PAGE 1 OP 12 SPENT FUEL POOL RERACK ABSTRACT This Engineering Package details the requirements for the design, fabrication, installation and inspection of high density spent fuel storage racks to replace the existing racks presently installed in the spent fuel storage pool. The purpose of this modification is to increase the number of spent fuel assemblies that can be stored in the pool from the present 728 assemblies to 1706 assemblies, thereby extending the onmite spent fuel storage capability until the year 2009.

The storage of consolidated fuel assemblies has been considered in the structural design of the spent fuel storage racks, and each individual storage cell has 'been sized to accept a fuel consolidation canister. However, additional analyses must be performed and NRC approval obtained prior to "

storage of consolidated fuel in the spent fuel racks.

This modification is classified Safety&elated because it interfaces with the spent fuel pool and the spent fuel cooling system, which will involve are Safety&elated.

1his modification has been evaluated in accordance with 10CFR50.59 and found to require a change to the Technical Specifications. Since prior NRC approval is required, a proposed amendment to the Facility .Operating License regarding this modification has been submitted to the NRC and prior Commission approval is required before this PCM may be implemented. This modification has been reviewed in accordance with 10CFR50.90 and 50.92 and found not to involve a significant hazards consideration and will have no negative effect on plant safety or operation.

The implementation of this modification requires changes to the plant Technical Specification and thus NRC approval is required prior to implementation. 2hese changes are discussed in detail in Section 3.0 of this Engineering Package.

Prior to the implementation of this modification, the plant shall verify with Engineering that the Engineering Package has been revised as required to reflect resolution of any issues raised by the NRC in granting a license amendment.

Su lement 1 This supplement provides details for the indexing strip supports and removes several hold points. The vendor manual/drawing list is updated to include the latest revisions of all the Joseph Oat drawings and the Joseph Oat and NUC procedures.

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rC"x veau-id'AGE 2 OF 12 This supplement retains the Safety Related classification originally designated. The modifications provided by this supplement do not..introduce an unreviewed safety question as defined in 10CFR50.59, nor do they alter any Technical -Specifications.

Su lement 2 This supplement provides a. modification to the 'procedure to

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that lift this the existing change will spent fuel racks and a safety evaluation to demonstrate not give rise to an unreviewed safety question, will not impact the sang operation of the Plant and does not require a change to any technical specification. Since this change concerns the spent fuel racks, this PCM supplement retains the original safety related classification.

Su lement 3 This supplement provides a modification to the procedure to install the temporary construction crane and a safety evaluation to demonstrate that this change will not give rise to an unreviewed safety question, will not impact the safe operation of the Plant and does not require a change to any technical specification. Since this change concerns the spent fuel racks, this PCM Supplement retains the original safety related classification.

This supplement provides the following:

a. A revision to the safety evaluation to address the installation of the

'indexing strip system.

b. A revision to the design bases and design analysis to replace the.

'conditions associated with a 24 month refueling cycle wi,th those for an 18 month refueling cycle.

c. A revision to the design bases, design analysis and safety evaluation to address'he impact of the sparger modification (which was implemented via PCM 142-186) on the local thermal hydraulic and bulk decay heat analyses for the racks.

This supplement also reissues, in a typed format, pages 29a through 29e,

which were previously issued handwritten as part of Supplement 3.

The safety evaluation demonstrates that the changes introduced by this supplement will not give rise to an unreviewed safety question, will not impact the safe operation of the Plant and do not require a change to any technical specification.

Since these changes concern the spent fuel racks, this PCM supplement retains the original safety related classification.

riii vib-iob PAGE 3 OF 12 SAFETY EVALUATION Safety 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.

With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed modification to a utilization facility requires prior NRC approval to accomplish i.f the modification requires a change to the facility operating licensing. Because this modification involves an operating licensing change, Reference 6.13 has been submitted to the NRC in accordance with 50CFR50.91 and 50.92. This submittal provides a description of the change and supports a finding of no significant hazards considerations for the modification. Implementation of this PCM cannot be accomplished until approval is obtained from the NRC. The following evaluates the rerack against the criteria of 10CFR50.92:

i) Involve a significant increase in the probability or consequences of an accident previously evaluated:

As shown in Reference 6.13, following potential accident scenarios have been considered:

1. A'spent fuel assembly drop in the spent fuel pool.

2~ Loss of spent fuel pool cooling system flow.

3. A seismic event.

4, A spent fuel cask drop.

5. A construction accident.

The probability of any of the first four accidents is not affected by the racks themselves; thus reracking cannot increase the probability of these accidents. As for the construction accident, FPL will ensure that no rack will be carried directly over the stored spent fuel assemblies. All work in the spent fuel pool area will be controlled and performed in strict accordance with specific written procedures developed for the rerack. The crane which will be used to "bring the racks into the fuel handling building has been evaluated and meets the requirements of Section 511 of NUREG612, "Control of Heavy Loads at Nuclear Power Plants".

In 'ddition, the temporary construction crane which will be used to move racks within the spent fuel pool area will meet the design, inspection, testing and operation requirements of Section 5.1.1 of NUREG-0612. This program provides for the safe ban/ling of heavy loads in the vicinity .of the spent fuel pool.

Accordingly, the modification will not involve a significant increase in the probability of an accident previously evaluated.

Pcs 016-186 PAGE 4 OF 12 The consequences of a spent fuel assembly drop in the spent fuel pool (Scenario 1) have been evaluated by the rack vendor and it was found that the criticality acceptance criterion, In keff less than or equal to 0.95, is not violated; addition, the radiological consequences of a fuel assembly drop are not dependent upon the installed racks and therefore are not changed from the present St Lucie 1 FSAR analysis. Because the calculated doses are less than 10 CFR Part 100 guidelines, the results of this accident are acceptable. This conclusion was confirmed by an independent NRC analysis. The results of the NRC analysis were reported in Section 15 of the Safety Evaluation of the St Lucie Plant Unit No. 1, Docket 50-335, 11/ 8/74. The results of the rack vendor analysis shows that a spent fuel assembly dropped on the racks will not distort the racks such that they would not perform their safety function.

Thus, the consequences of this type accident are not changed from the previously evaluated spent fuel assembly drops which have been found acceptable by the NRC.

The consequences of a loss of spent fuel pool cooling system flow (Scenario 2) have been evaluated by the rack vendor and it was found that sufficient time is available to provide an alternate means for cooling (i.e., the fire hose stations) in the event of a failure in the cooling system. The results of the analysis are found in Reference 6.13. Thus, the consequences of this type accident will not be significantly increased from -previously evaluated loss of cooling system flow accidents.

As discussed in Reference 6.13, the consequences of a seismic event (Scenario 3) have been evaluated and are acceptable. The new racks will be designed and fabricated to meet the requirements of applicable portions of the NRC Regulatory Guides and published standards referenced in Section 6.0 of this Engineering Package. The new racks are designed so that the floor loading from racks completely or partially filled with spent fuel assemblies (with or without CEAs inserted), or empty at the time of the incident, do not erceed the structural capability of the spent fuel pool. The fuel handling building and spent fuel pool structure have been evaluated for increased loads from the spent fuel racks in accordance with the criteria previously evaluated by the NRC, as discussed in Reference 6.13 Thus, the consequences of a seismic event are not significantly increased from previously evaluated events.

The consequences of a spent fuel cask drop (Scenario 4) have been evaluated and the results provided in Reference 6.13. The radiological consequences of the cask drop are well within the guidelines of 10CFR100 and the doses are not being increased to the doses analyzed for the present1y installed as'ompared racks. The cask drop analysis is ba'sed on Administrative and Technical Specification controls which ensure that minimum requirements for decay of irradiated fuel assemblies in the entire spent fuel pool are met prior to movement of the cask into the cask area of the spent fuel pool. Analyses also demonstrate that keff will always be less than the NRC acceptance criterion. In addition, the cask will not penetrate the spent. fuel pool as demonstrated by the'resent St Lucie 1 FSAR accident analyses. Thus, the consequences of a cask'rop accident will not be significantly increased from the previously evaluated accident analysis.

0 L C~l'l v J.G-L66 PAGE 5 OF 12 We consequences of a construction accident (Scenario 5) are enveloped by the spent fuel cask drop analysis previously performed in that the cask is heavier and provides a smaller crossmection than the new racks. In addition, all movements of heavy loads handled during the rerack operation will comply with the NRC guidelines presented in NUREG&612, "Control of Heavy Loads at Nuclear Power Plants". thus, the consequences of a construction accident are:not significantly increased from the previously evaluated accident analysis.

Iherefore, it isinvolve' concluded that the reracking of the spent fuel pool will not significant increase in the probability or consequences of an accident previously evaluated.

Create the possibility of a new or different kind of accident from any evaluated previously:

1he reracking has been evaluated in accordance with the guidance of the NRC position paper entitled, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", appropriate NRC Regulatory Guides, appropriate NRC Standard Review Plans, and appropriate industry codes and standards. In addition, a review of several previous NRC Safety Evaluation Reports for rerack applications similar to this EP has been conducted. As a result of this evaluation and these reviews, the modification does not, in any way, create the possibility of a new or different kind of accident from any accident previously evaluated in the Final Updated Safety Analysis for the St Lucie Unit 1 storage facilities.

iii) Involve a significant reduction in a margin of safety:

The NRC Staff Safety Evaluation Review process has established that the issue of margin of safety, when applied to a reracking modification, should address the'ollowing areas:

1. Nuclear criticality considerations.

2. Thermal-hydraulic considerations.

3. Mechanical, material and structural considerations-The established acceptance criterion for criticality's that the neutron multiplication factor in spent fuel pools shall be less than or equal to 0.95, including all uncertainties, under all conditions. This margin of safety has been adhered to in the criticality analysis methods for the new rack design.

The Reference 6.13 methods used in the criticality analysis conform with the applicable portions of the appropriate NRC and industry codes, standards, and specifications. In meeting the acceptance criteria for criticality in the spent fuel pool, such that k ff is always less than 0.95, including uncertainties at a 95K/95X probability confidence level, the reracking of the spent fuel pool does not involve a significant reduction in the margin of safety for nuclear criticality.

'll lJll 41$kv I',

J LPv PAGE 6 OF 12 As described in References 6.13 and 6.17, conservative methods are used to calculate the maximum fuel temperature and the increase in temperature of the water in the spent fuel pool.

The thermal-hydraulic evaluation uses the methods used for evaluations of the present spent fuel racks in demonstrating the temperature margins of safety are maintained for the conditions of increased fuel storage considering the modified sparger configuration. The modification will-allow an incr'ease to the heat load in the spent fuel pool. The evaluation shows that the exi,sting spent fuel cooling system will maintain the pool temperature at or below 150.8oF. Thus a margin of safety exists such that the'aximum allowable temperature of 217oF is not exceeded for the calculated increase in pool heat load. Thus, there is no significant reduction in the margin of safety for thermal- hydraulic or spent fuel cooling concerns.

The main safety function of the spent fuel pool and the racks is to maintain the spent fuel assemblies in a safe configur-ation through all normal or abnormal loadings, such as an earthquake, impact due to a spent fuel cask drop, drop of a spent fuel assembly, or drop of any other heavy object. As described in Reference 6.13, the mechanical, material and structural design of the modification is in accordance with applicable portions of the "NRC Position for Review and Acceptance of. Spent Fuel Storage and Handling Applica'tions",

dated April 14, 1978, as modified January 18, 1979; Standard Review Plan 3.8.4 and other applicable NRC and industry codes.

The rack materials used are compatible with the spent fuel pool and the spent fuel assemblies. The structural considerations of the new racks address margins of safety against tilting and deflection or movement, such that the racks are not damaged during impact. In addition the spent fuel assemblies remain intact and no criticality concerns exists. Thus, the margins of safety are not significantly reduced by the modification.

In summation, it has been shown that the modification has been analysed and the results of that analysis provided to the NRC in Reference 6.13. In addition, in accordance with 10CFR50.92, the implementation of this Engineering Package does not involve a significant hazards consideration because it does not:

~

l. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. create the possiblity of a new or different kind of accident from any accident previously evaluated; or

3. involve a significant reduction in a margin of safety.

PCH 018-186 PAGE 7 OF 12 3.2 Technical/S ecification Changes The implementation of this Engineering Package requires NRC approval of the following St Lucie 1 Technical Specification changes as described below and as shown on Attachment 7.3 (markedmp Technical Specification pages)

l. Specification 316.9.16 Bases is revised to reflect the assumptions used in calculations of the doses based on the decay time.

2. Specification 5.6.1.a.l is revised to correspond to the Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors (NUREG-0212 Rev 2) ~

3. Specification 5.6.1.a 2 is revised to show the nominal center-tomenter distance for the new storage racks.

4. Specification 5.6.1 a.3 is edited to discuss the boron concentration only.

5. Specification 5.6.1.a.4 is created to indicate the presence of Boraflex neutron absorber material in the cells

6. Specification 5.6.1.b and accompanying Figure 5.6-1 is created to show the increased spent fuel enrichment permitted in the pools

7. Sp'ecification 5.6.1.c is editorially changed from "b", to "c."

8. Specification 5.6.3 is changed to show the capacity of the new spent fuel storage racks.

Reference 6.13 provides a proposed amendment to the Facility Operating License regarding this modification and has been submitted to the NRC. Prior Commission approval is required for the implementation of this Engineering Package.

3.3 Attachment 7.6 provides a supplementary safety analysis for a change that has been introduced in.paragraph 9.3.1 (Special Instructions) ~

3.4 The following is a supplementary safety analysis for a change that has been introduced in Paragraph 9.5 (Special InstrucCiohs):

3.4.1 The safety 'valuation covers the reinstallation of the temporary construction crane following the initial placement of fuel in rack modules Fl', Gl and G2. It will show that no unanalyzed failure modes exist and that the potential consequences of any accidents are bounded by the Cask drop'ccident.

~ ~

3.4.2 With respect to Title 10 of the Code of Federal Regulations, -Tart 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.

PCM 018-186 PAGE 8 OF 12 (i) The probability of occurrence or the consequences of ~n accident or malfunction of equipment important to safety previously evaluated is not increased by the reinstallation of the temporary construction crane. The following are the bases for this conclusion:

The following potential accident scenarios associated with the spent fuel pool area have been previously evaluated:

Spent fuel assembly drop (FSAR Subsection 9.1;2.3)

2. Loss of spent fuel cooling system flow (FSAR Subsection 9.1.3.4)

3. Heavy load drop I

4. Seismic event (FSAR Subsection 3.7.5)

The probability of events 1, 2 and 4 is not related to the reinstallation of the temporary construction crane .or its presence in the spent fuel pool area; hence the reinstallation of the crane cannot increase the probability of occurrence of these accidents.

The probability of a heavy load drop in the spent fuel pool is not increased by the reinstallation of the temporary construction crane because the reinstallation procedure assures that no credible failure can occur which would result in the dropping of crane components into the spent fuel pool.

I The south "girder will be moved into its position at the south end of the bridge trucks using a'oller support at each a-end 4" which allows the girder to pass over th'e 'trucks with maximum clearance. The physical arrangement of the roller support assemblies is such as to prevent the cocking of the girder which would be required to cause the girder to fall into the pool. The minimal clearance 'between the bottom of the girder and the top of the trucks will allow the trucks to retain the support of the girder even if both roller assemblies fail. Once in. position, the girder will be secured the to both end concrete trucks. The positioning of the north girder over ledge at the north side of the pool will preclude any potential for dropping this girder into the pool.

Therefore,* the reinstallation of the temporary construction crane will not increase the probability of a heavy load drop.

I 'U'l VJO l,u PAGE 9 OF 12 The reinstallation of the temporary construction crane and its presence in the spent fuel pool area have no bearing on the consequences of the fuel assembly drop or the loss of fuel pool cooling; nor can they affect the consequences of the heavy load drop, since no heavy load other than the cxane components will be -brought into the fuel handling building until the"cxane has been removed.

\

A seismic event during the reassembly oT the crane cannot result in the dropping of crane components into the pool since at no time will any crane component be lifted over the pool.

The two bridge girdezs will be brought into the building, individually, over the 4 foot wide ledge at the north end of the pool. The south girder will be provided with a x'oiler support at each end which will be used to roll the girder into its position at the south end of the bridge trucks where will be secured. The north girder will be secured to the it trucks before the bridge assembly may be moved to'he south.

A postulated seismic event during any stage of this procedure cannot result in the dropping of any components into the pool. Thus, the consequences

. of a seismic event are unaffected by the reinstallation of the temporary crane.

Therefore, i,t is concluded that the reinstallation of the temporary construction crane will not involve an increase in

'he probability of occurrence or consequences of an accident previously evaluated.

(ii) The possibility for an accident or malfunction of a different type than any evaluated previously is not created by the reinstallation of the temporary construction cra'ne. The following are the bases for this conclusion:

The reinstallation of the temporary construction crane at a time after fuel has been placed in the rack modules Fl, Gl and G2 does introduce a modification to a condition previously evaluated in that there will now be fuel assemblies located in an area of the pool that was free of spent fuel during installation of the temporary crane. However, now, as the'nitial before, thexe will be no lifting of heavy loads over spent fuel or over any rack modules containing spent fuel; aU.

lifting of heav'y loads will comply with the guidelines of Section 5.1.1 of 'REG-0612, "Control oW Heavy Loads at .

Nuclear Power Plants." In addition, there is no potential for the interaction of the temporary construction crane with any safety related systems or equipment during its reinstallation in the spent fuel pool area, for the reasons stated in (i) above.

Therefore, it is concluded that the reinstallation of the temporary construction crane does not in any way create the possibility of an accident or malfunction of a different type than any previously evaluated-(iii) This modification does not change the margin of safety as defined in the basis for any technical specification. The following are the bases fox this conclusion:

ec>i olu-iu6 PAGE 10 OF 12 Technical Specifications 3.9.7, which requires that no load in excess of the weight of a fuel assembly be carried over spent fuel, has not been violated, since at no time'uring the reinstallation of the temporary construction crane will any load be carried over stored spent fuel.

Technical Specification 3.9.13, which limits the maximum load which may be handled by the spent fuel cask crane to 25 tons, has not been violated, since the heaviest temporary crane component to be handled by the cask crane weighs approximately 5400 lbs.

There are no other Technical Specifications applicable to this modification. Therefore, it is concluded that the modification does not change the margin of safety as defined in the basis for any 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 or a Technical Specification change.

3.4.3 Therefore, the reinstallation of the temporary construction crane following the initial placement of fuel in rack modules Fl, Gl and G2 will not have a detrimental effect on Plant Safety nor does it constitute an unreviewed safety question. No change to technical specifications is required. Thus, prior Commission approval is not required prior to implementation.

3.5 The following is a supplementary safety ana+sis for the installation of the indexing strip system (see Paragraphs 1.3.26, 2.3.26 and 8.2.8):

3.5.1 The safety evauation covers the installation of the indexing strip system. It will show that no unanalyzed failure modes exist and that the potential consequences of any accidents are bounded by the cask drop accident.

3.5.2 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.

PCM 018-186 PAGE 11 OF 12 The probability of occurrence or the consequences of - an accident or malfunction of equipment important to safety are not increased by the addition of the indexing strips, because neither the indexing strips nor the support for the strips will be carried over any racks containing spent fuel during their installation, nor do they have any interaction with the spent fuel racks or any other safety related system at any time. The indexing strips and their supports have been designed for all applicable loading conditions, including the safe. shutdown earthquake. Therefore, a failure of the indexing strips or their support system during a design basis event, which would result in the dropping of components into the spent fuel pool, is not possible. The spent fuel handling machine bridge girder, which provides support for the capt~est strips, has been evaluated for the additional loading 'mposed by the strips and the'ir supports and this loading has been found to be acceptable. In the unlikely event that the indexing strips should fall into the spent fuel pool during installation, . any accident scenario would be bounded by the existing cask drop accident analysis as discussed in St Lucie 1 FSAR Subsection 9.1.4.3.

The possibility for an accident or malfunction of a different type than any evaluated previously is not created. Because the weight of the heaviest indexing strip segment or the heaviest support assembly is insignificant when compared to a spent fuel cask, any accident involving a dropped indexing strip segment or support assembly would be bounded by the cask handling accident as discussed in St Lucie 1 FSAR Subsection 9.1.4.3.

The possibility of the misinstallation of the indexing strips, which could result in the incorrect identification of storage cells and the subsequent misplacement of a fuel assembly, has been considered. Although the design drawings show the proper location of the indexing strips, the criticality design has considered the misplacement of a freshly off-loaded fuel assembly in a Region 2 storage location. (See Paragraph 2.3.8 of this EP.)

The indexing strips and their supports have been designed for al1 applicable loading conditions, including the safe shutdown earthquake. Therefore, a failure of the indexing strips or their support system during a design basis event, which would result in the dropping of components into the spent fuel pool, is not possible.

(iii) The margin of safety as defined in the bases of the Technical Specifications is not reduced. Technical Specifications 3.9-7, which requires that no load with weight in excess of the weight of a fuel assembly be carried over spent fuel, is not violated, because no loads will be carried over spent fuel. There are no other Technical Specifications applicable to this modification.

PCM 018-186 PAGE 12 OF 12 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 Technical Specification change.

. Therefore, the installation of the indexing strip system will not have a detrimental effect on plant safety nor does it constitute an unreviewed safety question. No change to technical specifications is required. Thus, prior Commission approval for implementation is not required.

PCM 050-186 PAGE 1 OF 3 INSTRUMENT AIR UPGRADE This Engineering Package (EP) ls for the fnstallatlon of 2 new afr compressors, 2 new desiccant air dryers and removal of the exfstfng desiccant afr dryer, afterfilter package and refrigerant air dryer whfch do not have sufficient capacity to accomodate the new compressors. One of the two new afr compressors and one new air dryer will operate and the other will serve as a standby. We.exfsting compressors will remain as backup,.especially for loss of offsfte power, since only these compressors can be loaded on the diesel generator. The backup bottled air supply to the MSIUs and FCV9011 and FCV9021 will be retained.

This EP is classified as Non-Safety Related since the instrument afr (IA) system compressors and assocfated equipment performs no safety function. The safety evaluation has determined that this EP does not constitute an unrevfewed safety question and implementation of the EP does not require a change to the Plant Technical Specification Therefore, prior NRC notification for implementation of this EP is not required.

This EP has no impact on plant safety and operation.

The Supplement 1 provides revised design bases/analysis, safety evaluation, operation and maintenance guidelines and FSAR change package-

-Although the safety evaluatfon has been revised, the original results of evaluation as stated above remain unchanged.

Su lement 2 The Supplement 2 provides revised design for the piping and piping supports/restraints to install additional flexible hoses to minim'ize piping vfbration.

The original results of evaluation as stated in the safety evaluation remain unchanged.

Su lement 3 The Supplement 3 provides design for: addition of a strainer and associated valves for,the dryer prefflter drain valve, improved removability of after cooler automatic drain valve> replacement of thermal overload relays for compressor motors. It also provfdes revised design for individual control air lines to the compressors'nloaders and pressure switches to prevent compressor huntfng and replaces certain temperature gauges subject to vibration with fluid-filled gauges.

The original results of evaluation as stated fn the safety evaluation remains unchanged.

019 8L/

PCri 05v-186 PAGE 2 OF 3 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; (1) if the probability of occurrence or the consequences of an accident or malfunction of equipment 1mportant to safety previously evaluated in the safety analysis 'report may be increased; or (ii) if a poseib111ty for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) 1f the margin of safety as defined in the bases for any Technical Specification is reduced.

This EP is for the addition of two 100X capac1ty new compressors, two new desiccant air dryers and removal of the existing low capacity desiccant dryer, afterf1lter package, refrigerant air dryer and supplemental air bottle racks and associated piping.

Failure of the instrument a1r compressors and components resulting 1n loss of IA and consequent affects as stated in the FSAR Subsection 9.3.1.3 have been reviewed. Thi.s modification does not add any new failure modes for the safety related air operated valves. However, existing solenoid valves, without regulators, whose maximum operating pressure differential capacity is lees than 115 psi w111 be replaced via Design Equivalent Engineering Package (DEEP) No 154-188D. Malfunction, if any, of these soleno1d valves will lead to the Fail Safe mode of the process valves. The IA system design pressure and temperature downstream of the IA aftercooler remain unchanged, therefore there is no concern for the valve actuators. This modification is therefore classified as nonnuclear Safety ~lity Group D and non-class 1E.

The increase in IA requirements from 155 SCRM to 400 SCFM and pressure from 90-100 psig to 105-115 peig is based on FPL studies for the requirement of the IA.

The supplemental air bottle racks and associated. components are tied into the accumulators to maintain the MSIVs and feedwater FCVs air system pressure between 100"105 ps1g. The bottle racks and components were considered for removal because the new compressors (1C and 1D) will be able to provide adequate instrument air flow at the required pressure. However, they have been retained for backup air supply.

Based on the above description, the modification included in this EP is considered to be non-safety related. Ghie EP does not involve an unreviewed safety question, and following are the bases for this Justification:

~~ nrn

~ ~ nr PAGE 3 OF 3 (i) The probability of occurrence or the consequences of an accid"nt or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased. The instrument air system compressora and associated equipment are not used directly in any safety analysis for accidents or malfunction of equipment and as such are non-safety related and will have no effect on equipment vital to plant safety.

(ii) The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created. The components involved in this modification have no safety related function and 'no changes have been made to the normal operational design of the system with the compressors 1C and 1D in operation. In this mode the IA compressors 1A and 1B discharge valves V18109 and V18118 are closed to prevent IA leakage via these compressors.

Similarly, whenever the IA compressors 1A and 1B are required to operate, check valves V181195 and V181196 will prevent IA leakage via compressors 1C and 1D-(iii) The margin of safety as defined in the bases for any Technical Specification is not affected by this PCH, since the components involved in this modification are not included in the bases of any Technical Specification.

PCM 074-186 PAGE 1 OF 2 HEATER DRAIN PUMP DEMINERALIZED WATER SUPPLY,

~ST~RA:T This design package provides the required engineering for adding permanent piping from the demineralized water system to the Unit I heater drain pumps'echanical seals. The piping will make available to the seals the necessary backup flushing water meeting the appropriate chemistry requirements. The backup water source is required during initial plant startup whenever the pumps sit idle.

Based on the failure modes analysis and 10CFR50.59 review, this modification does not impact any safety related equipment and is not relied upon for any accident prevention or mitigation. Thus, it does not constitute an unreviewed safety question Implementation of this and is correctly classified as non-nuclear safety related.

modification, therefore, does not require prior HRC approval.

Su lement I This package revision provides valve drawings for valves added by this PC/H and modifies the expiration date to reflect the correct format. The scope of work specified by this Engineering Package has not been affected by this revision. The safety classification and the safety evaluation as stated is correct and is not impacted.

Su lement 2 This Supplement provides revisions to certain drawings to correct improperly identified generic valve tag numbers. Ho other changes are addressed. The safety classification does not change, the safety evaluation remains valid and the Technical Specifications are not affected as a result of this revision.

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I PObi U>4-18o PAGE 2 OF 2 SAFETY EVALUATION I

The Unit Heater Drain Pumps are located in a Non-Nuclear Safety Related system and as such are not required to function during any existing analyzed accident scenario. Therefore, modifications to these pumps affect only Non-Nuclear Safety Related, Quality Group D equipment.

Based on the failure mode analysis, failure of the demineralized water supply piping could result only in failure of the heater drain pumps. Since the piping and components are located remote from any safety related equipment or components, failure of this equipment will not inhibit operation of any safety related equipment or components.

Based on the above evaluation and information supplied in the design analysis it can be demonstrated that an unreviewed safety question as defined in 10CFR50.59 does not exist.

o 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.

Since this design change does not alter or affect equipment, used to mitigate accidents, the probability of occurrence of analyzed accidents remains unchanged.

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

There is no new failure mode introduced by this change that has not been evaluated previously in the FSAR.

o The margin of safety as defined in the basis for any Technical Specifications has not been reduced.

This change has no affect on any existing Technical Specifications.

PCM 149-186 PAGE 1 OF 2 PRESSURIZER MANWAY COVER STUD TENSIONER ATTACHMENT HOLES ABSTRACT This Engineering Package covers modifications to the pressurizer manway cover to provide a inethod of safely lifting the cover whenever it is removed froid the pressurizer. The original manway cover lifting lugs were previously removed since they interfered with use of the stud tensioning device. These modifications will provide a method for attaching the stud tensioning device directly to the nianway cover so the cover and tensioner can be removed as an integral assembly.

As described in Chapter 5 of the FSAR, the pressurizer and its manway cover are considered nuclear safety related. Therefore modifications to the pressurizer manway cover are considered nuclear safety related. Based on the 10 CFR 50.59 review as provided in the safety evaluation, this'modification does not involve un unreviewed safety question, nor does it require changes to the Tcchnical Specifications. Implementation of these modifications can thus be completed without prior NRC approvaL 019 8L/

pcM l&9-186 PAGE 2 OF 2 SAFETY EVALUATION This modification affects the Unit 1 pressurizer manway cover which is a pressure boundary component for the pressurizer and therefore is required to maintain Reactor Coolant System integrity. As described in Section 5.2 of the FSAR, the pressurizer assembly (including manway cover) is considered to be nuclear safety related Quality Group A, and was designed in accordance with AShlE Section III Class 1 criteria.

These retaining capabilities 'f modifications have been found to have no affect on the pressure the manway cover or the overall seismic qualification of the. pressurizer assembly. As indicated in the Design Analysis, the proposed modification has no affect on the original design bases for thc Reactor Coolant System as addressed in the FSAR. This design therefore has no affect on the functional capabilities of the pressurizer.

Based on the above evaluation and information supplied in the design analysis it can be demonstrated that an unreviewed safety question as defined by 10 CFR 50.59 does not exist.

o 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.

Since the design change does not affect the pressure retaining capabilities of the manway cover, the probability of occurrence of analyzed accidents remains unaffected.

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

This modification adds no new equipment. Therefore,'no ncw accidents have been created.

o The margin of safety as defined in the basis for any Technical Specifications has not been reduced.

This modification has no affect on any existing Technical Specifications.

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 Specifications is not required. 4s shown in the preceeding sections, the proposed change does not involve an unreviewed safety question since each concern posed by'0 CFR 50.59 questions concerning safety were answered positively. In addition no Technical Specifications have been affected.

In conclusion this design is acceptable with respect to nuclear safety, does not involve an unreviewed safety question and does not affect any Technical Specification. NRC approval of this design is not required prior to implementation.

PCM 152-186 PAGE 1 OF 3 DIESEL GENERATOR OIL LEVEL ALARM ABSTRACT The St Lucie iJnii No 1 Diesel Generators are furnished with an externally mounted level switch which provides a local alarm in the event of low oil level in the engine crankcase. The switch is physically connected by way of instrument tubing between the crankcase drain and the crankcase vent. The switch operates on the principle of a pivotal float and is attached to the side of the engine.

The alarm setpoint, therefore, is fixed by the installed location of the switch. above the crankcase drain.

In order to reduce the frequency of unwarranted crankcase oil level alarms and to provide visual determination of the actual crankcase oil level, the presently installed manufacturer's level switches will be replaced by level indicating switches, with ad5ustable setpoints.

The new switches will be mounted adjacent to the diesel generators. which locates them away from the mgintenance platforms. Since the new oil level indicating switches consists of engine oil tubing runs and have interaction with Safety Related equipment (Diesel Generators), this Engineering Package (EP) is classified as Safety Related.

A review of the changes to be implemented by this EP was performed against requirements of 10CFR50.59. As indicated in Section 3.0 of this EP, this EP does not involve an unreviewed safety question, nor does it require a revision to the technical specification; therefore, prior Commission approval is not required for implementation of this PC/M.

Supplement 1 This supplement revised Section 11.0, Drawings and Vendor Manuals, to reflect omitted drawing numbers. The original safety evaluation is not affected by this supplement.

Su lement 2 This Supplement revises the Abstract page and incorporates Nutherm International seismic test report, FPL 2382R. The Safety Evaluation indicated in Section 3.0 for this EP has been revised to include the seismic report and removal of HOLD points.

A review of the changes to be implemented by this EP was performed against requirements of 10CFR50.59. As indicated in Section 3.0 of this EP, this supplement does not involve an unreviewed safety question, nor does it require a revision to the technical specification; therefore, prior Commission approval is not required for implementation of this PC/M.

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PCM 152-l86 PAGE 2 OF 3 SAFETY EVALUhTI(%

Mith repect 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 possiblilty for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report mey be created; or (iii) if the margin of safety as defined in the bases for any technical specification is reduced.

This Engineering Package replaces the existing pivotal float swftch with a differential pressure indicating switch and relocating the switch to an area away from the dfesel generator and maintenance platform. An electrical terminal box will be installed to re-use the existfng conduit. In addition, proper isolation and test valves are provided for mafntenance. This PC/M has been classified as Safety Related since'here is interaction wfth safety related equipment.

Based on the above,'his Engineering Package does not constitute an unreviewed safety question because:

i) The probability of occurrence or the consequences to safety of an accident previously or malfunction of equipment important evaluated in the Safety Analysis Report is not increased since the new switches perform the same function. Also the replacement oil level indicating switches will be procured with proper seismic qualifications end the supports and installation have been seismically evaluated. The new of oil level switches will provide significant reductfon in the number of low crankcase oil level alarms with local visual indication of crankcase oil level.

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

a This installation is in accordance with the Code of Federal Regulation 10 CFR 50.48 and no impact fs incurred by this fnstallation.

b The desfgn of instrumentation and tube track supports are fn accordance with the ST LUCIE UNIT 1 and 2 DESIGN CRITERIA MANUAL, Rev 1, Vol 2 Section 7, paragraph 7 '

~

and the new oil level switches have been seismically qualiffed as noted in Nutherm letter of transmittal, Attachment 7.3 end Seismic Analysis Report FPL-2382R>

EMDRAC 8770-U.834.

Thfe installation is fn accordance with the Code of Federal'egulation 10 CFR 50.49 and hes been determined to have no impact on the 10 CFR 50.49 Environmental

(}ualification criteria because the equipment is located in the Diesel Generator Building which is a mild environment.

This PC/M provides a replacement of the existing oil level switches with the increased enhancement of a visual indication.

PCM 152-186 PAGE 3 OF 3 iii) The margin of safety as defined in the bases for any Technical Specifications is not affected by this PC/M since the components involved,in this modification are not included in the bases of any Technical Specifications The implementation of this PC/M does not require a change to the Plant Technical Specifications.

The foregoing constitutes, per 10 CFR50.59(b), the vritten safety evaluation which provides the bases that this change does not involve an unrevieved safety question and prior Nuclear Regulatory Commission approval for the implementation of this P(X is not required'

~

PCM 154-186 PAGE 1 OF 2 CCW PUMPS SUCTION PRESSURE GAUGES ABSTRACT This Engineering Package (EP) covers the modifications and details required to permanently mount the following CCW pump su'ction gauges:

Pump A PI-14-27A Pump B PI-14-27B Pump C PI-14-27C t

The pressure gauges are used to monitor the pump suction pressure during normal operation and periodic tests of standby pumps.

The Component Cooling Water System is classified Safety Related. The new installation is part of the CCW system and as such, this EP has been classified Safety Related.

This modification will improve pump monitoring and at the same time will reduce the valve manipulation when taking pressure readings. With this new installation, all the valves in the line between the pipe connection and the gauge will be normally open.

A review of the changes to be implemented by this PC/M was performed in accordance with the requirements of 10CFR50.59. As indicated in the Safety Evaluation (Section 3.0), this PC/M does not constitute an unreviewed safety question, nor does it require a revision to the Plant Technical Specifications. This modification will have no effect on plant safety or operation. Prior Commission approval is not required for the implementation of this PC/M.

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.

This Engineering Package (EP) provides the design for the installation of permanent suction pressure indicating gauges for each of the Component Cooling Water Pumps. The pressure gauges have been seismically tested and qualified to IEEE 344-1975 and have been seismically mounted. This PC/M is classified as Safety Related since this system as discussed in FSAR Section 9.2 provides a heat sink for safety related components associated with reactor decay heat removal for safe shutdown or LOCA conditions and is designed to withstand design basis earthquake loads, tornado loads and/or maximum flood levels without loss of safety function.

Based on the above, this Engineering Package does not constitute an unreviewed safety question for 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 is not increased since:

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salotuxl.ai <vu ss Vent'got:.i

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1%/, 104 I. The Ashcroft gauges have been qualified for Seismic Category I service per Nutherm Report No. FPL-3191R, EMDRAC 8770-12121.

b. This installation is located inside a structure that has been designed to withstand the effects of natural phenomena (tornado wind and tornado missile impact).

c. The new gauges are seismically tested and qualified to IEEE 344-1975.

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

a. The new gauges are only utilized for surveillance testing and improper indication will not affect the system.

b. The design of instrumentation and supports are in accordance with the ST LUCIE UNIT 1 and 2 DESIGN CRITERIA MANUAL, Rev 1, Vol 2, Section 7, paragraph 7.0 and the new pressure gauges have been seismically qualified.

C~ A pressure boundary failure of this new installation during normal operation would not result in failure of the system because the loss of fluid through the 1/2" tubing line is less than the makeup capability provided for the system (makeup is provided through valve LCV-14-1).

d. The component cooling system is arranged into two redundant and independent essential supply systems, each with a pump and heat. exchanger and the capability to supply the minimum complement of safety related equipment required for safe shutdown or LOCA conditions assuming a sIngle failure (loss of one CCW loop during an accident).

e. Provisions for isolation of this installation from the rest of the s'stem are provided.

f. No attachments to seismically designed masonry block walls are made for this installation.

g. This installation is in accordance with the Code of Federal Regulation 10 CFR 50.48 and no impact is incurred by this installation.

h. This installation is in accordance with the Code of Federal Regulation 10 CFR 50.49 and has been determined to have no impact on the 10 CFR 50.49 Environmental +alification criteria because the equipment is located in the Component Cooling Equipment area, which is a mild environment.

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

The implementation of this 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 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.

PCM 001-187 PAGE 1 OP 2 I&E BULLETIN 85-03 MOV SWITCH SETTINGS

~BSTRACT NRC IE Bulletin 85-03 requires that operating nuclear plants develop and implement a program to ensure that switch settings on selected safety-related motor-operated valves (MOV's) are correctly selected, set and maintaned to accommodate the maximum differential pressures expected on these valves during all postulated events within the design basis. Item a) of. the bulletin requires that the design basis for those HOV's located in AFW and HPSI systems be reviewed to determine the maximum differential pressure expected during both opening and closing strokes for all postulated events. This effort was performed for St. Lucie Units 1 and 2 by Combustion Engineering as part of the CE Owner's Group (CEOG) Tasks 528 and 531.

The results of the Item a) were subsequently transmitted to the NRC via FPL letter L-86-204, dated Hay 14, 1986.

Item b) of Bulletin 85-03 requires that the licensee establish the correct MCV switch settings based on the previously determined maximum differential pressure. All switches, including torque switches, torque bypass switches, position limit, position indication, overloads, etc., shall be considered. This design package provides the overall switch setting guidelines for each HOV, in addition to the specific design information necessary to set both the open and close torque switches and meet the requirements of Bulletin 85-03.

Once the correct switch settings have been incorporated into the respective MOV, Item c) of IE Bulletin 85-03 requires that each MOV be stroke tested against the maximum differential pressure established in Item a) to verify operability.

Because all of the HOV's associated with Bulletin 85-03 are safety-related, this engineering package has been classified as nuclear safety-reiated. A review of the switch setting changes to be implemented by this PC/H was performed against the requirements of 10CFR50.59, and it was concluded that these modifications do not constitute an unreviewed safety question and do not require a change to the plant Technical Specifications.

Su lement 1 This supplement revises the torque switch settings for valve V-3654 to account for" actual field testing. This condition had been previously justified via Safety Evaluation JPE-M-87-038, Rev. 1. The engineering package safety classification and safety evaluation are unaffected.

Su lement Supplement 2 provides minimum and maximum thrust values for valve V-3654. The original safety evaluation remains valid and concludes that no unreviewed safety question exists as a result of this supplement. Implementation of this supplement does not require prior NRC approval.

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PCM 001-187 PAGE 2 OF 2 SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, the modification described in this engineering package does not constitute 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 in the FSAR is not increased. This engineering package only provides the necessary design information required to set MOV switch settings utilizing hlOVATS signature analysis techniques. The recommended switch settings are considered enhancements to the existing settings to further ensure valve operability. Also, FSAR'esign bases were reviewed to determine the maximum loading conditions on each MOV to ensure the switch settings were properly selected. Furthermore, Item c) of BuDetin 85<3 requires that each MOV be stroke tested under maximum differential pressure conditions to ensure valve operability.

ii) The possibility for an accident or malfunction of a different type, than any evaluated previously in the safety analysis report is not created. No hardware modifications are performed as part of this PC/M. The proposed MOV switch settings alter accident mitigating equipment to further enhance operability. However, malfunctions of these MOV's do not in themselves initiate an accident. Therefore, no new accidents have been created.

- Additionally, the specified modifications do not introduce any new failure modes for the equipment. Therefore, no different malfunctions of the equipment than those previously'nalyzed are introduced.

The margin of safety as defined in the basis for any Technical Specification has not been reduced. This modification does not impact the Technical Specification requirements for the associated equipment. Valve stroke times are not impacted. Therefore, the margin of safety controlled by the Technical Specifications is preserved.

In conclusion, the change proposed in this engineering package is acceptable from the standpoint of nuclear safety does not involve an unreviewed safety question and prior NRC approval for implementation is not required.

PCH 076-'187 PAGE 1 OF 3 ERDADS/SAS UPGRADE ABSTRACT This Engineering Package provides for modifications to the computer room in preparation for implementing an upgrade to the Emergency Response Data Acquisition and Display System, which is also known as the Safety Assessment System (ERDADS/SAS), under PCH 076-187 Supplement 1 and PCH 077-287. Included in this work are the connection of the computer room to the adjoining office, relocation of computer room and office doors, installation of a false floor in

,the office, upgrade of lighting and convenience outlets, and installation of conduit and cables for the computer control terminals, the data loading terminal CRT 412 console, and disk drives.

This Engineering Package is classified as quality related due to the cable and conduit which are being installed to support SAS quality related components.

Implementation of this PCH does not involve an unreviewed safety question or a change to the Plant Technical Specifications. Therefore, it can be implemented without prior Commission approval.

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

SUPPLBENT 1 In addition to modifying the computer room, this Engineering, Package provides for an upgrade to the ZRDADS/SAS hardware and software including the Safety Parameter Display System (SPDS), in the St Lucie Unit 1 control room, computer room and technical support center. It will improve'the performance and display capabilities of the existing system and will include new display CRTs and keyboards, new color hardcopiers, additional printers, a data 1oading terminal, additional memory and new internal computer switching and communications components.

This EP remains classified as quality related since the function of the ERDADS/SAS system, which is to assist the operators in evaluating-the safety status of the plant, has not changed. The original safety evaluation has not been affected. Therefore, implementation of this EP does not involve an unreviewed safety question or a change to the Plant Technical Specifications.

It may be implemented without prior Commission approval.

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

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PCYi 076-187 PAGE 2 o f 3 SUPPLEMENT 2 Supplement 2 to this Engineering Package modifies the design to replace the CRTs, video generators, and supporting components which were originally specified in Supplement 1 due to hardware compatibility problems. The overall design remains. the same.

This EP remains classified as quality related since the function of the ERDADS/SAS system, which is to assist the operators in evaluating the safety status of the plant, has not changed. The original safety evaluation has not been affected. Therefore, implementation of this EP does not involve an unreviewed safety question or a change to the Plant Technical Specifications.

It may be implemented without prior Commission approval.

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

SUPPLEMENT 3 Supplement 3 to this Engineering Package incorporates the Human Pactors engineering rev'iew on the Safety Parameter Display System (SPDS) and Non-SPDS portions of the Emergency Response Data Acquisition and Display System (ERDADS). The overall design issued by the previous supplements remains the same.

This EP remains classified as quality related since the function of the ERDADS/SAS system, which is to assist the operators in evaluating the safety status of the plant, has not changed. The original safety evaluation has not been affected. Therefore, implementation of this EP does not involve an unreviewed safety question or a change to the Plant Technical Specifications.

It may be implemented without prior Commission approval.

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

I e

PCM 076-187 PAGE 3 of 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 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 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 are not increased since the existing input isolation of the ERDADS/SAS equipment will not be modified and will maintain the same level of protection for safety-related equipment.

There is no possibility for an accident or malfunction of a different type than any previously evaluated since no new safety-related functions or interfaces with safety-related systems are created by this EP.

iii) This modification does not change the margin of safety as defined in the basis for any Technical Specification, since no equipment installed or modified by this EP affects any parameter referenced in the Technical Specifications.

This EP modifies equipment which is not nuclear safetymelated.

However, since the ERDADS/SAS system assists control room personnel in evaluating the safety status of the plant," this EP is classified as quality related.

The Human Factors Engineering evaluation of the SPDS portion of the ERDADS system found seventy-four (74) HEDs. All four (4) Priority 1 HEDs have been corrected. Therefore, the HEDs found through this Human Factors Engineering review do not affect plant safety.

This EP has no effect on cables or components necessary for safe shutdown of the plant. Changes to equipment and structures involving 10CFR50 Appendix "R" fire protection requirements and changes to equipment on the Essential Equipment List have been addressed. (See .1). Thus, the proposed design is in compliance with applicable requirements for fire protection.

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

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

t

PCM 006-188 PAGE 1 OP 2 RCP SEAL COOLER HX LEAK DET ABSTRACT This Engineering Package addresses the replacement of existing limit switches for Component Cooling Water (CCW) outlet valves HCV-14-11-Al, A2, Bl and B2 and minor wiring modifications to the valve control circuits. The replacement limit switches will modify valve position indication so that -the indicating lights will discriminate between two (2) conditions: valve fully closed and not fully closed. The wiring modification to the valve control circuits consists of rewiring existing time delay relays to introduce a 60 second, time delay. This time delay will aU.ow sufficient CCW flow through the RCP Seal Cooler Heat Exchangers to normalixe the temperature, thus, prohibiting the initial temperature differential from initiating inadvertant valve control lockout.

Component Cooling Water (CCW) to the RCP is classified as Non"Nuclear Safety Related and nonmeismic according to St Lucie Plant - Unit 1 (PSL-1) PSAR Section 9.2.2.3. Also, the valve position indication circuits are Non-Nuclear Safety Related. However, considering the consequences of reactor coolant leaks'ge from a tube leak into the component cooling system, which is a loss of function of a radioactive confinement system, this package is classified as Quality Related.

The safety evaluation of this package indicates that neither the replacement of the limit switches nor the valve control circuit wiring modifications constitute an unreviewed safety question, and do not require a change in the Plant Technical Specifications. Therefore, prior NRC notification for implementation of this EP is not required.

This EP has no impact on plant safety and operations.

SUPPLK2KNT 1 Supplement 1 to this PCM provides clarification of the nonmeismic mounting of the limit switches,'esolves the ambiguity of the safety classification and incorporates .the appropriate revision numbers to the attachment list.

Verification of the TEDB classification and the identification of procedure changes have also been provided with this supplement.

The original safety evaluation has not been affected/remains unchanged except for a minor clarific'ation of the safety classification.

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PCM 006-188 PAGE 2 OF 2 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) ff the probability of occurrence or the consequences of an accident or malfunction of equfpment fmportant to safety previously evaluated fn 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 margfn of safety as defined fn 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 consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report will not be increased by this modification. Electrical separation is mafntafned between safety related wiring and components. The modifications provided by thfs package have no impact on equipment important to safety and introduce no new failure modes. Therefore, this modification does not increase the probability of an accident or malfunctfon of equfpment important to safety.

(ii) The possibility for An accident or malfunction of a different type than any evaluated'reviously in the Safety Analysis Report will not be created by this modification. No new failure modes have been introduced as stated in section 2.1.8 of this EP.

(ffi) The margin of safety as defined in the bases for any technical specification is not reduced sfnce this modificatfon does not degrade the CCW system, and the CCW Seal Coolers do not'form the bases of any Technical Specification.

As described in FSAR section 9.2.2 the Component Cooling System is a closed loop cooling water system that utflimes demineralfxed water to cool various components. The modificatfons described in this PCM involve replacfng existing limit switches and rewiring the associated CCW outlet valve circuits. These changes do not interrupt the closed loop Component CooUng Water System and are to a Non-Nuclear Safety Related valve fndication function which discriminates between a fully closed and not fully closed valve position. The equipment involved with this modification is nonnuclear safety related. However, based on the consequences of a tube leak from the heat exchanger into the Component Cooling System resulting in a loss of function of a radioactive confinement system, this Engineering Package is classiffed as Quality Related.

PCM 025-188 PAGE 1 OF 2 SFP TEMP JIB CRANE ABSTRACT In order to me6t the future spent fuel storage needs of the St.

Lucie Plant, REA-SLN-85-186 has been issued requesting the engineering services necessary to rerack the Unit 1 spent fuel pool. PC/M 142-186 has been issued for related modifications to support this rerack. PC/M 018-186 has been issued for the actual reracking effort. This engineering package is being issued for the temporary installation of a two (2) ton jib crane on the north wall of the Unit 1 Fuel Handling Building. This 'ib crane will facilitate installation and removal of the rack handling crane in the spent fuel pool area. The jib crane will be removed upon completion of the rerack operation.

Note: This jib crane will not be used when any fuel is stored within its operating limits. While installing the jib crane, no fuel shall be located within the associated load path.

The temporary jib crane does not perform or affect any safety related function. However, this PC/M is classified Quality Related since there is a potential for 'the jib crane to interact with safety related items. Quality Related requirements are applied to this modification.

This PC/M does not constitute an unreviewed safety 'uestion. The implementation of this PC/M does not require a change to plant technical specifications. This modification does not affect or safety. Based on the above, implementation of this plant'perations PC/M does not require prior NRC approval.

Su lement 1 This supplement adds the prepared I

and independent verification signatures for the Electrical interface and documents the power feed for the crane.

This supplement retains the Quality related classification originally designated. The modifications provided by this supplement do not introduce an unreviewed safety question as defined in 10CFR50.59, nor do they alter any Technical Specifications. Based on the above, implementation of this PC/M does not require prior NRC approval.

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PCI1 025-F88 PAGE 2 OF 2 Safet hnal sis 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:

the if (i) the of an probability of occurrence or consequences to safety accident or malfunction of equipment important previously, evaluated in the safety analysis report may be increased; or (ii) if 'he 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.

The temporary jib crane does not perform or affect any safety related system or'unction. This PC/M is classified as quality related to assure Q.C. inspection of .the installation and independent verification of the involve design.

The modifications included in this PC/M do not any unreviewed safety questions because:

(i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased since this modification will have no effect on equipment required to shut down the plant and monitor the plant. in a safe shutdown condition. hdditionally, the jib has been seismically designed and qualified,to the requirements of NURBG 0612. Therefore, the probability of a load drop accident has not been increased.

(ii) There malfunction is of a no possibility for an accident or different type than any previously evaluated since the jib crane does not perform a safety to any function and no changes have been made operational design. Installation and operation of the jib crane is enveloped by the existing cask drop analysis.

(iii) This modification does not change the margin of safety as defined in the basis for any technical specification since there will be no fuel located within the operating limits of the jib crane.

hdditionally, the jib has been seismically designed and qualified to the requireme'nts of NURBG 0612. Therefore, the probability of a load drop accident has not The implementation of this PC/M does not been'ncreased.

require a change to plant technical specifications. ',

The foregoing constitutes, per 10 CFR 50 '9(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 PC/M is not required.

PCM 033-188

,PAGE 1 OF 2 INST CHGS FOR HUMAN FACTORS ABSTRACT This Engineering Package (EP) includes engineering and design necessary to implement instrumentation changes to resolve twelve (12) outstanding Human Engineering Discrepancy Reports against the St Lucie - Unit 1 (PSL-1) control panel. The control panels affected by this modification are the Post Accident Panel, the Hot Shutdown Control Panel, remote Hydrogen Analyzer Panels 1A & 1B and RTG Boards 101, 102, 105, and 106.

This package is classified as Nuclear Safety Related since it involves modifications to instrumentation which serve Nuclear Safety Related functions.

The modifications are of a type which do not affect the function, availability or capability of the Nuclear Safety Related instrumentation. The safety evaluation has determined that this EP does not constitute an unreviewed safety question and does not require a change in the Plant Technical Specifications. This PCM can be implemented without prior NRC approval.

This PCM has no impact on plant safety or operation.

SUPPLEMENT l This Engineering Package Revision covers the removal of spare indicating switch LIS-07-SPARE and the associated blank nameplate from Control Room Auxiliary Console Panel No 1. Unlike the modifications of Revision 0 to this package, no Human Engineering Discrepancy Report was written against this instrument. This discrepancy was identified as a result of the ongoing commitment to the Human Factors Review process.

The original Safety Evaluation has been revised. The Safety Evaluation still concludes, however, that this EP does not involve an unreviewed safety question, or a change to any of the, Technical Specifications. Therefore, prior commission approval is not required for'mplementation of the PCM. The intent of the original Safety Evaluation is not affected by this supplement.

This PCM supplement has no impact on plant safety or operations.

I PCM 033-188 PAGE 2 OF 2 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. Q) 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.

I 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.

(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-(iii) The margin of safety as defined in the bases of any technical specification is not reduced since the instrumentation changes provided by this package involve only the removal of unused devices, the rearrangement and installation of indicating lights, the relocation of a recorder and the re-orientation of key-lock switches. The Safety Related circuits which were modified have been analyzed, and it has been determined"that there is no effect on the purpose, function or operation of the control circuits.

this equipment that is identified as it is affects Since EP Nuclear Safety Related, considered Nuclear Safety Re1ated.

Due to the fact that neither Revision 0 nor,Revision 1 to this EP involves any fire protection systems, fire rated assemblies or systems associated with achieving and maintaiding 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 1 FSAR requirements for fire protection equipment.

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 nor does implementation of Nuclear Safety Related PCM 033-188 or Supplement 1 to the same require a change to Plant Technical Specifications. Therefore, prior Commission approval for the implementation of this PCM or Supplement 1 to the same is not required.

PCM 069-188 PAGE 1 OF 2 DOUBLE INSUL EXC BEARING PED ABSTRACT This Engineering Package covers the modifications to install a double insulated exciter bearing pedestal.

This Engineering Package will provide the engineering and design details required to replace the existing f9 'xciter bearing pedestal with a double insulated exciter bearing pedestal. The installation of the new pedestal will reduce the potential of a grounding incident on the f9 exciter bearing. The new bearing pedestal is insulated, not only from the exciter bearing but from the exciter base. This will protect the exciter bearing from becoming grounded if the pedestal is grounded.

This Engineering Package will also address and document the removal of the bearing metal thermocouple at the P9 exciter bearing.

The Turbine Generator, which the exciter bearing and its pedestal are part of, is classified as nonmafety related, but since there will be modifications made to the RTG Boards, this package is classified as Quality Related.

This EP does not constitute an unreviewed safety question since the modifications described above were reviewed in accordance with 10CFR50.59 and were determined to have no adverse impact on plant operations or safety 0 related equipment. Prior NRC approval for the implementation of this PCM is not required.

The

'I implementation of Technical Specifications.

this PCM does not require a change to 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 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.

This Engineering Package provides the engineering and design details required to implement the replacement of the existing d9 exciter bearing pedestal with a double insulated pedestal. The instaHation of the double insulated pedestal will reduce the potential of grounding the exciter bearing from the pedestal, as well as from the exciter base.

This Engineering Package also provides for the removal of the bearing metal thermocouple, TE-22-39. This thermocouple has contributed to bearing failures when the thermocouple wiring introduced a ground on the bearing pedestal. Although the bearing metal thermocouple provides desirable remote indication and recording capabilities, it is not absolutely necessary. The bearing oil drain thermocouple, TE-22"35, provides remote indication and the bearing oil drain thermometer TI-22-43A, provides local indication of the same temperatures as the bearing oil drain thermocouple.

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PCM 069-188 PAGE 2 OF 2 The replacement of the Turbine Generator f9 exciter bearing pedestal and the removal of the bearing metal thermocouple are in the Turbine Building, are not safety related and do not affect any safety related plant systems.

Subsection 3.5.3.2 of the FSAR addresses External Missiles, with subpart (b) addressing Turbine Missiles, specifically, missiles generated by discs.

the high pressure turbine rotor and the low pressure turbine There are no changes to the high pressure turbine rotor nor the the low pressure turbine discs. The modifications required to replace bearing pedestal are located at the exciter end. The consequences of turbine failure and the potential for damage to criticalhasplant not structures, systems, and components from the resulting missiles been increased by this modification.

The modifications to the RTG Board will involve the de-termination of the cable for the bearing metal thermocouple. Although this modification does not affect the safety related functions of the RTG Board, this Engineering Package has been classified as Quality Related.

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

(i) 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, since the replacement of the existing pedestal with a double insulated pedestal and the removal of the bearing metal thermocouple will enhance the operability of the equipment. The modificati'ons will reduce the potential of grounding the bearing, by insulating the bearing from the bearing pedestal as well as from the exciter base and by preventing the thermocouple wiring from introducing a ground on the bearing pedestal.

(ii) As a result of this modification, there is no possibility for an accident or malfunction of a different type than any previously evaluated. There are no additional missiles generated by the replacement of equipment on the turbine generator There is no introduction of any new failure mode for the equipment.

(iii) This modification does not reduce the margin of safety as defined in the bases for any Technical Specification. The proposed design ensures that the exciter bearing is protected from becoming grounded if the pedestal is grounded. Since the exciter bearing is part of the turbine generator, which is a non-safety piece of equipment, the margin of safety provided by the Technical Specifications is preserved.

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.

PCM 269-188 PAGE 1 OF 3 GUIDE TUBE PLUGGING DEVICE REMOVAL ABSTRACT This Design Package covers the removel of six (6) Guide Tube Plugging Devices (30 Guide tubes) at St. Lucie Unit 1. These plugs were inserted into core locations previously occupied by Part Length Control Element Assemblies (PLCEA'S) to simulate the original hydraulic characteristics of the reactor.

The basis for the previous removal of the PLCEA'S is documented in PC/M Package No. 398-78 (Reference 6.1). The removal of the plugging device will cause a slight increase in the core bypass flow and a slight decrease in the calculated DNBR, therefore this is a "Safety Related" issue.

The benefits for removal of the plugging devices are outlined'n RAA NO. 1024 and FPL memo No. RE/PSL 8704 (References 6.2 & 6.3). The benefits include manpower and critical path time savings, Person REM reduction, fewer core alterations and saving of wear and tear on refueling equipment. Arguments for removal of the plugging devices are that part length plugs are not in use on St Lucie 2 and similar plug assemblies were removed at .Calvert Cliffs Unit 2 in August 1985 (Reference 6.2) with favorable results.

An analysis summary by Advance Nuclear Fuels Corporation showing acceptability of the plug removal is included in this design package.

The Safety Evaluation addressed 'in Section 3.0 concluded that the change implemented by this document is acceptable from the standpoint of nuclear safety as it does not involve an unreviewed safety question and does not I

change the Technical Specification. The results of this design package show no adverse ef fects on plant safety or operation. To this end, prior NRC approval is not required to implement this change.

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PCH 269-188 PAGE 2 OF 3 SAFETY EVALUATION This Engineering Package covers the removal of the Guide Tube Plugging Devices. The Guide Tube Plugging Device Removal is considered Safety Related since satisfactory performance prevents or mitigates the consequences of accidents that could cause undue risk to the health and .safety of the public.

10CFR50.59 allows changes to a facility as described in the safety analysis report (SAR) without prior NRC approval unless the proposed change involves a change to the Technical Specifications or an unreviewed safety.

question.

A change to the Technical Specifications is not required since the operational parameters of the CEDM are not altered by this change and core bypass flow 'and MDNBR are not specified in the Technical Specifications.

The increase in Core Bypass Flow and reduction in MDNBR were found to be within acceptable limits as reported by Advanced Nuclear Fuels Corp. {Attachment I).

The GTPD Removal results in a negative 10CFR50.59 determination as shown in Attachment V because:

A) The change described does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. The change has no effect on the operability of the CEA systems.

PCM 269-188 PAGE 3 OF 3 The change generates no functional change, no deletion, no overriding, or no degradation of the system to perform its design function. The ability to achieve required scram times is not altered and is tested at each refueling in accordance with the Technical Specifications.

8) The change does not create the possibility for an accident or malfunction of a different type than evaluated previously in the safety analysis report.

The change of this engineering package generates .-

no functional change, no deletion, no overriding, or degradation of the ability of the CEA system to perform its design function. Furthermore, the change implemented by this engineering package will have no effect on the operability of systems which interface with the CEA's.

C) The change does not reduce the margin of safety as defined in the basis for any Technical Specification. The change has been reviewed and it was found that the change will not alter the operational parameters of the CEDMs (e.g., scram time), and therefore< does not impact any margin of safety.

The change proposed is acceptable from the standpoint of nuclear safety as it does not involve an unreviewed safety question and does not change the Technical Specifications. Therefore, prior NRC approval is not required to implement this change.

PCM 018-189 PAGE 1 OF 2 COND PUMP CURR BAL RELAY ABSTRACT This Engineering Package covers the modification -of current balance (negative sequence) relays for condensate pump motors. The negative sequence protection is provided for the pump motor to protect against an open phase condition in the fused transfer svitch. Tso (2) plant trips have been attributed to condensate pump trips which resulted from 'spurious actuation of the condensate pump negative sequence relay. The existing relay vill be rewired to eliminate the condensate pump trip function leaving only the alarm function. The alarm vi11 continue to be annunciated in the Main Control Room utilising the ezfsting Condensate Pump Motors dedicated windows.

The system being modified is nonmafety, therefore this EP is- classified as Non-Safety Related.

This EP was reviewed in accordance w1th 10CFR50.59 and determined not to constitute an unrevieved safety question nor require a change to the Plant Technical Specifications. Prior NRC'pproval for implementation of this PCM is not required. This PCM has no adverse impact on plant safety or operations.

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PCM 018-189 PAGE 2 OF 2 SAFETY EVALUATION respect to Title of the Code of Federal Regulations, Part

'ith 10 50.59> a proposed change shall be deemed to involve an unreviewed if safety questions (i) the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis reports 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 reports may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.

This Engineering Packege provides the engineering and design details required to rewire the negative sequence relay for the condensate pumps. The ezisting trip function of the relay wiU. be deleted, but the alarm function will be maintained, so that the condition wiU. be alarmed in the Main Control Room. This wiU. eliminate the spurious tripping of the plant caused by the negative sequence relay.

This EP has been classified as Non&afety Related as the system modified is nonmafety and tripping of the condensate pump will not cause a direct reactor trip.

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

(i) 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. 2>is modification will eliminate the trip function of the sub)ect relay and will prevent condensate pump trip due to the relay actuation, regardless actual current unbalance.

if the actuation is spurious or caused by Currently, relay actuation'ither spurious or actual results in the trip of the associated condensate pump which will lead to trip of the feedwater pumps on low suction pressure and result in a unit trip on low steam generator level. After this modification, trip of a condensate pump can be caused only by an actual current imbalance leading to motor failure.

(ii) As a result of this modification, there is no possibility for an accident or malfunction of a different type than any previously evaluated. This modification does not affect safety related equipment. There is no introduction of any new failure modes for the safety related equipment.

(iii) Tld.s modification does not reduce the margin'f safety as defined in the bases for any Technical Specification. The margin of safety provided by the Technical Specifications is not affected as the equipment modified does not form a basis of any Technical Specifications.

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.

PCM 139-189 PAGE 1 OF STM GEN TUBE PLUG REPL

~BSQA~C This PC/M documents Engineering review and concurrence for removal of Primary Water Stress Corrosion Cracking susceptible Westinghouse mechanical tube plugs and replacement with Combustion Engineering expanded type plugs in the St. Lucie Unit 1 steam generators. The tube plugs manufactured from heat NX-3513 are to be removed.

The removal method will utilize electro-discharge machining if there is a high probability of cracked plugs. This PC/H also provides the information necessary to as-build affected documents.

Since the steam generator tubes are nuclear safety related, the tube plugs described herein and this Engineering Package 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.

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Pt.'M i&9-ii9 PAGE 2 OF 2

~AEU L This modification involves documenting the maintenance practice of removing existing plugs and replacement plugging of 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 removal of Westinghouse tube plugs and the use of the Combustion Engineering expanded tube plugs as a replacement design (previously utilized on the Unit I steam generators only as initial plugs), the required 50.59 revi'ew of the modification, and the information required for update of affected documents.'elded tube plugging is allowed for in the original steam generator design and will not require review and concurrence with the exception of weld procedures and qualification documents.

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

i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR 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 I components), and since

.no existing accident mitigation equipment or systems are altered by this modification.

ii) The possibility of an accident or malfunction of a different type than any previously evaluated in the SAR has not been created 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.

iii) 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 generators following this modification does not change.

Since the above arguments demonstrate than an unreviewed safety question does not exist, and since a revision to the Technical Specifications is not required, the replacement of Westinghouse tube plugs with the Combustion Engineering tube plugs in the Unit I steam generators does not require prior NRC approval.

PCM 239-984 PAGE 1 OF I SODIUM HYPOCHLORITE INJECTION PUMPS 1.0 OPERA'DON The operating parameters of the sodium hypochlorite injection system will be affected by this modification. 'Ibis will be achieved by altering the output of the injection pumps.

2.0 FUNCTION This modification will provide a lower discharge pressure from the sodium hypochlorite injection pumps. It will assure leak-tight shutoff of the diffuser control valves and extend, the operating life of pump and control valve components.

3.0 DESIGN DESCRIPTION The size of the injection pump impellers will be modified by this design change.

The existing impellers are to large for the present system conditions. By decreasing the impeller size, the same volumetric flow will be maintained at a lower discharge pressure.

SAFETY ANALYSIS 1.0 This modification has been reviewed with respect to 10CFR50.59 and has been deemed not to involve any unreviewed safety question because of the following:

1.1 The Sodium Hypochlorite System is non-safety related, nonseismic, and does not perform any function related to plant safety.

1.2 The new impellers will be the existing ones modified to the specified size, therefore, materials willsatisfy the original intent of the design.

1.3 There is no safety related equipment in the location of the Sodium Hypochlorite System. 'Iherefore, failure of the pumps, with the modified impeOers, will not increase the probability of an accident or malfunction of equipment important to safety previously evaluated.

1.4 The proposed modification does not require a change of the Technical Specifications.

2.0 The foregoing constitutes, per 10CFR50.59, the safety evaluation which provides the basis for not requiring prior Commission approval for the implementation of this modification.

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PCM 199-985 PAGE 1 OF 2 WATER TREATMENT PLANT REGENERATION WASTE NEUTRALIZATION TANK ABSTRACT The subject REA requested a neutralization tank be added to the Water Treatment Plant (WTP) to meet current Department of Environmental Regulation (DER) regulations governing discharge of hazardous wastes. The neutralization tank modification (PC/M 116-985) provides the necessary details for installation of this tank and the associated piping, equipment and components necessary to allow for regeneration wastes to be automaticaU.y directed to the tank during the appropriate times in the regeneration process. .During the caustic infection steps of regeneration, caustic solutions must be directed to the tank. The existing system, however, is unable to provide the necessary flows and pressures required to accommodate these regeneration steps due to the additional headloss in the new piping runs. Thus, to accommodate the new arrangement, a booster pump must be added to the caustic dilution water demineralized water supply. In addition, the caustic dilution water flow control valve and flow indicator/transmitter must be replaced to accommodate the flow requirements. This system is not required for plant safe shutdown, therefore, this modification is non-nuclear safety related and its implementation does not create an unreviewed safety question.

Su lement 1 The 'Supplement provides revisions to certain package drawings to correct improperly identified component, instrument and line designations. No other changes 'are addressed. The safety classification does not change, the safety evaluation remains valid and the Technical Specifications are not affected as a result of this revision.

Su lement 2 f

This Supplement provides a vendor drawing or the booster pump demineralized water supply flow control valve to accommodate as-building. This Supplement does not impact the original safety classification, Technical Specifications or safety evaluation whose results remain valid.

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pm 199-985 PAGE 2 OF 2 ShFEIT EVALUATION The subject modification provides for addition of a booster pump and flow control valve in the caustic dilution water supply to the WTP. In addition, the modification provides for replacement of certain caustic dilution water fiow transmitter components to accommodate the required flow rates. As defined in Section 9 of the Unit 1 PSAR, the WTP and its associated systems are classified as non-nuclear safety related and are not required to perform a safety function. Based on the failure mode analysis, as addressed in the Design Analysis, the modification has no affect on nuclear safety. Therefore, the modification is adequately classified as Non-Nuclear Safety Related Quality Group D.

Based on the above evaluation and information supplied in the design analysis, it can be demonstrated that an unreviewed safety question as defined by 10CFR 50.59 is not created. Since the modification affects only the WTP which is classified as Non-Nuclear Safety Related and cannot affect any other safety related equipment or components as addressed in the faQure mode analysis, the consequences of all analyzed accidents remains unchanged. Also, with respect to nuclear safety, no new accidents or mali'unctions are introduced as a result of this design change.

Additionally, the margin of safety as defined in the Technical Specifications has not been reduced.. Therefore, an unreviewed safety question does not, exist.

Since this modification does not involve an unreviewed safety question, nor require a change to the Technical Specifications, this modification is acceptable with respect to nuclear safety thus prior NRC approval is not required for implementation of the modification.

053-987 PCM PAGE l OF 2 CONDENSATE POLISHER CROSS-TIE INS ABSTRACT This Engineering Package (EP) is for the installation of the non outage portion of the 24 inch cross-tie piping required for the future connection of the Condensate Polisher System (CPS) to the St Lucie Unit 2 Condensate System. It-also includes the installation of the 8 inch cross-tie piping that will connect the CPS backwash pump suction to an existing Unit 2 Condensate Storage Tank non-safety class connection.

This EP is classified non-safety related since the portions of the Condensate System and Condensate Storage Tank piping where this modification will be implemented do not perform any safety function.

The safety evaluation has determined that this EP does not constitute an unreviewed safety question and implementation of the EP does not require a change to the Plant Technical Specif1cation. Therefore, prior NRC notification for implementing this EP is not required.

This EP has no impact on plant safety and operation.

SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulation, Part 50.59, a proposed'hange shall be deemed to involve an unreviewed safety question; (1) 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 1s reduced.

This Engineering Package (EP) is for the installation of the 24 inch condensate cross-tie piping from two (2) existing connections on the Unit 2 condensate system to the v1cinity of the Condensate Polisher Building.

l This is for the future connection of the Unit Condensate Polisher System (CPS) to the Unit 2 Condensate System. It also includes the installation of an 8 inch line from an existing non-safety class connection of the Unit 2 condensate tank to the vicinity of the Condensate Polisher Building. This piping will allow the use of the Unit 2 condensate for backwashing the Unit l condensate polishers. The portions of the Condensate System, Condensate Storage Tank piping and the CPS that this modification will be implementing do not perform any safety function or interact with safety related equipment, therefore this package is classified as non-nuclear safety related.

Based on the above description, the modification included in this Eng1neering Package (EP) is considered to be non-safety related. This EP does not involve an unreviewed safety question, and the following are bases for this justification:

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PCH 053-987 PAGE 2 OF 2 (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. The portions of the Condensate System, condensate storage tank piping and CPS where this modification will be implemented are not used in any safety analysis for accident mitigation. Malfunction of this equipment will have no effect on equipment vital to plant safety.

(ii) The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created. The components involved in this modification have no safety related function. Failure of these systems will not affect plant safety.

(iii) The margin of safety as defined in the bases for any Technical Specification is not affected by this PCH, since the components involved in this modification are not included in the bases of any Technical Specification.

The implementation of this PCH does not require a change to the plant Technical Specifications.

The foregoing constitutes, per 10CFR 50.59 (b), the written safety evaluation which provided the bases that this change does not involve an unreviewed safety question and prior Commisision approval for the implementation of this PCH is not required.