ML17308A479

Rept of Changes Made Under Provision of 10CFR50.59 for Period Ending 890122. W/890720 Ltr

ML17308A479
Person / Time
Site:	Saint Lucie
Issue date:	01/22/1989
From:	Woody C FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L-89-260, NUDOCS 8907250248
Download: ML17308A479 (171)

Text

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REGULATORY INFORMATION DISTRIBUTION SYSTEM (RZDS)

I SSION NBR 8907250248 DOC. DATE 89/01/22 NOTARIZED: NO DOCKET N XL:50-'335 .'St:.Lucie Plant,:Unit 1, Florida Power & Light Co. 05000335 AUTH. NAME AUTHOR AFFILIATION

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"St Lucie Plant Unit 1 Rept of Change's Made Under Provision of 10CFR50.59 for Period Ending 890122." W/890720 ltr. D DISTRIBUTION CODE: XE47D COPIES RECEIVED:LTR ENCL SIZE-

<' TITLE: 50.59 Annual Report of Changes, Tests or Experiments Made W/out Approv, h ~ w RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR E CL PD2-2 LA 1 0 PD2-2 PD D

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.. Box14000, Juno Beach, FL 33408.0420 JULY 2 0 lS89 L-89-260 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 of plant changes/modifications (PCM) which were made under the provisions of 10 CFR 50.59. Included with the brief description of each PCM is a summary of the safety evaluation. This report includes PCMs completed between January 23, 1988 and January 22, 1989 and correlates with the information included in Revision 8 of the Updated Final Safety Analysis Report.

Very truly yours, C. O.

Actin enior Vice President Nuclear COW/EJW/gp Enclosure cc: Stewart D. Ebneter, Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, St. Lucie Plant

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OSOOO335 PDC an FPL Group company

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St Lucie P1ant Unit 1 Report of Changes Made Under the Provisions of 10CFR 50.59 for the period ending January 22, 1989 8907250248

PLANT CHAN6E/NOD RFVIEMED FOR PSL1 FSAR AHENDHENT 8 NUtlBER REVISION TITLE 583-879 e-1 ESFAS POMER SUPPLY 875-882 DIESEL 6ENERATOR HI6H CAPACITY TURBOCHAR6ER INSTALLATION 297-177 8-3 REACTOR CAVITY FILTRATION SYSTEH 889-182 REACTOR VFSSEL HEAD SHIELDIN6 114-182 8-5 TURBINE SUPERVISORY INSTRUMENTATION 355-183 THERHAL SHIELD REMOVAL-PHASE III 177-184 8-3 ROSENOUNT TEtlPERATURE TRANSMITTER REPLACEHENT 839-185 ED6 SUBSYSTEH FLOM DIA6RAtlS eee-18s REFUELIN6 C FUEL XFR MACHINE 892-185 ESFAS POMER SUPPLY

-185 . REPLACEMENT OF VALVE SOLENOIDS 2e4-18s ICM BACKUP LUBEMATER BACKFLOM PREVENTER REPLACEMENT 828-186 CCM PUNP JOURNAL BEARIN6 NATL CH6 858-186 8-1 INSTRUMENT AIR UP6RADE 852-18e e-1 UPPER 6UIDE STRUCTURE LIFT RI6 REPAIR 874-186 HEATER DRAIN Pvtlp DEHINERALIZED MATER SUPPLY 877-186 18CFR58.49 EQ LIST REVISION 899-186 U6S LIFT RI6 LOAD TEST FIXTURE 114-186 CONDENSATE PUMPS EXPANSION JOINT REPLACENENT 119-186 18 CFR 58.49 EQ LIST REVISION 128-186 S.U. TRANSF LOCKOUT DISC SMITCH 131-186 AUTO LEAK RATE TESTER FOR PERSONNEL AIR LOCKS33-186 8-1, QSPDS SOFTMARE NODIFICATION

-186 BECKHAN MASTE 6AS SYSTEM OXY6EN ANALYZER REPLACEllENT 148-186 1,2 ANNUNCIATOR NUISANCE ALARMS

PLANT CHAN6E/NOD REVIEMED FOR PSL1 FSAR AHENDHENT 8 .

NUNBER REVISION TITLE 147-186 IC'M DISCH PIPE ZINC RIBBON 813-187 SIHULATOR TRAININ6 FACILITY 6AI-TRONICS 816-187 CCM. TCMi OBCM VALVE ACTVATOR REPLACEHENT 818-187 DRAIN FOR PIPE LINE R-MH-848 834-187 CONDENSER OUTLET TUBE SHEET AND MATERBOX COATIN6S 836-187 CONDENSER TUBIN6 STRAIN 6AUBE INSTALLATION 839-187 CONDENSER RECIRC TO COND PNEUHATIC SQRT EXTR REPLACENENT 841-187 HAIN FEEDMATER RE6 VALVE POS IND REHOVAL 854-187 CONDENSATE POLISHER TIE-INS 875-187 FIRE DETECTOR HODIFICATIONS

-187 8-2 - ERDADS/SAS UP6RADE 878-187 REPL OF F 8 P CONTROLLERS 885-187 TURBINE 6ENERATOR SEAL OIL SYSTEH ENHANCEHENT 888-187 8-1 REMOTE REACTOR VESSEL LEVEL INDICATION 185-187 CHAR6IN6 PUNP BLOCK NATL CH6 116-187 - REPLCHNT OF S.R. BATT 1A418 119-187 6ROUTIN6 OF NASONRY BLOCK MALLS 123-187 CEA N6 SETS LOCK-OUT RELAY 128-187 8-1 SI TANK 8 CONT FAN COOLER INST UP6RADE 141-187 488V PCB TRANSFORHER REPLACEHENT 142-187 488V LOAD CNTR 1A3 4 183 TRANSFORHER REPLACEHENT 143-187 488V PCB TRANSFORHER REPLACEHENT 2-187 SIT SAMPLE VALVE AS BUILD NODIF ICATION

-187 CEDS COIL PMR PR06 PART LEN6TH REHOVAL 881-188 HOISTURE SEPARATOR REHEATER SHELL REPAIR

PLANT CHANGE/NOD REVIEMED FOR PSL1 FSAR AllENDtlENT 8 NUtlBER REVISION TITLE 883-188D CONDENSER EXPANSION JOINT IllPINGEllENT PLATE llODIFICATION r" ee5-1ss 8-1 HETRASCOPE REPLACEHENT has'CP 886-188 COOLER HEAT EXCHANGER TUBE LEAK DETECTION 887-188 8-2 RCP VIB HONIT EQUIP UP6RADE ee9-Iss EQ DOC PACK 4 DISCONN tlOV SPACE HTRS 818-188 STATION AIR/INST AIR PRESS IND RPLCHNT 811-188 RAB/RCB 'MALKMAY 812-188 1B 4 1D INSTRUHENT INVERTER DRAMING CHAN6ES 813-188 8-1 LIGHTIN6 PANEL RELAY 815-188 ICM LUBE MATER PIPE RESTRAINT HODIFICATION

-188 'ONDENSATE PUllp DISCHAR6E SAtlPLING LINES e19-1ss TURBINE LUBE OIL SYS/RESERVOIR PERNANENT FLUSH CONNECTIONS

'28-188 LP 122 CKT EXCHANGE 821-188 DG BUILDIN6 DELUGE VALVES CLAPPER LATCH ASSEMBLY REPLACEMENT 822-188 MI RE DELETION FROM SMI 7 CH S S-2/388 826-188 FLOOD PROTECTION STOP L06 419 829-188 B.A. CONC RED DOC PAC UPDATE 831-188. e-1 RCB PROTECT IVE COAT IN6S llAINTENANCE 833-188 INSTRUMENT CHAN6ES FOR HUllAN FACTORS CONCERNS 835-188 EXTRACTION STll PIPIN6 NATL UP6RADE 838-188 FDMTR RE6 SYS CONTROLLER REPLACEllENT 843-les EQ LIST REV- SPARE PARTS 6-188 ED6 DMG 8 INSTR LIST CORRECTIONS

-1880 REACTOR HEAD 0-RIN6 RETAININ6 RING llODIFICATION 855-188D ICM & CM PUllP PACKA6E REPLACEllENT

PLANT CHAN6E/tlOD REVIEMED FOR PSL1 FSAR AtlENDtlENT 8 NUtlBER REVISION TITLE e59-1ss CONDENSER INLET TUBE SHEET AND MATERBOX COATIN6 ese-188 DIESEL 6ENERATOR 60VERNOR INSTABILITY 864-188 REACTOR CAVITY INFLATABLE SEAL 874-188 HAIN 6EN LINKS tlOD 875-188 PT INDICATION ENHANOENENT 876-188D REPLACEMENT OF PRESSURE IND PI-18-3 877-188D REPLACEtiENT OF FLOM TRANStlITTER FT-89-3B1 882-188 REPL BLDMN CONTROL VLV POSITIONERS 893-188D tlAIN 6EN SUR6E CAP REPLACEtlENT 894-188 BORIC ACID CONCENTRATION REDUCT 188 S/U TRANSFtlR 1A818 DIFF RELAY REPLCtlT 187-188 8 FUEL POOL PURIF SYS PUtlPS tlECH SEAL REPLACEtlENT 189-188 CONDENSATE RECOVERY SYSTEtl PUMPS tlECHANICAL SEAL REPLACEtlENT 111-188 TURBINE 6LAND SEAL SYSTEtl PUtlPS tlECHANLCAL SEAL REPLACEtlENT 113-188D SECONDARY SIDE MET LAYUP SYS PUtlPS tlECH SEAL REPLACEtlENT 114-188 DUAL CEA EXT SHAFT REPLACEtlENT 115-188D CONDENSER INLET MATER BOX DRAIN 117-188 REACTOR COOLANT PUtlp CASE TO COVER 6ASKET REPLACEtlENT 119-188 SAFETY INJ SYS BLANK FLAN6E REtlOVAL 122-188 ICM PRESSURE INDICATOR UP6RADE 125-188 CONTAINMENT FAN COOLER SHORT TERN RESTORATION 127-188 CONDENSATE POLISHER SYSTEtl PUtlPS .tlECHANICAL SEAL REPLACEtlENT

-188 STEAN 6EN TUBE PLUS DESI6N

-188 688V TAPIN6 PROC 143-188 FIRE PUtlp BKR O.L. TRIP DEV

PLANT CHAN6E/NOD REVIEWED FOR PSL1 FSAR AHENDHENT 8 NUHBER REVISION TITLE 158-1SS RCB EQUIP HATCH DOOR OPER DM6 159-1880 38 INCH STEAN 6ENERATOR NOZZLE DAN SEALS 161-1880 HAIN STEAN NOZZLE BLOCK DRAIN PIPE REPLACEHENT 162-188 TURBINE DRAIN VLV REPLACEHENT 165-188 ITT BARTON TRANSHITTER REPLACEHENT 167-188 HFMTR VLVS SPRIN6 RETAINER 178-188 IN-CORE INSTRUHENT THIHBLE FLANBE REPLACEHENT 171-188 IN-CORE INSTRUHENT THIMBLE FLAN6E REPLACEHENT 175-1880 HISC. SNUBBER HODIF I CAT IONS 177-188D HISC. SNUBBER UP6RADE

-188D . ICM LUBE MATER FLAN6E, REPLACEtlENT 179-188D AFM PIPE 8 RESTRAINT CORROSION 188-1880 FEEDMATER FLOM INST LIST RAN6E CORRECTION 181-1880 . FMRV TECH HANUAL UPDATE TO REFLECT SNUBBER INSTALL 183-188D RCP SEAL CARTRID6E 0-RIN6 PART NUHBER CHAN6E 188-188 D 8-1 SS/996 REPLACEHENT FOR 18 DIESEL 6ENERATOR 288-188D THROTTLED VALVE CMD LS DEV 285-188D ADD SETPOINT INFO OR RPS TO SETPOINT INDEX 218-1880 ICC VENDOR HANUAL UPDATE (1988) 217-188 688V CABLE SPLICES 228-1880 PACIFIC VLV 28" CHECK VLV HIN6E NATL CH6 221-188 D ZURN STRAINER HODEL 595A PART NUHBER CHAN6E

/42-1880 VELAN VALVE PARTS HATERIAL CHAN6ES HCV3615

-188D CHECK VALVE FEED FROH SHUTDOMN HTEXCH 1A 1B SEAL HAT CHN6 246-188D PORV LOWER SEAL BUSHIN6 6ASKFT V1482,,V1484

PLANT CHAN6E/NOD REVIEMED FOR PSL1 FSAR AHENDHENT 8 NUNBER REVISION TITLE 285-1880 BETA ANNUNCIATOR INST MANUAL 338-1880 FT-3321 CMD CHAN6E 335-1880 SAS POWER FEED CMD CHAN6E eea-985 8-2 UP6RADE OF NORTH MASTE MATER TREATHENT FACILITY 887-985 HYPOCHLORITE CELL FLUSH SYSTEH 171-985 HYPOCHLORITE SYSTEH INSTRUHENT ENHANCEMENT 199-985 MATER TRTHNT PLT RE6ENTN MSTE NEUTRLZTN TANK NOD-BOOSTER PHP 812-986 MPT 6ROUND EROSION REPAIRS 828-986 INTAKE CANAL DRED6IN6 AND SLOPE RESTORATION 839-986 BLOMDOMN BUILDIN6 RADIATION HONITORIN6 S Y ST EH

-986 e-1 . SINULATOR TRAININ6 FACILITY PIPIN6 TIE-INS 832-988 SECURITY BLD6S ENHANCEHENTS e9s-988 S6BTF MONITOR STORA6E TANKS-VENT STACK REPLACEHENT 186-988 S6 BLOMDOMN TREATHENT FACILITY SYSTEH PUHPS HECH SEAL RPLHT 137-988 6ENERATOR RETAININ6 RIN6S REPLACEHENT 1eo-9880 REPAIR OF FIRE PROT LINE 6"-FP-153 298-988 REHOVAL OF FORNS BL06 FROH SECURITY PERIHETER N/A RBfOVAL OF GUIDE TUBE PLUGGING DEVICES N/A r P RELOAD SAFETY ANALYSIS CYCLE

8' PCM 583-079 ESFAS POMER SUPPLY This change request is to cancel the replacement and relocation of a new ATl power supply (Consolidated Controls Corporation (CCC)) Device No. PS105) proposed under the original PC/M.

The original ATl power supply (CCC Part No. KDD1907 (Lambda Part No. LXs-C-15) will be retained as in the MA ESFAS cabinet as a result of this change request. The proposed replacement ATi power supply (CCC Part No. LXS-D-15-R)) will not be utilized, as a result of this change request. The location of the new ATI power supply as proposed under the original PC/M will be cancelled. Therefore, the use of the mounting hardware for the new ATl proposed power supply will be cancelled under this change request.

The original ATI power supply will be relocated under approved PC/M 92-1 85.

PCM 583-079 PC/M 583-79 CHANGE REQUEST 1 This change request to the Engineered Safeguards Features Actuation System Cabinets (ESFAS) has no effect on nuclear safety and does not alter the fntent of the original safety analysfs for PC/M 583-19.

<<'"P The existing ATI power supply (Lambda Part No. LXS-C-15) will be maintained ancf its location will not change as a result of this change request. Performance of the existing ATI power supply has been reliable for at least ten years of operation.

The power supply is used for test circuit and does not perform a safety related function.

The power supply is located in a mild environment and therefore the requirements of 10CFR50.49 do not apply.

The ATI power supply will be relocated under approved PC/M 92-185 to improve ventilation which will result in lower operating temperatures and provide improved performance.

since the ESFAS Is not utilized in determining the probabilities of accidents.

~ this change request does not affect the availability, redundancy, capacity, or function of any equipment required to mitigate the effects of an accident.

This change request does not affect any other safety related equipment.

Redundancy, function, or failure mocfe capacity have not been changed, Y " Y failure could directly result In a no~ontrolled release of radloactlve material.

The possibility of equipment malfunction of a different type than analyzed ln the FSAR has not been increased.

The margin of safety as defined in the basis for any Technical Specification has not been changed since this change request does not change the performance, capabilities, or operating characteristics of the ESFAS.

ln conclusion this change request to PC/M 583-19 does not Involve an unrevlewed safety question.

PCM 075-082 DIESEL GENERATOR HIGH CAPACITY TURBOCHARGER INSTALLATION SYSTEM DESCRIPTION 1.0 Design Description The diesel turbochargers supply co@pressed air to boost the perfor-mance of the diesel engines. When the engines operating below 50X load, the turbochargers are gear driven from the crankshaft. The St. Lucie diesel generators have been experiencing numerous turbo-charger'ailures due to the gear train. Electro-motive diesel has produced a new high capacity turbocharger with a new, stronger gear train that extends time between scheduled overhauls up to 1500X. The new high capacity turbocharger is a commercial grade item manufactured by EMD of General Motors to the same high quality standards as the original assembly.

2.0 Function The diesel turbocharger supplies compressed air to the diesel engine to provide more power output per cubic inch of piston displacement.

3.0 Operation This modification will not affect the diesel generators'oad capabilities or operating characteristics and thus do not change the operating procedures.

FCN 075-082 SAFETY AHALYSIS This change does not involve an unreviewed safety question because:

l. a) The probability of occurrence of an accident previously evaluated in the FSAR has not been affected since the diesel generators are not utilized in determining the probabilities of accidents.

b) The consequences of an accident previously evaluated in the FSAR have not been changed since this modification does not affect the operability of any equipment required to mitigate the effects of an accident.

c) The Probability of malfunction of equipment important to safety previously evaluated in the PSAR has not been adversely affected since EHD states, "leigh capacity turbochargers have not exhibited any adverse effects from the heavier components even when used in locomotives which experience relatively high "g" forces during hard couplings and high speed crossovers. In our opinion, it is highly improbable that the slight turbo weight increase would sufficiently change resonant frequencies to create seismic sensitivity where it does not now exist. From a functional stand-point, we expect identical response from the 17.9:1 gear ratio high capacity turbo as we received from the 18:1 ratio standard turbo."

I'n addition, this modification does not affect any other safety related equipment.

d) The consequences of the malfunction of equipment important to safety previously evaluated in the-FSAR have not been affected for the same reasons give in 1 (c).

2~ a) The possibility of an accident of a different type than analyzed in the FSAR has not been created since this modification does not affect any systems vital to safety or whose failure could result in an uncontrolled release of radioactive material.

b) The possibility of equipment malfunction of a different than analyzed in the FSAR has not been increased for'ype the same reasons given in 1 (c).

3~ The margin of safety as defined in the basis for any Technical Specification has not been changed since this modification does not change the performance, load capabilities, or operating characteristics of the diesel generators.

0186L

PCM 297-177 REACTOR CAVITY FILTRATION SYSTEM During refueling operations, the refueling cavity is filled wi.th water. The water provides protection from radiation given off by individual fuel bundles in addition to cooling them off khan the reactor head is first x'emoved, radioactive impurities (crud) may ba released into the water As a result, refueling cavity water bccomcs turbid, making it difficult to observe removal and replacement of fuel assemblies below the watex'evel, In order to ensure water clarity in the refueling cavity during refueling operations, a reactor cavity filtration system is needed. At present, the purification portion of the fuel pool system performs this function during refueling; however, it may not have sufficient capacity for ensuring the water clarity that is needed.

The fuel pool purification system, located in the FHB maintains clarity and puri.ty of water in t:he fuel pool, refueling water tank and the refueling cavity.

The purification loop consists of the purification pump (150 gpm capacity), ion exchanger, filter, strainers and surface skimmers, Fuel pool water is circulated by the pump through a filter which removes particulates larger than 5 micron size and through an ion exchanger to zemove ionic material ~

During refueling operations,'his same system is used for purification of the refueling cavity. The 3 inch suction and discharge piping axe routed from the FHB, through the penetration room in the RAB and into the refueling cavity in-side the RCB This same system is also used for filling and draining the re-fueling cavity The existing suction anddischarge nnzzles inside the refueling cavity are 9 inches apart and are located at the far end of the refueling cavity away from the reactor'vessel. In order to provide better filtration, the "uction and discharge locations would require more separation in order that filte 'a8 water has a chance to disperse before being sucked back through the filtration system.

Also, the suction line would need to be extended to the vicinity of the reactor vessel where it <<an do the most good. It is at the reactor vessel head that water

'first becomes contaminated requiring removals It is also where we require visually clear water This contaminated water needs t:o be dxawn out before has a chance to drift into the farther reaches of. the refueling cavity.

it Since we must be able to drain t:he refueling cavity using the fuel pool puri-fication system, the existing suction location at the bottom of the refueling cavity must be maintained; However, wa also need to have suction at the vicinity of the react:or vessel head which exists at a higher elevation. Thezeforc, addi-tional suction line must taa off the existing suction pipe and ba routed t:o tha reactor vessel Valves would need to be provided so that this line could b valved out during draining operations In addition to routing additional .,uction piping to thc vicinity of tha reactor head, t:hc water clari.ty can bc improved by installing a scpaxate rcactox cnvi.ty filt'.ration system. This system would hnntllc purification of thc refueling cavity alone, rclicving t:hc fuel pool p<<rifi.cat:ion system in thc FHB from t:his additional duty. With a scpar..t:c rcactot, cnvt.t:y filtrntion syst:cm, wc would have full time purificat:ion of bltc rcfucling cnvi,ty as opposed .to intcrmittant: purificat:ion

<<ntlcr thc old system. Also, a scpnrat:c sy:;tom could bc installed having grcatcr capacity than t:hc prrscnt.- fuel pool purificat:ion system, Since the fua1 pool purification pump is required for filling and dzaining, the reactor cavity filtration system would tea off the existing suction and di.scharga lines and we would thus mai.ntain the capability of filtering the refueling cavity with the purification portion of the fuel pool syst:cm,

I PCM 297-177 This PC/H does not constitute an unreviewed safety questions or involve a tech. spec. change.

The reactor cavity filtration system will operate only during plant shutdown while refueling. The addition of this system should reduce the time required for removal and replacement of the fuel assemblies.

0198L/

PCM 089-182 REACTOR VESSEL HEAD SHIELDING ABSTRACT THIS REPORT DEMONSTRATES THE ACCEPTABILITY OF THE TEMPORARY R.V.

HEAD SHIELD AND ITS PERMANENTLY ATTACHED SUPPORT STRUCTURE, AS PROVIDED BY NUCLEAR POMER OUTFITTERS OF CRYSTAL LAKE, IL., FOR USE IN ST. LUCIE UNIT 1. IT ALSO FULFILLS THE REQUIREMENT FOR SUBMITTAL OF A "DESIGN PACKAGE" AS REQUIRED BY FLORIDA POWER AND LIGHT SPECIFICATION FOR ENGINEERING PROCURMENT, AND INSTALLATION FOR PC/M 89-82, DATED OCTOBER, 1982.

IY. SAFETY ANALYSIS The reactor vessel head shielding consists of lead wool blankets hung around the reactor head during plant outages. The blankets are hung from a support system which is permanently attached to the reactor vessel head liftrig.

The shielding system does not perform a nuclear safety related function. It has been designed to withstand all applicable loads specified in the original plant design. A NUREG 0612 analysis has been performed to include the additional loading of the shielding system on the reactor head lift rig. The analysis shows that all NUREG 0612 requirements are satisfied.

Apropriate QA and QC requirements have been identified as well as procedures to assure the installation and use will conform to the design criteria.

The probability of occurrence or the consequences of a design basis accident or malfunction of equipment important to the safety of the plant, previously evaluated in the FSAR, has not been increased. There is no possibility of accident or malfunction different than those previously evaluated. Also, there are no changes to the technical specification of the plant. Therefore, it can be concluded that the addition of the reactor vessel head shielding system does not pose an unreviewed safety question pursuant to 10CFR 50.59.

~ ' ~ l q y ~

PCM 114-182 ST LUCIE PLANT - UNIT 1 TURBINE SUPERVISORY INSTRUMENTATION INTRODUCTION The turbine is a Westinghouse tandem compound four-flow exhaust 1800 RPM unit vith one high pressure and tvo lov pressure elements. The AC generator and brushless"type exciter are directly connected to the turbine generator shaft. The unit is provided with throttle valve steam chest assemblies located on each side of the high pressure turbine casing. The structural shapes of the casings and their methods of support are carefully designed to obtain free but symmetrical movements resulting from thermal changes.

Turbine Supervisory Instrumentation (TSI) is desigaed to provide optimum insight into the mechanical integrity of the turbine generator.

This system utilizes a combination of monitoring, recording and logging to collect data on the operation of the turbine. The TSI system is used to sense subtle changes in the operation of the turbine generator.

The items listed below are considered very important in the control of safe starting, loading and monitoring of the turbine:

h. Radial Vibration and Vibration Phase Angle B. Rotor Eccentricity C. Differential Expansion D. Thrust Bearing Monitor E. Case Expansion P. Turbine Speed and Acceleration G. Instrumentation Racks and Cable Terminations H~ Mimic Display aad Annunciation Lights I. PRobes, Cables and Conduit Installation The Bently Nevada TSI system vill ultimately replace the existing Westinghouse Turbine Supervisory equipments However, the replacement vill be accomplished in two (2) stages. The first stage is designed to install the Bently Nevada system vithout removing the exisitng Westinghouse equipment. While the second stage vill be to disconnect the Westinghouse equipment and to complete the connection of the new system.

Included in the first etage of implementation are the installation of the brackets, the probes, the conduits, the electrical boxes, the TSI cabinet, and the annunciator mimic display for the Bently Nevada system.

The existing Westinghouse thrust bearing probe located at the coupling epacer betveea the jackshaft and the lov pressure (LP) turbine 1B is to be replaced with nev Bently Nevada probes. The mounting bracket for the Westinghouse probe vill be modified to accept the new Bently Nevada robes. The existing Bently Nevada thrust bearing probes at the alance ring being longer, could aot be relocated and therefore vill be removed.

PCM 114-182 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; if (i) the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) 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 margin of safety as defined in the basis for any technical specification the is reduced.

The turbine supervisory instrumentation is non-safety related and non-seismic. The modification is being implemented for added protection of the turbine generator and to reduce potential unscheduled downtime. It improves plant reliability but does not otherwise affect the existing turbine design.

An evaluation for the impact of the added masses on the RTGB-101 and the change in the dynamic characteristics of the RTGB will be made by Civil Dept. It was found that the required modification has no significant impact on the dynamic characteristics of the board. The TSI cabinet is non-safety related but is located on the RAB elevation 43.0 level in the vicinity of safety related equipment. The cabinet was seismically analyzed and it was determined that it will maintain its structural integrity during a seismic event. Furthermore the displacement are sufficiently small to preclude interaction with ad)scent equipment. The implementation of this PCM does not require a change to the plant technical specifications.

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

0186L

PCM 355-183 THERMAL SHIELD REMOVAL PHASE 3 SYSTEM DESCRIPTION'.0 Function Inspection of the core support barrel (CSB) and thermal shield (TS) revealed damage at the CSB to TS connection. Phase III of the Thermal Shield Removal allovs the thermal shield segments to be removed from the Unit eventual offsite shipment.

fl refueling canal for 2.0 Desi Descri tion Removal of,. the TS from the Unit 01 refueling pool will involve the folloving stepsc

1. Loading 24" vide x 6'ong TS segments into th>> transfer shield in the Unit refueling pool.

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2. Transport of the transfer shield with contents out of the Unit fl containment and onto a truck vhich will transport the shield to Unit t2 via a prescribed route.

3. The transfer shield will be lifted from the truck and placed in the Unit f2 cask handling area where the transfer shield contents vill be put into a.

shipping cask for ofisite transport.

3.0 ~Oeraalan The operation of all equipment utilised in this effort shall be performed by a qualified operator.

0186L

PCM 355-183 SAFETY ANALYSIS The <<transport of thermal shield segments does not involve an unreviewed safety question becausec

l. (a) The probability of occurrence of an accident previously evaluated in the FSAR has not been affected since Unit 01 will not be operating while the transfer is taking place and no work vill be performed in or over the Unit spent tl fuel pool. Work being performed in Unit 02 is in the cask handling area and vill not affect operation of Unit f2 aa there is no fuel in the Unit t2 spent fuel pool. The haul route has already been evaluated for heavier loads than a fully loaded shield.

(b) The consequences of an accident previously evaluated in the FSAR have not been affected since the shield drop accident will not result in offsite doses in excess of those accidents previously evaluated.

(c) The probability of a malfunction of equipment important to safety has not been affected since the failure modes assumed as a result of the transfer process will not impact safety related equipment operation.

(d) The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR remain unchanged based on item la, lb, and lc above.

2. (a) The possibility of an accident of a different type than any analysed in the PSAR has not been created. The results of the analysis for a shield drop in excess of ll feet showed that corrective action could be taken while still maintaining offsite doses within a fraction of 10 CFR NO limits, thus remaining within the bounds of previous analyses.

(b) The possibility of a malfunction of equipment important to safety of a different type than any analyzed in the PSAR remains, unchanged for the reasons outlined above.

3. The transport of the thermal shield has no effect on the margin of safety as defined in the basis for any technical specification.

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PCM 177-184 ROSEMOUNT TEMPERATURE TRANSMITTER REPLACEMENT INTRODUCTION This PC/M is for the installation of nineteen (19) new Rochester model temperature transmitters to replace the existing Rosemount models. The existing Rosemount model 442 temperature transmitters are no longer available from Rosemount. A comparable model by Rochester Instrument Systems will satisfy the operational conditions as well as meet the safety/seismic requirements. This modification will also incorporate the implementation of two (2) signal transmitters to increase the load capability of two (2) transmitter loops mentioned in the system description.

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 consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be incrased; or (ii) if a possibility for an accident or malfunction of a differenty type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specifications is reduced.

These new Rochester model temperature transmitters are located in a mild environment and are seismically qualified to IEEE-344-1975. In addition, these units are manufactured to the Rochester Quality Assurance Program for nuclear devices; therefore meeting traceability and reportability requirements. This PC/M is for replacement of nineteen (19) transmitters, thus providing a more accurate and reliable model.

Therefore, this modification will not increase the probability of the occurence of any accident, whether previously evaluated or of a different type than previously evaluated and will not reduce the safety of the plant.

This PC/M does not reduce the margin of safety as defined in the basis of any technicaly specification, nor does it require a revis&n of a technical specification.

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

EDG SUBSYSTEM FLOW DIAGRAMS PCM 039-185 Modifioation D~eaori tion This PC/M releases the new Diesel Generator Subsystem Flow Diagrams to the site.

The following activities must be completed before the new flow diagrams can be issued as permanent plant drawings:

1.) All valves and instruments must be tagged in the field as per the new flow diagrams.

2.) Affected operating procedures must be reviewed to determine if revision is required to reflect the new tag numbers or flow diagram numbers.

Safet Anal sis 1a With respect to the probability of occurrence of an accident previously evaluated in the FSAR:

Plow diagrams are not considered in evaluating FSAR accidents.

lb. With respect to the consequences of an accident previously evaluated in the FSAR:

Flow diagrams are not considered in evaluating FSAR accidents.,

1c. With respect to the probability of malfunction of equipment important to safety pret! iously evaluated in FSAR:

Plow diagrams are not considered in determining the probabilities of safety related equipment malfunctions.

1d. With respect to the consequences of malfunction of equipment important to nuclear safety previously evaluated in the PSALM:

Plow diagrams are not considered in determining the probabilities of safety related equipment malfunctions.

2e. With respect to the possibility of an accident of a different type than analyzed in the FSAR'-

Flow diagrams are not considered in evaluating FSAR accidents.

2b. With respect to the possibility of a malfunction of a different type than analyzed in the FSAR:

Flow diagrams are not considered in determining the probabilities of safety related equipment malfunctions.

3. With respect to the margin of safety as defined in the basis for any technical specification:

Flow diagrams do not impact technical specification safety margins.

Based on the above, the new flow diagrams and the tagging/retagging of diesel generator valves and instruments are determined not to involve an unreviewed safety question. There are no system modifications involved.

PCM 090-185 ST LUCIE UNIT I REPLACEMENT OF LOAD WEIGHING. SYSTEM FOR REFUELING AND FUEL TRANSFER MACHINES hBSTRACT This engineering design package covers the replacement of the load~eigning system for the refueling and fuel transfer macnines. The ex>>ting system is manufactured by W C Dillon which no longer have spare

.parts available. Tne original equipment manfacturer, PAR Sys em Corporation,'will design, supply and install the new load weighing system. As discussed in UFSAR Chapter 9, this system is designed for safe handling and storage of fuel to and from the reactor. Tne equipment is normally used at 18 month intervals for a period of approximately three (3) weeks during which time it must operate continuously without maintenance or service. Also this sys em must be able to withstand loadings induced by the design base earthquake.

Therefore, this PCM is classified as "+ality Related". This item does not require revision to tne plant technical spe"ifications, nor does it meet the criteria for an unreviewed safety question. Tnerefore, pursuant to 10CFR50.59 this modification can be made witnout prior commiss ion approval.

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pCH 090-185 SAFETY EVE.UATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question; (i) if the probability of occurence or the consequences of an accident or malfunction of equipment important to safety previously evaluated xn the safety analysis report may be increased; or (ii) if a possiblility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

The modification described in this PC/M replaces existing components associated with the refueling machine. Tne refueling machine is only required to operate for approximately three (3) weeks at eighteen (18) month intervals as per UFSAR Section 9. 1.4.2. This system is not required for normal operation of the plant. It is only required for fueling and removal of fuel from the reactor, therefore this equipment is not required for safe shutdown of the plant. With regard to spent fuel handling accidents as described in VFSAR Section 15.4.3, the results of the fuel handling accident are not affected by this equipment change out. The digital readout will ensure that when removing a fuel assembly it is not damaged by excessive lifting forces. The digital scale should in fact provide more accurate information to the operator to better preclude this event.

The new equipment has been seismically analyzed to preclude its disalignment during a seismic event. Therefore, this will not impa"t the "light loads" accident analysis.

The failure of this component, not to function, would preclude further fuel movement until its repair. Redundancy of this system is not required. In as so much that this system is not required to shutdown the reactor, cool the core or cool another safety system or the reactor containment (after an accident), nor is it part of any system that reduces radioactivity released in an accident. Note, only these portions of a system that are designed primarily to accomplish one ef the above functions, or the failure of which could prevent accomplishing, one of the above functions, is designated 'safety related. The system is required to withstand loadings induced by the design bases earthquake. Therefore, this PCM is classified "Quality Related".

Tne modifications to tne load cell supports shall be designed and analyzed by PAR Systems as to maintain tne seismic integrity of the equipment. Tne results of their analysis will be reviewed by Ebasco.

Tne implementation of this PC/M does not require a change of the plant spe"ifications.

"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 Commission approval for the implementation of this PCM is not required."

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PCM 092-185 ST LUCIE UNlT 41 ESFAS POWER SUPPLY This modlficatlon provides for the replacement of ESFAS measurement cabinet's Instrument loop power supplies which are no longer available from the orlglnal equipment manufacturer. These supplies furnish the sensor current for containment pressure, and refueling water tank level.,

This change performs a nuclear safety related function and is powered from Class 1E safety sources. This PC/M does not involve as unreviewed safety question.

a.

since the ESFAS is not utilized in determining the probabilities of accidents.

b.

this modification does not affect the availability, redundancy,-capacity, or function of any equipment required to mitigate the effects of an accident.

c This modification does not affect any other safety related equipment.

d.

~~~ Redundancy, function, or failure mode capacity have not been changed.

Be this modification does not affect any systems vital to safety or whose failure could directly result in an uncontrolled release of radioactive material.

f. The possibility of equipment malfunction of a different type than analyzed in the FSAR has not been increased.

The margin of safety as defined ln the basis for any Technical Specification has not been changed since this modificatio does not change the performance, capabilities, or operating characteristics of the ESFAS.

In conclusion this change does not Involve an unreviewed safety question.

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pCM 177-185 REPLACEMENT OF VALVE SOLENOIDS INTRODUCTION Oe The purpose of this PC/M is to replace twelve (12) existing ASCO and two (2) AVCO valve solenoids, no longer manufactured, with "Qualified" ASCO solenoid NP-8316 Series on various valves. The replacement solenoids IEEE-382-are environmentally qualified IEEE-323-1974, IEEE-344-1975, and to be 1972. Also four (4) of the existing ASCO solenoid valve seats are rebuilt, using ASCO spare parts kit components (Elastomers)

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 questionl (i) if the probability of occurence or or the consequence 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 does not involve an unreviewed safety question and the following provides the bases for this conclusion.

The replacement of ASCO valve solenoids have been qualified to IEEE-323-1974, and IEEE-383-1972 requirements. The qualification documentation package has been provided by ASCO via their report AQS-21678/TR Rev A.

The qualification test program simulated the effects of long-term operation under normal operating conditions and the effects of Design'asis Accident .

(DBA) the effects included exposure to the environmental extremes of temperature, pressure, humidity, radiation, vibration, and chemical spray.

With adequate margin, the qualification program demonstrated that the equipment can perform its specified function under the anticipated normal operating and DBA conditions.

The replacement of ASCO valve solenoids are one-for-one replacement of the existing ASCO solenoids. For the AVCO solenoids, valves FCV-25-2 and FCV-25-5, the addition of tubing fittings for the air supply system is required. However, there is no effect on the seismic qualification of those valves due to the additional weight of the required fittings.

The replacement of ASCO valve solenoids for the CCW isolation valves is also a one-for-one replacement of the existing ASCO .solenoids. The re" placement of these solenoids enhance the existing valves by installing environmentally qualified solenoids.

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

The foregoing constitutes, per 10CFR50.59 (b), the written safety evalua-tion which provides the bases that, this cnange does not involve an unre-viewed safety question, therefore, prior Commission approval is not re-quired for implementation of the PC/M.

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peg 204-185 ST LUCIE UNITS 1 R 2 ICW BACKUP LUBEWATER BACKFLOW PREVENTOR (REANLN<5<1)

ABSTRACT Engineering Package covers replacement of the existing ICW service water

'his backup lubewater supply backflow preventor with one that is currently manufactured. The existing backflow preventor vendor no longer manufactures these thus spare parts are difficult to obtain. This package is classified as non-seismic, nonnuclear safety related.

SAFETY EVALUATION With respect to Title 10 of the code of Federal Regulations, Part 50.59, a proposed change shaQ be deemed to involve an unreviewed safety question: (1) if the probability of occurence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iio if the margin of safety as defined in the bases for any technical specifications is reduced The subject modification provides for replacement of the existing backflow preventor.

This backflow pr eventor ensures that seawater does not backflow into the service water system thus contaminating the domestic water supply. With respect to 10CFR 50.59, failure of this backflow preventor: (1) does not increase the probability of an'accident or malfunction of equipment important to safety since the supply is separated from ail safety related equipment by a double check valve class break and is located remotely with respect to such safety related equipment and cannot fall on or hit such equipment; or (2) does not create possible accident scenarios not previously addressed by the Safety Analysis Report since it functions only as a system enhancement and does not have to function in any postulated accident conditions; or (3) does not affect or require changes to the Technical Specifications.

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

Therefore prior commission approval is not required for implementation of this PC/5I.

Additionally per the FSAR Section 9, this backup lubewater supply is not required to perform any safety related functions nor is the modification within any safety related or seismic boundaries. Therefore the modification is considered to be nonwafety related, Quality Group D.

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PCM 028-186 ST. LUCIE UNXT 1 Component Cooling Water Pump Journal Bearhg Shell Material Clumge (REA SLN-421-Ss-S)

This engineering package covers replacement of the existing cast Iron'journal bearing shells on the component cooling water pumps 1A, 1B 8 1C with shells made of carbon steel. The existing cast iron shells are no longer available and the manufacturer's replacement part Is the carbon steel shell. As addressed in the Safety Evaluation, this modification is considered nuclear safety related.

Based on the 10 CFR SO.S9 review, it has been demonstrated that this change does not involve an unreviewed safety question, and the change will not affect plant safety. Additionally, no change is required fo the Technical Specifications.

Accordingly, prior NRC approval is not required for implementation of this design.

SAFETY EVALUATION The Unit 1 Component Cooling Water pumps are nuclear safety related and are classified as ASME Section III, Class 3 QuaHty Group C components.

They are required to provide a heat sink for safety related components associated with reactor decay heat removal for safe shutdown or LOCA conditions. The journal bearing shell material change affects both journal bearings in the IA, 1B and 1C pumps.

Failure of the bearing shell (regardless of material utilized) and respective journal bearing will result in failure of the component cooHng water pump.

However, failure of a single pump has been previously evaluated and has been accounted for in the Component CooHng Water System design bases as identified in the FSAR. Measures exist to ensure adequate decay heat removal for safe shutdown or LOCA conditions should a single pump faIL Since the new shell parts are internal to the bearing housing, failure of an additional component cooling water pump simultaneous to the first pump failure is not possible based on single failure criteria. In addition, since the new shell material is functionally equal or better than the existing cast iron material, the probabiHty of pump failure remains unchanged Based on the above evaluation and information provided in the Design Analysis, it can be demonstrated that an unreviewed safety question as defined by 10 CFR 50.59 is not created. Since no other accident beyond what has been previously addressed in the FSAR has been identified and no other safety related equipment or components are affected as addressed in the failure modes analysig the probability of occurence of analyzed accidents has not been increased. The replacement is equal or better to the equipment replaced. No new accidents or malfunctions are introduced as a result of this design change. Additionally, the margin of safety as defined in .the Technical Specifications has not been reduced and no Technical Specification changes are requIred. Therefore an unreviewed safety question does not exist.

Since this modification does not Involve an unreviewed safety question and does not change or alter the Technical Specifications, this change is acceptable with respect to 10 CFR 50.59 and does not require NRC appr oval prior to Implementation.

- PCM 050-186 ST LUCIE PLANT UNIT NO 1 INSTRUMENT AIR UPGRADE REA-SLN-481 Qo

'ABSTRACT This Engineering Package (EP) is for the installation of 2 new air compressors, 2 new desiccant air dryers and removal of the existing desiccant air dryer, afterfilter package and refrigerant air dryer which do not have sufficient capacity to accomodate the new compressors. One of the two new air compressors and one new air dryer will operate and the other will serve as a standby. The existing compressors will remain as backup, especially for loss of offsite power, since only these compressors can be loaded on the diesel generator. In addition, the backup air supply to the MSIVs and FCV9011 and FCV9021 will be removed since the new compressors will be able to supply adequate air flow at the required pressure.

This EP is classified as Non-Safety Related since the instrument air (IA) system compressors and associated equipment performs no 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 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 a aintenance guidelines and FSAR change package.

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

SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulation, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question; (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.

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

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PCM 050-186 afety Evaluation (Continued)

Pailuze of the iastrumeat air compxessore aad components resulting in loss of IA aad consequent affects as stated fa the PShR Subsection 9.3.1.3 have been reviewed. Thfs modification does not add any new failure modes for the safety related air operated valves. However, exlstfag solenoid valves, without regulators, whose mazimum operating pressure differential capacfty is less than U.5 psi will be replaced via Design Equfvalent Engineering Package (DEEP) No 154-188D. Mal&mction, if aay, of these solenoid valves will lead to the Pail Safe mode of the process valves. The Ih system design pressure and tempex'ature downstream of the Ih aftezcooler remain unchanged, therefore there is no concern for the valve actuators. This modification is thex'efore classified as nonnuclear Safety Qud.fty Group D and nonmlase 1E.

The increase in Zh requirements from 155 SCRM to 400 SCFM aad pressure from 90-100 peig to 105-U.5 psfg 1s based on FPL studies for tlat requixement of the IA.

Removal of the supplemental air bottle racks whfch arc tied into the accummulators to maintaia the MSIVs and feedwater PCVs air system pressure between 100-105 psig is considered acceptable'because the new (1C and lD) will be able to pxovide adequate instrument aiz 'ompressors flow at the required pressure.

Based on the above 'descriptioa, the modificatfon included fn this EP ie considered to be non-safety related. Thfs EP does not involve an unreviewed safety question, aad following are the bases for this Justification'.

(1) The probability of occurxence or the consequences of an accfdent or malfunction of equipment important to safety previously evaluated ia thc safety analysis report is not increased. The instrument air system compressors and associated equipment ax'e not used directly in any safety analysis for accideats oz malfunction of equipment and as such aze noa-safety related and will have no effect on equipmeat vital to pleat safety.

(ii) The possibility for an accident or malfunction of a different

. type than any evaluated previously fn the safety analysfs report is not created. The components involved in this modification have no safety related function and no changes have been made to the aormal operational design of the system with the compressors 1C and 1D fa operatfon. In thfs mode the IA compressozs XA and 1B discharge valves V18109 and V18119 aze closed to prevent IA lcakagc via these compressors.

Similarly, whenever the Ik compxessors LL and 1B are rcquircd to operate, valve V18586 is closed to prevent Ih leakage via compressors 1C aad 1D (iii) The margin of safety ae defined in the bases for any Technical Spec1fication is aot affected by this PCM, since the components fnvolved ia this modification are not included in the bases of any Tcchnical Specfficatioa.

pCM 052-186 ST LUCIE UNIT 1 Upper Guide Structure (UGQ Lift Rig Repair REA-SLN-86-010 ABSTRACT This engineering package covers the repairs and modification to the damaged UGS Lift Rig. The lift rig was damaged on November 6, 1935 when one of its attachment points failed during a UGS lift operation. Although the UGS Lift Rig is non safety-related, its failure could result in damage to nearby safety-related equipment. Therefore, quality-related design requirements have been imposed to assure QC Inspection of the repairs and modification. The implementation of this PCM does not pose any unreviewed safety questions nor does it affect any safety-related equipment.

S lement 1 This supplement documents the "as-built" configuration of the repaired UGS lift rig. It includes the actual column chord dimensions and incorporates the minor design.

changes that were made in the field during the repair effort. The original safety evaluation has been reviewed for the impact of the changes addressed by this supplement and it has been determined that it remains valid.

Safe Evaluation This engineering package provides for the repair and modification of the UGS lift rig to comply with its original functional requirements. The UGS lift rig is not a safety-related piece of equipment and is only used during refueling operations when the plant Is in the cold shutdown mode. Since failure of the lift rig while lifting the UGS could result in a load drop onto the reactor and irradiated fuel assemblies, this component is considered important to safety. For this reason, quality-related design requirements have been imposed to assure QC inspection of the repairs and modification.

The modified lift rig has been structurally reanalyzed for dead and seismic loads subject to the requirements of NUREG 0612, ANSI N14.6, and the applicable ASME and ASTM codes. The results of this analysis demonstrates that the new and existing components are all within allowable stress levels.

The containment heat sink, hydrogen generating source, and free volume analyses described In FSAR Section 6.2 are not affected by this modification, since the lift rig replacement parts are the same as or similar to those installed originally.

This modification does not change any assumptions made or conclusions drawn In the St. Lucie FSAR, and there is no'new failure mode introduced that has not been previously evaluated in the FSAR. However, FSAR Figure 9.1-8 must be updated to reflect the repairs and modification to the liftrig.

For the above reasons, the repairs and modifications to the UGS lift rig will Pot inctease the probability of occurrence nor the consequences of a design basis accident or malfunction of equipment Important to the safety of the plant.

Additionally, there will continue to be no possibility of an accident or malfunction different than those already evaluated In the FSAR. Finally, the margin of safety as defined in the Plant Technical Specifications has not been reduced. It Is therefore concluded that this modification does not pose an unreviewed safety questions pursuant to 10 CFR 50.59 and does not affect any Technical Specifications.

PCM 07I4-186 HEATER DRAIN PUMP DEMINERALIZED WATER SUPPLY msmACr This design package provides the required engineering for adding permanent piping from the demineralized water system to the Unit 1 heater drain seals. The piping will make available to the seals the necessary back pumps'echanical up flushing water meeting the appropriate chemistry requirements. The back up water source is required during initial plant startup whenever the pumps sit idle.

Based on the failure modes analysis and 10 CFR 50.59 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 and is correctly classified as Non-Nuclear Safety Related.

Implementation of this modification, therefore, does not require prior NRC approyal.

Su lement 1 This package revision provides valve drawings for valves added by this FC/M 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.

ShFETY EVhLUhTION The Unit 1 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 corn ponents.

i Based on the above evaluation and information supplied in the design analysis it can be demonstrated that an unreviewed safety question as defined in 10CPR50.59 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 077-186 10 CFR 50.59 EQ LIST REVISION ABSTRACT several areas This engineering design package provides the vehicle for updating to the of equipment qualification. This package includes corrections requirements, and various 10CFR50.49 list, changes in maintenance documentation corrections.

This design package is considered nuclear safety related because itdoes not affects equipment faDing under the scope of 10CFR50.49. This package represent an unreviewed safety question since it deals strictly with enhancing the present documentation used to qualify equipment at St. Lucie.

Safet Evaluation This engineering design package provides for several documentation changes to the present St Lucie Unit No. 1's equipment quaHfication documentation. This documentation will affect the future procurement of various safety related components and assist in validating the to function during a design basis accident. Therefore, this design components'bility package is considered safety relate*

The documentation changes addressed in, this package range from corrections of typographical. errors on the 10CFR50.49 list to reviews of a vendor's equipment qualification test report. None of the changes require physical modification to any plant system. They do however, affect the future maintenance of various instruments.

Based on the above and the information supplied in the design analysis it can be deomonstrated that an unreviewed safety question as defined by 10CFR50.59 does not exist. Since this change does not alter any equipment used to mitigate accidents, the probability of occurrence of an analyzed accident remains unaffected. This design package only enhances the environmental documentation of various instrumentation and in no way affects the plant design, therefore the possibiHty of an unanalyzed accident or malfunction has not been created.

The surveiDance requirements of the Technical Specifications were rhviewed against the equipment qualification maintenance requirements addressed in this package in the design analysis. No Technical Specification changes are required by this design package.

In conclusion, the changes proposed by this design package are acceptable from the standpoint of nuclear safety because they do not involve an unreviewed safety question and no Technical Specification changes are required.

SUPXramKHT fl The revisions incorporated by sappleaent fl do not affect the orig inal safety evaluation.

PCM 099-186 ST LUCIE PLANT - UNIT NO. 1 UPPER GUIDE STRUCTURE LIFT RIG LOAD TEST FIXTURE REA SLN-86-010 ABSTRACT This Plant Change/Modification (PC/M) consists of the fabrication and installation of a temporary structure which will be used to load-test the Upper Guide Structure (UGS) Lift Rig after it is repaired and modified. The structure will be attached to the reactor missile shields in their laydown area. The static load test will be performed by the reactor polar crane using the missile shields as test loads. After the load test, this temporary structure will be removed from the containment.

This PC/M is not classified as safety-related since the load test structure will not perform or affect any safety-related function. Although failure of the test fizture will not result in any interaction with safety-related equipment or functions, Quality Related requirements will be applied to the design because of the importance of thd UGS Lift Rig to plant operations. The Quality Related design requirements assure Q. C. inspection of the installation and independent verification of the design of the load test structure.

This PC/M does not constitute an unreviewed safety question. The only effect on plant operations will occur during the refueling outage UGS Lift Rig load test. The implementation of this PC/M does not affect- any safety-related equipment.

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PCM 099-186 SAFETY EVALUATION With respect to title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question: (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

This modification is not classified as safety related since the load test structure does not perform any safety function. Failure of the load test structure before, during, or after the load test could not affect any safety-related equipment or function since the structure will be installed in the containment only during the refueling outage and is located away from any safety-related components. Failure of the load test structure during the load test could damage the UGS Lift Rig. The UGS Lift Rig is not a safety-related component, but it is important since its failure during lifting of the UGS could result in a load drop onto the reactor and irradiated fuel assemblies. For this reason, the load test structure has been designed using ality Related requirements.

The containment heat sink analysis inventory of hydrogen generating items and free volume assumptions described in FUSAR Section 6.2, are not affected by this modification, since the load test structure is temporary and will be removed from the containment prior to plant operation.

The modifications included in this PC/M do not involve 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 performing a safety function.

ii There is no possibility for an accident or malfunction of a different type than any previously evaluated since the load test structure performs no safety function and no changes

.have been made to any operational design.

, Failure of the load test structure will have no effect on any safety-related equipment or function.

This modification does not change the margin of safety as defined in the basis for 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; therefore, prior Commission approval is not required for implementation of this PC/M.

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PCM 114-186 ST LUCIE PLANT - UNIT HO 1 CONDENSATE PUMPS EXPANSION JOINT REPLACEMENT REA-SLN-85-153 hBSTRACT The existing expansion )oints in the Condensate Pumps suction are made of elastomeric material which has deteriorated due to aging. The deterioration is so severe that the Condensate System is susceptible to air permeation. To correct this problem the existing expansion )oints will be replaced with new stainless steel expansion )oints.

The Condensate System considered in this Engineering Package is non-safety related. Accordingly, this Engineering Package is classified as non-safety related. The safety evaluation has shown that this EP does not impact plant safety and operation and does not constitute an unreviewed safety question or require a Technical Specification change.

Therefore, prior NRC approval is not required for implementation.

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PCM 114-186 Safet Evaluation With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question; (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.

The Condensate Pump expansion joints are located between the Condenser and Condensate Pumps and are utilized to absorb differential thermal expansion between these components. They do not perform any safety related function and therefore are classified as non-safety class, Quality Group D. The failure analysis has shown that failure of these components vill not affect any safety related equipment.

This non-safety related modification does 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 in the safety analysis report is not increased. The St Lucie Unit No 1 PSAR, Section 10.4.6 describes the Condensate System. This Section indicates that the portion of Condensate System being modified and the expansion joint are not designed to seismic Class I standards and they are not used in any safety analysis for accidents or malfunction of equipment. This system is 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 is not created. The components involved in this modification do not perform any safety related function. No changes have been made to the operational design of the Condensate System.

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

The implementation of this PCM does not require a change to the plant Technical Specification.

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

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PCM NO. 119-186 REV NO 1 ST LUCIE PLANT - UNIT NO 1 10CFR50.49 ENVIRONMENTAL QUALIFICATION LIST REVISION REA SLN-85-58 ABSTRACT This Engineering Package provides the vehicle for updating several areas of equipment qualification. This package includes corrections to the 10CFR50.49 list, changes in maintenance requirements, and various documentation package corrections.

This Engineering Package (EP) is considered Nuclear Safety Related because it affects equipment falling under the scope of 10CFR50.49. This package does not represent an unreviewed safety question since it deals strictly with enhancing the present documentation used to qualify equipment at St Lucie Unit No 1 and no physical plant modifications are required by the EP. The safety evaluation of this package indicates that a change to the Plant Technical Specifications is not required. Removal of equipment from the 10CFR50.49 list does not affect plant safety and operation.

Su lement 1 This EP revision adds terminal blocks to the 10CFR50.49 list and thei,r associated Equipment Qualification Documentation Package 8770-A-451-17.0 "Amerace Terminal Blocks". The equipment and EQ Documentation Package does not affect the original safety evaluation.

PCM 119-186 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 unrevised 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 possib1lity 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 marg1n of safety as defined in the bases for any technical specification is reduced.

This Engineering Package provides for several changes to the present St Lucie Un1t No. l's 10CFR50.49 list. This documentation &ll affect the future procurement of various safety related components and assist in validating the components'bility to function before, during and after a design basis accident. Therefore, this EP is considered Nuclear Safety Related.

The documentation changes addressed in this package range from corrections of typographical errors on the 10CFR50.49 list to additions and deletions of equipment as a result of Eg documentation packages reviews. None of the changes require physical modification to any plant system. They do, ho@ever, affect the future maintenance of various equipment.

The possibility of new Design Basis Events (DBEs) not considered 1n the UFSAR is not created since this change does not alter any equipment used to mitigate accidents. This modification is an enhancement of the environmental qualification documentation of various equipment and in no @ay affects the plant design.

PCM 119-186 Due to the fact that this EP does not affect or modify any cables essential to safe reactor shutdown or systems associated with achieving and maintaining shutdowns, this package has no impact on 10CFR50 Appendir "R" fire pxotection requirements. Therefoxe the proposed design of this package is in compliance with the applicable codes and UFSAR requirements for fire protection equipment.

Since this modification involves no physical modifications to safety related equipment and changes in the maintenance schedules will not result In failure of equipment, the degree of protection provided to Nuclear Safety Related equipment is unchanged. Removal of equipment from the 10CFR50.49 list does not affect the plant's safety. The probability of malfunction of equipment is unchanged. The probability.

of malfunction of equipment impoxtant to safety previously evaluated in the UFSAR remains unchanged. The consequences of malfunction of equipment important to safety previously evaluated in the UFSAR are unchanged. The possibility of malfunctions of a different type than those analyzed in the UFSAR is not cxeated.

Based on the above, the modifications included in this Engineering Package do not involve an unreviewed safety question because of the following reasons:

The probability of occurrence and the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report will not be increased by this modification because it'oes not affect the availability, redundancy, capacity, or function of any equipment required to mitigate the effects of an accident.

(ii) The possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report will not be cxeated by this modification. Function, mounting and the ability to withstand harsh environmental conditions have not been altered and this modification does not affect any other safety related equipment.

(iii) The maxgin of safety as defined in the bases for any technical specification is not reduced since this modification does not change the requirements of the Technical Specifications.

The Implementation of this PCM 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.

PCM 128-186 ST LUCIE PLhNT UNIT NO 1 STARTlJP TBhNSFORMER LOCXOUT DISCONNECT SWITCHES REA-SIR&77-10 ASSXRhCT This Engineering Package (EP) provides for the installation of disconnect switches in the plant startup transformers lockout relay circuits. The purpose of this change is to facilitate lockout relay maintenance testing while eliminating the possibility of inadvertent plant trip by propagation of a lockout relay trip during lockout relay maintenance test.

This EP is classified as ~1ity Related since lockout circuit actuation vill trip the startup transformer and would result in plant operation under Limiting Conditions for Operation as defined in the Plant Technical Specification. Subsequent loss of offsite power to the station buses could affect plant trip, starting and loading Baergency Diesel Generators. h review of the changes to be implemented by this PCM @as performed in accordance Wth the requirements of 10CFR50.59. hs indicated in the Safety Evaluation (Section 3.0), this PCM does not involve an unreviewed safety question, nor does it require a revision to the plant Technical Specifications. This modification vill have no effect on plant safety or operation. Prior Comnission approval is not required for the implementation of this PCM.

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SAFETY EVALUATION Paf 128-186 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 barris for any Technical Specification is reduced.

The modifications included in this Engineering Package 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, in FSAR Section 8.3.1.1, is not increased since the startup transformers and their lockout trip circuits are not Nuclear Safety Related equipment. Failure of the test switches will not affect the availability of the Emergency Diesel Generators in the event of loss of offsite power (LOOP).

ii) There is no possibility for an accident or malfuncgion of a different type than any previously evaluated since the startup transformers are used for plant startup and shutdown. Za the event of test switch failure which may result in unavailability of the preferred offsite power source (start~ transformer), -the emergency diesel generators can provide the power required for safe shutdown as previously evaluated in FSAR Section 8.3.

(iii) This modification does not change the margin of safety as defined in the basis for any Technical Specification. This has been determined based on the fact that this modification does not exceed the limitations of Plant Technical Specification and does not affect safe reactor shutdown, the mitigation of the consequences of a design basis event (DBE), or the control of radioactive releases to the environment.

This EP affects equipment that is Nonnuclear Safety Related.

However, since startup transformer 'failure, and startup transformer trip signal actuation wiU. result in plant operation under Technical Specification limitations, this EP is classified as Quality Related.

This EP has no effect on cables essential to safe reactor shutdown or components listed on the Essential Equipment List. There are no changes to equipment involving 10CFR50 Appendix "R" Fire Protection requirements (see attachment 7.1). Thus, the proposed design of this package is in compliance with the applicable codes and FSAR requirements for fire protection equipment.

Implementation of this PCM does not require a change to the Plant Technical Specifications and may be implemented without prior Commission approval.

The foregoing constitutes, per 10CFR 50 59 (b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question nor does it require a revision to the plant Technical Specifications,and prior Commission approval for the implementation of this PCM is not required.

PCM 131-186 ST LUCIE UNIT NO 1 AUTO LEAR RATE TESTER POR PERSONNEL AIR LOCKS REA-SLN-86&05 ABSTRACT This engineering package allows for the replacement of the ezisting Volumetrics Automatic Leak Test System model 14324 (obsolete) with the currently available Volumetrics model 14330-2. This system will provide both local and remote (main control room) alarm on failure of leak rate test.

Since there are no essential cables associated with this EP, this package has no impact on 10CFR50 hppendiz "R" requirements.

The leak rate test system is not required for safe reactor shutdown and does not serve to mitigate the consequences of a design bases event (DBE) and is therefore not safety related equipment. However, since this package includes modifications to control board annunciators, and is required to maintain the limits of St Lucie - Unit 1 Technical Specification Section 3/4.6, "Containment hir I,ocks," it is considered Quality Related.

This engineering package will restore automatic test capabilities to the personnel air locks and reduce manpower requirements to manually operate the ezisting leak rate testers. The ezisting interior door tester, currently inside the containment vessel, is relocated outside the ezterior door so as to

<<inimize personnel contact with the RCA.

The implementation of this PCN does not require any change to the St Lucie-Unit 1 Technical Specifications. The modifications do not involve an unreviewed safety question and prior Commission approval for the implementation of this package is not required.

SAFETY EVALUATION PCN 131-186 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 occurxence or the consequences of an accident or

<<alfunction of equipment i<<portant 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 <<ay be created, or (iii) if the margin of safety as

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

The proposed modification affects personnel air hatch leak rate testing which provides locaL and remote (main control zoom) alarm on test failure.

The pxobability of occurence of a DBE pxeviously addressed in the FOSAR is not affected by this <<odification. This system will in fact decrease the probability of a breach of containment by assuring containment integrity.

Failure of this system to operate propexly will be annunciated thereby preventing the performance of inaccurate testing. The possibiLity of new DBEs not considered in the FUSAR is not created since. the design philosophy has been previously discussed in the FOSAR. This <<odification is an enhancement to a pre~isting system as is being performed to provide the highest caliber equipment possibLe.

Due to the fact that this EP does not involve any cables essential to safe reactor shutdown ox systems associated with achieving and maintaining safe shutdown conditions, this package has no impact on 10CFR50 hppendiz "R" fire protection requirements. Therefore the proposed design of this package is in compliance with the applicable codes and FOSAR zequirements for fire protection equipment.

'he leak detection system is not necessary for safe reactor shutdown nor does it serve to <<itigate the consequences of a design bases event (DBE. Hence, this package is not safety related. since re<<ote alarm in the However, a Main Contxol Board is provided by the Containment Personnel Air Lock Automatic Test System and <<odifications to the annunciatox panels are included in this design, this package is Quality Related. ~Lity Control Engineer shall witness the installation of the new annunciator tile in the Contxol Room as well as the auto leak rate tester in the personnel access area.

As the evaluation of failure <<ode (Section 2.2 7) indicates, the failure mode of this system has no effect on safety related systems ox'quipment. Bence the degree of protection provided to nuclear safety reLated equipment is unchanged. The probability of malfunction of equipment i<<portant to safety, pxeviously evaluated in the FOSAR remains unchanged. The consequences of

<<alfunction of equipment important to safety previously evaluated in the FOSAR ax'e unchanged. The possibility of <<alfunctions of a different type than those analysed in the FOSAR is not created.

The implementation of Qality Related PC/M 131-186 does not requixe a change to the plant technical specifications, nor does it create an unreviewed safety questions.

The foregoing constitutes, per 10 CFR 50.59(b), the written safety evaluation which provides the bases that this change does not involve an unzeviewed safety question and prior Commission approval for the implementation, of this PCM is not requiredi

pCM 133-186 ST. LUCIE UNIT 1 SPDS SOFTMARE MODS ABSTRACT This Engineering Package covers the modifications to the previously certified software of the gualified Safety Parameter Display System (gSPDS). The modifications consist of additions to assist the plant operator in accident

.monitoring. There is no major gSPDS hardware modification as a result of this PC/M. However, the exchange of identical Erasable Programmable Read Only Memory (EPROM) chips were required as a results of software modifications.

This Engineering Package is safety related because it involves modifications to a safety graded system gSPDS. The gSPDS is a safety grade class lE processing and display system used for post-accident monitoring. The hardware and software changes of'his PC/M were evaluated against IOCFR 50.59. The results of the evaluation are that there is no unreviewed safety question.

PCM 133-186 SAFETY EVALUATION This engineering package is safety related because it involves a modification to a safety graded system. We have evaluated the effects of this PC/M with respect to regulation 10CFR50.59. The two applicable items for the QSPDS are:

a) Unreviewed Safety Questions There are no major hardware changes due to this PC/M, since the exchanged hardware (EPROM's) are identical to original. The software changes consist of the addition of one display page which is consistent with the requirements of format, content and visibility of the original design. Therefore, there is no increase in the probability of occurrence or consequence of an accident, or malfunction of equipment because of this modification to the QSPDS. The possibility of an accident or malfunction of a different type than any evaluated reviously in the FSAR has not been created. In addition the margin prev of safety is not decreased by this PC/M. Instead, the safety margin is considered to be increased due to the increased visibility of the safety parameters by the operator as a result of this PC/M.

b) Technical Specifications The requirements established in the Technical Specification for the QSPDS are unaffected by this PC/M. The changes of this PC/M did not affect design, nor previous function, it merely improved Human Factors Engineering considerations.

PCM 138-186 BECKMAN WASTE GAS SYSTEM OXYGEN ANALYZER REPLACEMENT ST LUCIE PLANT - UNIT NO 1 REA-SLN-86-030 In order to increase the availability of the oxygen analyzers for frequent monitoring of the oxygen levels in the waste gas decay tanks, the existing oxygen analyzers will be replaced with updated oxygen analyzers having an analytical element designed for sampling services in either liquid or gaseous sample streams.

The inherent design features of the replacement analyzers will include the design and operational criteria for sample monitoring and installation in potentially hazardous locations, therefore, this design shall be considered as Quality Related.

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

A review of the changes to be implemented by this PCM was performed against the requirements of 10CFR50.59. As indicated in Section 3.0 of this safety Engineering Package (EP), this PCM does not involve an unreviewed -"t question, nor does it require a revision to the technics sp- i therefore, prior Commission approval is not required for implementation of this PCM.

SAFETY EVALUATION PCM 138-186 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:

if (i) the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the 0 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.

i) The probability of occurrence or the consequences of an accident or malfunct5on of equipment important to safety previously evaluated in the Safety Analysis Report is not increased since the Oxygen Analyzers are used for frequent monitoring of oxygen concentrations in the waste decay tanks and as described in PSL-1 FSAR Subsection 11.3.2.1 this system's function is not essential for the safety of the plant. The replacement of Oxygen Analyzers will provide control improvements to.maintain the Waste Gas Analysis System functional with significant reduction in system maintenance and component replacements.

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

a) This installation is in accordance with the Code of Federal Regulation 10 CFR 50.49 and no impact is incurred by this installation.

b) The new equipment mountings and added components have been analyzed in accordance with the specification for the Design Fabrication and Erection of Structural Steel for Building, and it has been determined that the stresses with the new equipment are less t5an the panel stresses with the original equipment.

c) 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 Environmental Qualification criteria since the equipment does not monitor or mitigate the event causing the harsh environment.

d) The Waste Gas System Oxygen Analyzers are neither required for safe shutdown nor for mitigating the consequences of an accident.

iii) The margin of safety as defined in the bases for any Technical Specifications is not affected by this EP 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 CFR50.59(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 140-186 ST LUCIE PLANT - UNIT 1 ANNUNCIATOR NUISANCE hIhLNS REA-SLN-86-052 ABSTRACT This Engineering Package (EP) covers the modifications of five annunciator circuits in the Main Control Room. Existing logic, circuit configuration and components will be changed in the Reactor Turbine Generator Boards (RTGBs) so as to eliminate existing nuisance conditions caused by erroneous alarm indication of these five annunciator circuits. By implementing this EP, these circuits will be consistent with the "Dark Annunciator" concept which allows for lighted annunciators during off~ormal conditions only.

This EP is classified as Nuclear Safety Related since it involves the interposing of a control relay in a safety related circuit (hydrogen analyzer) and the extension of safety related power supply cables (10482E, 10482L, and 10485H). The safety evaluation has determined that this EP does not involve an unreviewed safety question and does not require a change in the plant technical specifications. This PCM may be implemented without prior NRC approval.

SUPPLANT 1 This Engineering Package Revision covers modification of the six annunciator circuits associated with annunciated windows P"30, P-35, P-36> P-42, Q-40 and X-5 in the Control Room. These modifications, which include relocation of-local reset switches, installation of reflashers and logic modifications, will eliminate the nuisance alarm status of the six annunciators. By implementing this PCM Supplement, these six annunciators will be brought into compliance with the "Dark Annunciator" concept of NUREG 0700 "Guidelines for Control Room Design Revie~" ~

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 the technical specifications. Therefore, prior NRC approval is not required for implementation of the PCM.

The intent of the original Safety Evaluation is not affected by this supplement.

SUPPLEMENT 2 This Engineering Package Revision covers modification of the three annunciator circui.ts associated with annunciator windows N-45, R-50, and S-24 in the Control Room. These modifications, which include the installation of four (4) relays, evaluation to support setpoint modifications and drawing corrections, will eliminate the nuisance alarm status of the annunciators.

The Safety Evaluation of Supplement 1 to this PCM has been revised. The Safety Evaluation still concludes that this EP does not involve an unreviewed safety question or a change to the Technical Specifications. Therefore, prior Commission approval is not required for implementation of the PCM. The intent of the original Safety Evaluation is not affected by this supplement.

PCM 140-186 SAFETY EVALUATION Pith respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed changed 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:

(i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased since the annunciators serve no function in the control of plant operations or safe shutdown.

Electrical separation is provided between redundant safety related wiring and components and annunciator logic which is separated to protect control functions from being affected by annunciation circuit failure.

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

Since this EP affects equipment that is identified as Nuclear Safety Related (Hydrogen Analyzer and SI Tank Isolation Valves 3614, 3624, 3634. 6 3644) and requires the extension of Nuclear Safety Related power supply cables (10482E, 10482L> and 10485H),

Safety Related.

it is considered Nuclear Due to the fact that the EP does not involve any cables essential to safe reactor shutdown or systems associated with achieving and maintaining safe shutdown conditions, this package has no impact on 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.

Implementation of Nuclear Safety Related PCM 140-186 and Supplements 1 4 2 to the same PCM do not require a change to the plant technical specifications and may be implemented without prior NRC approval.

The foregoing conititutes, per 10CFR 50.59 (b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question and prior NRC approval for the implementation of this PCM, as well as Supplements 1 6 2 to the same, is not required.

PCM 147-186 LUCIE UNIT I ICM OISCHARGE PIPf ZINC RIB80N (REA-SLN-SS-137-12)

~BSTRACT acka e covers the installation of zinc ribbon

'cial anodes in' the Intake Cooling M a t er { ICM) discharge piping.

he h anodes will be ins a lle d i n the pipe beginning at the Component Mater CCM) wall and extending to the discharge canal, Th e ovide cathodic protection for the internal surface Th PC/H i 1 if' t Related because the sacrificial anodes are to be installed f ty Related ICM pipe. The anodes perform no safety

.f t Pl t safety or operation and the installation does not constitute an unreviewed safety question or require a chan Q e to the plant Technical Specifications.

SAF VA'ATIO The sacrificial anodes to be installed in the ICW discharge pipe as described in this design package do not have a safety function. As demonstrated by the failure modes evaluation in the design analysis, the principle effect on safety is the potential for internal pipe coating damage in the event that a zinc ribbon pipe attachment fails. For this reason, guality Related design requirements have been applied and the modification is classified as guality Related.

Based upon the above and information supplied in the design analysis, it can be demonstrated, that an unreviewed safety question as defined by IOCFR50.59 does not exist.

i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety evaluated in the safety analysis report has not been increased because the zinc anodes are installed downstream of all active ICW System components evaluated in the FSAR.

Therefore, there is no interaction with the evaluated system components.

ii) The possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis report has not been created because, as demonstrated in the design analysis, the worst case failure of the zinc anode pipe attachments would have no impact on the ICW System capability to perform its design functions as specified in FSAR Section 9,2.1. In addition, the pipe attachments (thermit welds) have been evaluated and the determination has been made that the bonding process will not cause detrimental metallurgical conditions or impact the pipe coating systems.

iii) The margin of safety as defined in the basis for any Technical Specification has not been reduced. The installation of the zinc anodes will have no impact on the structural integrity of the ICW system piping or the design flow requirements of the system. For this reason, it is concluded that the margin of safety has not been decreased.

IOCFR50.59 allows changes to a facility as described in the if an unreviewed safety question does not exist and if FSAR a change to the Technical Specifications is not required. As shown in the preceding sections, the change proposed by this design package does not involve an unreviewed safety question because each concern posed by IOCFR50.59 that pertains to an unreviewed safety question can be positively answered.

In conclusion, the change proposed in this design package is acceptable from the standpoint of nuclear safety, does not involve an unreviewed safety question, does not require a change to the Technical Specifications and does not require prior NRC approval.

PCM 013-187 FLORIDA POWER 6 LIGHT COMPANY ST LUCIE PLANT UNIT NO 1 SIMULATOR TRAINING FACILITY GAI-TRONICS REA-NONE ABSTRACT This Engineering Package (EP) includes modifications to provide Gai-Tronics communication capability, including emergency alarms and instructions, for the Simulator Training Facility. The new Gai-Tronics equipment will be tied into the existing St Lucie Unit 1 Gai-Tronics System at the Service Building.

The modifications presented by this Engineering Package impact only non"safety related equipment. However, two conduit supports are being added to a safety related block wall. The additional loading has been reviewed and determined to have no effect on the structural integrity of the wall. The Gai-Tronics modifications are required in order to assure compliance with the St Lucie Plant Emergency Plan. Also, the power supply for the Gai-Tronics System is supplied from a vital AC source. Therefore, this package is classified as Quality Related.

A review of the changes to be implemented by this PC/M was performed against the requirements of 10CFR50. 59 as indicated in Section 3.0 of this EP. As a result, the expansion of the Gai-Tronics System to include the Simulator Training Facility does not constitute an unreviewed safety question and will not affect plant safety and its operation. The implementation of this PCM does not require a change to the Plant Technical Specifications. Therefore, prior commission approval is not required for implementation of this EP.

C Safet Evaluation PCM 013-187 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 expand the St Lucie Unit 1 Gai-Tronics Communication System to include the Simulator Training Facility. Gai-Tronics speakers will be located throughout the Simulator Training Facility in order to assure complete coverage for the emergency alarm signals (e.g., site evacuation alarm) One handset/speaker amplifier will be

~

located in the Simulator Facility to provide twomay Gai-Tronics communication capability-Based upon the expansion of the Gai-Tronics System presented by this EP, the breaker for circuit 33 of non-safety related Vital AC Bus No 1 (the power feed for the Gai-Tronics System) will be increased from a 20 amp to a 30 amp breaker. Also, the feeder cable from Vital AC Bus No 1 to the Gai-Tronics Power Distribution Cabinet (Cable 11201F) will be changed from 1-2/C f10 to 2-1/C f4. All other supplemental equipment (i.e., the 70 amp fuse, isolation transformer and feeder cables 11201Y and Z associated with the power supply) remain unchanged. The additional load presented by this modification (3.6 amps maximum at 120V AC) is considered insignificant and will have no impact on loading. Also, the increase in breaker size will not affect circuit breaker coordination. A fault on Vital AC Bus No 1, circuit 33 will not result in the loss of the entire vital bus (i.e.f the circuit breaker and/or fuse for circuit 33 will clear the fault before any upstream breaker opens).

Based upon the feeder cable changeout, a new non-safety related conduit is required in the RAB from Vital AC Bus No 1 to cable tray C3. Two new conduit supports are added to seismically designed block wall 4167. The additional loading has been reviewed and concluded that'neither the stress levels in the wall nor its it has been fundamental natural frequency are significantly affected by this modification.

The Gai-Tronics System and the 120V Vital AC System are not safety related systems. The expansion of the Gai-Tronics System to include the Simulator Training Facility has no impact on any other plant systems or operations.

The Gai-Tronics System is not required to mitigate or monitor any result of an accident.

Failure of this system has no impact on previously generated safety analysis reports. The margin of safety as defined in the bases for any Technical Specification is not reduced ~

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

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

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PCM 016-187 CCWi TCW AND OBCW VALVE ACTUATOR REPLACEMENT ST LUCIE PLANT UNIT NO 1 REA-SLN-86&31 ABSTRACT This Engineering Package (EP) is to replace two (2) existing Bettis actuators, including all accessories, on the shutdown cooling heat exchanger isolation valves, I-HCV-14-3A and 3B. These actuators open the valves on the safety injection actuation signal (SIAS) and supply cooling water to the heat exchanger. During the outage of 1985, the actuators were inspected and found to have cylinder wear. The existing actuators, model 746-X-2SR-42, are no longer manufactured and spare parts are not available, therefore they are. being replaced with Bettis actuators, model NT312-SR4&3. The new actuators operate in the same manner and will perform the same function as the existing actuators.

The modification considered in this EP is on the Component Cooling Water System. The valves and actuators are Class 3 Seismic Category I>

therefore this EP is classified as Safety Related.

Design details are provided for the installation of the new actuators and all accessories on the exifting valves.

The safety evaluation has shown that this EP does not constitute an unreviewed safety question and implementation of this EP does not require a change to the Technical Specification. Therefore, prior NRC approval is not required for implementation.

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

PCM 016-187 SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59> a proposed change shall be deemed to involve aa unrevieved safety question: (i) if the probability of occurrence ox the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be or (ii) if the possibility for an accident or malfunction 'ncreased; of a different type than any evaluated previously in the Safety Analysis Repoxt may be created; or (iii) if the,margin of safety as defined in the basis for any technical specification is reduced.

The modification considered in this EP is classified as safety related because the shutdown cooling heat exchanger isolation valves, I-HCV".14-3A and 3B, and actuatoxs are Safety Class 3, Seismic Category I. In the modification the tvo (2) existing Bettis actuators, model 746A-X-2SR-42 vill be replaced with nev actuatore, model NT312"SR4%3, because the existing actuators are no longer manufactured and spare parts are not available. The actuatore open the isolation valves in the event of aa, SIAS. The nev actuators vill perform the same function in the same manner as the existing actuators. No new failure modes are created. On lose of power or loss of air the valves "fail open".

The modifications included in this Engineering Package do not iavolve an unreviewed safety questioa because:

i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated are not increased because the nev actuators vill perform the same function in the same manner as the existing actuators.

ii) There is no possibility for aa accident or malfunctioa of a different type than any previously evaluated. This EP does not modify the intended opexation or test requirements of the system because the nev actuators will perform the same function in the same manner as the existing actuators.

iii) This modification does not change the margin of safety as defined in the basis for any Technical Specification because it neither changes the design parameters of the CCW system nor does it change the CCW design flov or functional requirements.

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 vhich provides the bases that this change does not involve an unreviewed safety question therefore, prior NRC approval for the implementatioa of this PCM is not required.

PCM 01 8-1 8 7 SY LUCK UNrr I Drain for Pipe Line 2-VM-DOO (RE A-SLN-86-86)

ABSTRACT This Engineering Package (EP) is for preparation and issuance of necessary changes in documentation to reflect permanent installation of a drain connection/tap installed in the spool piece downstream of FIT 6608 in line 2-WM-D00 by Circuit Alteration Tag No. 2025. FIT 6608 is located in a vertical run of piping in the Waste Management System off gas header in the RAB. The off gas header must be periodically drained due to condensation in the piping. FIT 6608 acts like a check valve and prevents water from going to a low point drain. The drain connection/tap located in the spool piece allows water above (downstream) FIT 6608 to be drained out of the header.

The existing piping at the drain connection/tap location is non-seismic, Ouality Group D, performs no safety related function, has no affect on safety related equipment, has no affect on plant safety and operation, and the gas flowing through the pipe is acceptable for discharge to the environment. But, since the gas and condensation in the piping has the potential to be radioactive, the EP is classified as quality related.

Based on a failure mode analysis and 10 CFR 50.59 review, the change proposed by this EP is acceptable from the standpoint of nuclear safety, it does not involve an unreviewed safety question, and does not require any changes to Technical Specifications. Therefore, prior NRC approval is not required for implementation of the modification.

pCN 018-187 SAFETY EVhLUATION This EP is for preparation and issuance of necessary changes in documentation to reflect permanent installation of a drain connection/tap installed in the spool piece downstream of FIT 6608 in line 2-WM-D00 by Circuit Alteration Tag No.

2029. FIT 6608 is located in a vertical run of piping in. the Waste Management System off gas header. The off gas header must'be periodically drained due to condensation in the piping. FIT 6608 acts like a check valve and prevents water from going to a low point drain. The drain connection/tap located in the spoolpiece allows water above (downstream) FIT 6608 to be drained out of the header.

The existing piping at the drain connection/tap location is non-seismic, Quality Group D, performs no safety related function, has no affect on safety related equipment or functions, and the gas flowing through the pipe is acceptable for discharge to the environment. But, since the gas and condensation in the piping has the potential to be radioactive, the EP is classified as quality related.

Based on the above and the 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.

The drain connection/tap and associated piping are not used in any safety analysis for accidents or malfunction of equipment. This modification is non-nuclear safety related and will have no effect on equipment vital to plant safety. Based on this, the probability of occurrence or the consequences of all analyzed accidents remain 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.

This modification is non-nuclear safety related and based on the failure modes analysis will have no effect on safety related equipment and functions.

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

No function of the subject drain and associated piping is controlled by or in the basis for, any Technical Specification. Thus, the margin of safety as defined in the basis for any Technical Specification has not been reduced.

10 CFR 50.59 allows changes to a facility as described in the FSAR if an unreviewed safety question does not exist and if a change to the Technical Specification is not required. As shown in the preceding sections, the change proposed by this design package does not involve an unreviewed safety question because each concern posed by 10 CFR 50.59 that pertains to an unreviewed safety question can be positively answered. Also, no change to the Technical Specifications is required based on the above evaluation.

In conclusion, the change proposed in this design package is acceptable from the standpoint of nuclear safety, does not involve an unreviewed safety question, and does not require any change to Technical Specifications. Therefore, prior NRC approval is not required for implementation of the modifications.

pcs 034-187 ST. Mcm mtre 1 COHDEHSER OQTIZT TUBE SHEET PHD QhTERBOX COhTIHGS hBSTRhCT This engineering package addresses the addition of an epoxy coating to the condenser outlet tube sheets and vaterboxes. This modification vill enhance the corrosion resistance of the tube sheets and waterboxes and allov reduction of the cathodic protection system potentials and current densities.

The condensers and the plant circulating water system are classified as non-nuclear safety related and therefore, the modification addressed in this engineering package does not consistute an unreviewed safety question.

Furthermore, the addition of a protective coating to the condenser outlet tube sheets and waterboxes does not require a change to the plant Technical Specifications.

mm "

This supplement consists of the correction of the drawing number listed under Section 11.2 nf this Engineering Package and the correction of a typographical error in the abstract. These changes do not affect the original design bases and do not alter the conclusions of the original design analysis or safety evaluation.

PCM 034-187 As noted in FSAR Sections 9.2.3 and 10.4.5, the condensers and circulating water system perform no nuclear safety related function. A failure mode evaluation of the proposed condenser outlet tube sheet and waterbox coatings has determined there is no potential for interaction with equipment or functions important to nuclear safety. Accordingly, the modification addressed by this engineering package is classified as non nuclear safety related.

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

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 there is no potential for interaction between the modification addressed by this engineering package and equipment of functions important to safety, previous safety analysis report evaluations related to safety remain unaffected.

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

No new accidents or malfunctions associated with the failure of the condenser outlet tube sheet and waterbox coatings have been created.

The margin of safety as defined in the basis for any Technical Specification has not been r"educed.

Since there is no potential for interaction between the modification addressed by this engineering package and equipment or functions important to safety, the margin of safety as defined in any Technical Specification remains unaffected.

ln conclusion, the modification proposed in this engineering package is acceptable from the standpoint of nuclear safety, does not involve an unreviewed safety question and does llnot require a change to any Technical Specifications. Accordingly, N RC approval prior to implementation is not required.

i PCM 036-187 ST+ LUCIK UNIT Condenser Tubing Strain Sage Installation 8QQXBBt X As part of the investigation into titanium condenser tube hydriding (hydrogen embrittlement) which has been discovered at the St. Lucie Nuclear Plant, strain gage instrumentation will be used to measure actual tube strain following the unit's return to power operation. after the present refueling outage. Data on actual tube strain levels during full. power operation is required in order to develop "realistic" criteria for future tube plugging which may be required due to hydriding.

This design package provides the engineering necessary to'install a 1 1/2 " diameter penetration into the condenser steam space to allow for routing of strain gage wiring. Also provided 're guidelines for installing the strain gages and the lead wiring in the condenser and through the new penetration. Following testing, the strain gage lead wiring is to be cut> and the new penetration is to be capped and all joints are.to be seal welded.

During the ne~t refueling outage, the wiring and "piping" conduit are to be removed from inside the condenser.

This design package is classified as "Non-Nuclear Safety Related" since it affects only nonseismic, Quality Group D piping and structures in Non-Nuclear Safety Related systems.

Based on .the failure modes analysis and 10 CFR 50 59 review, this modification does not impact any safety'el'ated equipment and is not relied upon for any accident prevention or mitigation. Thus it does not constitute an unreviewed safety question.

there are no unreviewed safety questions, and since no changes to Since technical specifications are involved, this PC/N may be implemented without prior NRC approval.

PCN 036-187 SAFETY EVALUAITON n t 1 Conden ser is a Non-Nuclear Safety Related component and as such is Th e Ui not required to function duriag any existing analyzed accident scenarios.

Therefore, modifications to the condenser affects only Non-Nuclear Safety Related, Quality Group D equipnent.

The added penetration will meet all design criteria of existing penetrations to insure that the condenser pressure boundary is maintained.

Postulated failures of the materials would have no impact on safe shutdown of the plant, or safety related systems. Any materials involved in this modification which could be postulated to become dislodged would be caught in the condenser hotwell pump screen None of the materials are large enough to impact pump suction. Additionally, postulated failures of the condenser would have no impact on safe shutdown of the plant, or safety related systems. The condenser is not used to prevent postulated accidents, mitigate the consequences of such accidents, maintain safe shutdown conditions, or adequately store spent fuel.

The following statements demonstrate that an unreviewed safety question, as defined by 10 CFR 50.59, does not exist: .

• 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.

Failure of the condenser is not considered as an accident initiating event or considered in determining the probability of an accident.

Also, since this design change does not alter or affect equipnent used to mitigate accidents, the probability of malfunction of equipment important to safety remains unchanged.

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

There is no new failure mode introduced by this change that has not been evaluated previously in the FSAR. Additionally, no failure modes analyzed by the FSAR are affected by this design.

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

This change has no effect on any existing Technical Specifications and does not require any changes to the Technical Specifications.

Since no unreviewed safety questions have been determined to exist, and since no revisions to the Technical Specifications are required, NRC approval is not required prior to implementation.

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p~ 039-187 CONDENSATE RECIRCULATION TO CONDENSER SQUARE ROOT EXTRACTOR REPLACEMENT ST LUCIE PLANT - UNIT NO. I REA-SLN6-OI I ABSTRACT This Engineering Package covers the replacement of one (I) square root extractor.

The presently installed square root extractor is no longer being manufactured and a suitable replacement is being provided for maintenance reasons. This Engineering Design Package is considered quality related since the replacement device is an Integral part of the condensate recirculation system and a direct replacement for previously approved instrument. The instrumentation loop, of which this device is part of, is not used to mitigate incidents.and accidents and, therefore, this PC/M ls neo considered to be safety related.

h review of the changes to be implemented by this PCM was performed against the requirements of 10CFR 50.59. As indicated in Section 3.0 of this PCM, this PCM 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 t}w implementation of this PCM.

SAFETY EVALUATION The changing out of the Square Root Extractor in this PC/M does not involve an unreviewed safety question \

because:

This EP reflects no interference with the safety equipment in that they are not required for a safe reactor shut-down and could not be used to mitigate an accident. The square rpot extractors are non-safety related. This modification will have no effect on equipment performing any safety function. There is no possibility for the creation oi an accident or malfunction. In the event of a total failure of this square-root extractor, it will have no effect upon any safety related equipement.

The probability of occurrence of the consequences of an accident. or malfunction of equipment important to safety previously evaluated is neither increased nor occurs since this system is non-safety related. This modification will have no effect on equipment performing any safety function.

This system and/or component parts are not used in any accident scenario and there is no possibility for creating an accident or malfunction of a different type than any evaluated previously in the safety report. Its failure will have no impact on the plant safe shut-down.

It has no effect upon the margin of safety as defined in the basis for any technical specification since the replacement of the square root extractor does not change the original design or operation and the proposed new extractor's are functionally identical to existing units. There are no changes to the plant technical specifications.

The foregoing constitutes, per 10CFR 50.59, the written safety evluation which provides the basis that this change does not involve an unreviewed safety question. Therefore, prior commission approval is not required for implementation of this PC/M.

PCM 041-187 T UCIE UNIT NO MFRV POSITION INDIChTORS REMOVhL REh-SLN-85-043-10 BSTRhCT This engineering package covers the removal of two Main Feedwater Regulating Valve position indicators (ZI-9011,9021) from RTG Board 102 along with associated wiring, cable, and conduit. h steel plate will be fastened to the control board to cover the exposed area.

Since . these indicators are operationally unreliable, the potential exists for incorrect interpretation of regulating valve position. Removal of the indicators will accomplish the resolution of a Human Factors Discrepancy (HED). No modifications to the valve control circuitry will be performed. Hence, routine valve operations will continue to be controlled from signals received automatically via the Feedwater Regulating System.

Therefore, this 'modification will not have any adverse effect upon plant safety or operation.

There are neither any Technical Specification nor Regulatory Guide 1.97 requirements for these devices.

Since this design requires a modification to the RTG board, Quality Related requirements shall be imposed.

/

These changes were reviewed against the requirements of 10CFR50.59. hs verified in the Safety Evaluation, this change neither requires a Technical Specification revision nor is it an unreviewed safety question. Therefore, prior NRC approval is not required.

PCN 041-187 ShFHTY HVhLUhTION This BP is classifie d as Quality Related because the components being removed, while performing a Non-Nuclear Safety Related function, are installed in the RTQ Board where the potential exists for impacting Safety Related equipment through modification of the wiring in the RTG Board, the removal of equipment, and the installation of cover plates that could potentially have an effect on the seismic integrity of the RTG Board.

This design proposes to remove the Main Peedwater Regulating Valve (MFRV) position indicators currently installed in the RTG Board 102.

The indicators are unreliable and could provide misleading valve position indication. Removal of the indicators will not affect the operator's ability to determine feedwater flow or steam generator level. hmple instrumentation is available to monitor these parameters from the control room. In addition, indicating lights in the control room will remain to determine whether the subject flow control valves are fully open or fully closed.

The indicators being removed do not perform a Nuclear Safety Related function and are not included under any Technical Specification or Regulatory Guide 1.97 requirement.

Internal wiring changes are being performed in the RTG Board to disconnect the subject indicators and to remove (SIS) wiring.

When required, only qualified wire jumpers will be installed inside the RTG Board. No conduit is being removed adjacent to, or in the vicinity of the RTQ Board or control room.

The restoration of the RTG Board through appropriate cover plates to replace the removed indicators has been evaluated within this package. This evaluation concluded both that the seismic integrity of the RTQ Board will be retained and that no missiles could be generated during a seismic event which could adversely impact Safety Related equipment.

PCM 054-187 ST LUCIE PLANT - UNIT NO 1 CONDENSATE POLISHER TIE-INS REA-SLN-85-14 ABSTRACT Thiss En gne ineering Package (EP) is for the installation of the 24 inch Unit 1 tie-in piping and manual isolation valves require uired for the future connection of the Condensate Polisher System (

t 2 C on d casa e System. It also includes the installati,on of the 8 inch tie-in piping that will connect the CPS backwash pump suc This it on ensa e Storage Tank non-safety class connection.

existing Un 2 Condensate is for providing the capability of using the Unit 2 c ondensate for backwashing the condensate polisher.

This EP is classified quality related since it also involves modifications to t hee RTGB-102 which is seismically qualified - and located in te h Un it 1 Control on Room. The modifications to the RTGB-1 02 involve nvo ve tea h i f the misalignment.

ddi tono tie-in isolation valve position indicat ng g hts in li and alarm for valve The safety evaluation has determined that the EP does not constitute an unreviewed safety question and implementation of the EP does not require a change to the Plant Technical Specification. 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 change, shal1 be deemed to involve an unreviewed safety question; (i) if the probability of occurrence or the consequences of an acc id en t o r m alfunction of equipment important to safety previously evvaluated ua in the safety analysis repoxt may be increased; or ((ii))

~

if a ssibility for an accident or malfunction of a different type than a ny evaluated previously in the safety analysis report may be created; or if (iii) the margin of safety as defined'n the bases for any Technical Specification is reduced.

This Engineering Package (EP) is for the installation of the 24 inch condensate Unit 1 tie-in piping required for the future connection of the Condensate Polisher System (CPS) to the Unit 2 Condensate System.

It also includes the installation of an 8 inch connection to the condensate polisher backwash pump suction from an existing non-safety class connection to the Unit 2 condensate storage tank. This is for providing the capability of using, in the future, Unit 2 condensate for backwashing the 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; however, since this EP also involves modifications to the RTGB, which is seismically qualified, for the addition of an annunciator and indicating lights for the condensate polisher isolation valves, it is classified quality related.

p~ 054-187 SAPETY EVALUATION (CONTINUED}

The new annunciator and indicating lights that will be added to the RTGB have been designed and will be installed to the same requirements as existing annunciators and indicating lights in the RTGB. This addition of components to the RTGB has been reviewed and considered acceptable and in compl1ance with the seismic requirements applicable to the RTGB.

Based on the above description, the modification included in this Engineer1ng Package (EP) 1s considered to be quality related. This .EP does not involve an unreviewed safety question, and the following are bases for this justification-(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated 1n the safety analysis report is not increased. The portions of the Condensate System, where this modification will be implemented, and the CPS are not used 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. The addit1on of the new annunciator and indicatIng lights to the RTGB has been reviewed and considered to be acceptable and in compliance with the seismic requirements applicable to the RTGB.

The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report 1s not created. The components involved in th1s modificat1on have no safety related function and no changes have been made to the operational design of the system.

(111) The margin of safety as defined in the bases for any Techn1cal .

Specification is not affected by this PCM, 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'ases that this change does not involve an unreviewed safety question and prior Commisision approval for the implementation of this PCM is not required.

PCH 075-l87 ST LUCIE UNIT NO 1 FIRE DETECTOR MODIFICATIONS REA-SLN-86-63-10 ABSTRACT This Engineering Package covers the modifications to the fire detection system which is part of the Fire Protection System.

This Engineering Package will provide the engineering and design details required to implement the replacement of the existing ionization smoke detectors. The existing detectors are divided into two (2) groups: The originals (installed eleven (ll) years ago) which are obsolete; and their replacements (installed as the originals failed) which are no longer manufactured. To ensure the reliability of the fire detection system, new state of the art ionization smoke detectors will be installed.

The Fire Detection System is non-safety related, but is provided in areas that contain or present a fire exposure to equipment essential to safe plant shutdown. Therefore, this Engineering Package is classified as Quality Related.

The safety evaluation has determined that this modification does not involve an unreviewed safety .question and implementation of this PCM does not require a change to Plant Technical Specifications. Therefore, prior NRC approval for the implementation of this PCM is not required.

This EP has no impact on the plant safety and operation.

SAFETY EVALUATION kith 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 important to consequences of an accident or malfunction of equipment 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 ionization smoke detectors with new detectors and new wiring bases. The existing detectors are either obsolete or no longer manufactured.

The implementation of this Engineering Package ensures the availability of the individual detectors to detect a fire.

PCM 075-187 SAFETY EVALUATION (CONTINUED)

Fire detection systems are provided in areas that contain or present a fire exposure to equipment essential to safe plant shutdown.

Therefore, this Engineering Package has been classified as Quality Related.

Based on the preceding, this EP does not involve an unreviewed safety question and the following are the bases for this 5ustification:

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 ionization smoke detectors enhances the operability of the equipment. The replacement of the obsolete detectors with new detectors is on a one to one basis, with the new detectors having the same characteristics as the existing detectors. The possible failure of the detectors will not prevent safety related equipment from performing their intended functions. The detectors are not required during an accident condition.

(ii) As a result of this modification, there is no possibility for an accident or malfunction of a different type than any previously evaluated. The detectors are not required during an accident condition nor will they prevent safety related equipment from performing their functions. The existing detectors are being replaced on a one to .one basis. This modification does not affect any safety related equipment.

(iii) This modification does not reduce the margin of safety as defined in the bases for any Technical Specification. The functions of the Fire Detection System that are controlled by the applicable Technical Specifications are maintained by this change.

The implementation of this PCM 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.

PCM 076-187 ST LUCIE PLANT UNIT 1 ERMDS/SAS UPGRADE REA-NONE 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 ie also known ae the Safety Assessment SYstem (ERDADS/SAS), under PCM 076-187 Supplement 1 and PCM 077-287. Included in this work are the connection of the computer room to the adgoining office, relocation of computer room and office doors, installation of a false floor in the office, upgrade of lighting and convenience oulete, and inetaH.ation of conduit and cables for the computer control terminals, the data loading terminal, CRT 812 console, and disk drives.

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

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

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

SUPPLEMENT 1 In addition to modifying the computer room, this Engineering Package provides for an upgrade to the ERDADS/SAS hardware and software including the Safety Parameter Display System (SPDS), in the St Lucie - Unit 1 control zoom, computer room and technical support center. It vill improve the performance and display capabilities of the existing system and will include new display CRTs and keyboards, new color hazdcopiers, additional printers, a data loading terminal, additional memory and new internal computer switching and communications componentsi 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 RP will not affect the safety or operation of the plant.

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076-187 SUPPLEMENT 2 Supplement 2 to this Engineering Package modifies the design to replace the CRTs, video generators, and supporting components which were original11 y 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 unrertewed-safwt~uestion 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.

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 unx'eviewed safety question: (i) if the probability of occuxrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be incx'eased; 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 fox any technical specification is reduced.

076-187 SAFETY EVALUATION (continued)

The modifications included in this Engineering Package do not involve an unrev3.ewect'afety question because:

i) 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.

ii) 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 Attachment 7.1). 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.

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 Specif ications.

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.

PCN 078-187 ST LUCIE PLANT - UNIT 1 REPLACEMENT OP PISCHER AND PORTER CONTROLLERS REA-SLN&6-91-10 This Engineering Pac'kage (EP) covers the replacement of the now obsolete Pischer 6 Porter controllera with the currently manufactured and functionally equivalent Pischer 6 Porter controllers. The controllers are used to maintain the level and pressure parameters in the pressurizer within the required limits during the normal plant operation.

These controllers perform Non-Nuclear Safety Related functions. However, being located on the main control board, they are expected to maintain their structural integrity during the design basis seismic event. The controllers are classified Quality Related.

The safety evaluation (Section 3.0) indicates that this Engineering Package does not involve an unreviewod safety question, and does not require a change in the Plant Technical Specifications. Therefore, NRC approval for these modifications, prior to their implementation, is not required.

This EP has no impact on plant safety or op'eration.

PCM 078-187 SAPETY 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 malfunction probability of of occurrence or the consequences of an important to safety previously accident or equipment evaluated in the Safety Analysis Report may be increased, or (ii) if a possibility for an accident or malfuntion of a different type than any evaluated previously in the Safety Analysis Report may be created, or (iii) if the margin of safety as defined in the bases for any technical specification is reduced.

The modifications included in this Engineering Package do not involve an unreviewed safety question because of the following reasons'.

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

The possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report will not be created by this modification because the function of the controllers has not been altered by this modification.

(iii) The margin of safety as defined in the bases for any technical specification is not reduced since the new controllers perform non-nuclear safety related functions and are not included in the bases of any technical specification.

The new controllers replace the obsolete controllers on Class 1E main control board, therefore, this EP is classified Quality Related.

The implementation of this EP does not require a change to the Plant Technical Specifications, nor does Therefore, the it create an unrevtewed safety be implemented without prior question. PCM may Commission approval.

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.

ST LUCIE UNIT 1 TURBINE GENERATOR SEAL OIL SYSTEM ENHANCEMENTS

@GAEA: SLN-86-092-10)

ABSTRACT This Engineering Package covers modifications to the Turbine Generator Seal Oil System as recommended in V/estinghouse Operations and Maintenance Memo 005l {Reference 6.3). This modification provides for the installation of a "drip leg" in the air side seal oil pump suction line and an additional vent line between the existing vent line and the hydrogen side drain regulator tank. These system enhancements should minimize oil intrusion into the generator housing, and decrease the amount of dirt and contamination that would lead to damage/wear to system components.

The Turbine Generator Seal Oil System performs no safety related functions nor does it interact with safety related equipment. Therefore, this modification is classified as non-nuclear safety related.

Based on a failure mode evaluation and a 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. This modification has no adverse effect on plant safety or operability.

PCM 085-187 SAFETY EVALUATION This Engineeri'ng Package covers the modifications to the Turbine Generator Seal Oil System. A "drip leg" will be installed in the air side seal oil pump suction line. Also, an additional vent line will be installed between the existing vent line and the hydrogen side drain regulator tank.

This modification is classified as non-nuclear safety related, since the Seal Oil System performs no safety related function and does not interact with safety related equipment, components, or functions.

Based on the above and information supplied in the design analysis, it can be demonstrated that an unrev.'ed 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.

Due to the location of the "drip leg" and the vent line, their failure would not cause interaction with any safety related equipment. Also, the turbine generator seal oil system is not considered by the FSAR in determining the probability of accidents, possible types of accidents, or in the evaluation of consequences of accidents. Therefore, it can be concluded that the probability of occurrence of accidents previously addressed in the FSAR 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.

The components involved in this modification do not perform safety related functions. The operability of the turbine generator seal oil system has not been adversely affected by the modification. Also, the location of the "drip leg" and vent line eliminates the possibility of interaction with safety related equipment. Therefore, the possibility of an accident of a different type has not been created.

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

Since the components involved in this modification are not directly included in the bases of any technical specification, the margin of safety has not been reduced.

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. As shown in the preceeding sections, the change proposed by this design package does not involve an unreviewed safety question because each concern posed by 10 CFR 50.59 that pertains to an unreviewed safety question can be positively answered. Also, no change to the Technical Specifications is required based on the above evaluation. Therefore, prior NRC approval is not required for implementation of this modification.

PCM 088-187 EQMOTE REACTOR VESSEL LEVEL INDICATOR ST LUCIE PLANT -

REA-SLN-8706 UNIT NO l ABSTRACT This Engineering Package (EP) is to install a remote level 1ndicator for the reactor vessel. this indicator will prov1de reliable level indication during refueling.

The modifications considered in this EP are on the Reactor Coolant System. The connections are designated as Nuclear Safety Related and seismically qualified since they are within the Reactor Coolant Pressure Boundary, and therefore, this modification is classified as Safety Related. The instrument side of the system downstream of the piping isolation valve is designated as non-safety, seismic design.

Two transmitters (one wide range, one narrow range) and associated cables will be installed. Indication will be added to the Control Room to allow monitoring of refueling water level. The safety evaluation has shown that this EP does not constitute an unrev1ewed safety question and pr1or NRC approval is not required for implementation.

The implementation of this EP does not require a change to the Technical Specification and does not reduce the margin of safety for any Technical Specification.

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

Supplement No l The purpose of this supplement is to remove all hold points associated with th1s EP. The reactor coolant piping supports and the conduit supports within the containment area have been evaluated, so the hold points are no longer necessary.

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

PCM 088-187 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; if (i) the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.

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

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

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 this modification provides a means whereby an accurate Reactor Vessel water level can be readily determined during refueling. During power operation this system is isolated from the RCS. 3he portions of this modification within the normal RCS pressure boundary have been designed to the original requirements of the RCS pressure boundary.

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

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

The implementation of this P(M does not require a change to the plant Technical Specifiqation.

The foregoing constitutes, per 10CFR 50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question or a change to the Technical Specifications, and prior commission approval for the implementation of this PCM is not required.

pCN 105-187 ST LUCIE UNIT 1 CHARGING PUMP BLOCK MATERIAL CHANGE E SLN -9 ABSTRACT This design package covers the replacement of the current charging pump block material" of 316 stainless steel (ASTM-A-182 F316) with 17-4 PH stainless steel (ASME-SA-705 Gr. 630 1150 HT). The 17-4 PH material has a tested fatigue strength approximately twice that of 304 .or 316 stainless steel. Field testing of charging pump systems indicate that strong pressure pulsations exist. at times in the system. These pulsations are in part responsible for the fatigue failures of the charging pump blocks. Reduction of pressure pulsations is a current concern. The increase in fatigue strength of the new material should result in a substantial improvement in block life. Based on a failure mode analysis and 10CFR50.59 review, the changes proposed by this engineering package are acceptable from the standpoint of Nuclear Safety. This modification does not involve an unreviewed safety question and a Technical Specification change is not required, therefore, prior NRC approval is not required for implementation of this modification. The function of the charging pumps is not altered by this modification. This engineering package is classified as Nuclear Safety Related.

PCM 105-187-S ET Y VALUA 0 This modification consists of replacement of the existing 316 stainless steel charging pump blocks with 17-4 PH 1150 HT stainless steel. This modification does not affect the design function of the charging pumps and does not introduce any new active components to the system.

The new material is stronger than the existing material and should provide a substantially longer service life for the block. Since the system and components modified by this engineering package are ASME Section III, Class II, this package is classified as Nuclear Safety Related.

The following constitutes an evaluation to determine implementation of this engineering package will result in an if the unreviewed safety question as defined by 10CFR50.59:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis is not increased since no new active components are being added, and the failure modes of existing components are not being altered.

Accident probabilities and consequences are not affected by this modification.

The probability of an accident or malfunction of a different

. type than created.

previously evaluated in the FSAR has not been Since the system design bases as described in FSAR Sections 9.3.4.3.2 (f) are not affected by this modification, no new accidents are made possible.

The margin of safety as defined in the basis for any technical specification has not been reduced since no system design parameters are being altered. The technical specifications have been reviewed and that no changes are required.

it has been determined In conclusion, the change proposed in this design package is acceptable from the standpoint of nuclear safety, does not involve an unreviewed safety question, and does not require any change to technical specifications. Therefore, prior NRC approval is not required for implementation of the modifications.

ST LUCIE UNIT NO 1 REPLACEMENT OF SAFETY RELATED BATTERIES 1A and 1B (REA-SIN-87-008-11)

ABSTRACT This Engineering Package covers the modifications to the Safety Related Station Batteries IA and 1B which are part of the 125V DC Distribution System.

This Engineering Package will provide the engineering and design details required to implement the replacement of the existing batteries with new batteries. The existing batteries are showing signs of degradation (the battery acid is contacting the copper poets). The new batteries will also have an increased spare design margin (capacity) of 3Z over the existing batteries, which were Installed in the early 80s, for future load growth capability.

The station batteries, which are part of the 125V DC system, are classified as Class lE, are seismically qualified and perform a safety related function.

This EP will be classified as Safety 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 related equipment The implementation of this PCM does not require a change to the Plant Technical Specifications.

This change does not involve an unreviewed safety question and prior Commission approval for the implementation of this PCM is not required.

SAFETY EVALUATION Vith 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; o'r (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 116-187 This Engineering Package provides the engineering and design detai1s required to implement the replacement of the existing batteries with new batteries. The existing batteries are showing signs of degradation which cou1d reduce the capacity of the battery cells.

The implementation of this Engineering Package increases the availa-bility of the batteries, upon loss of the AC power system,chargers to prov1de power sufficient to supply the DC loads until the battery are loaded onto the diesel generators. The 125V DC systems, which include the station batteries, are safety xelated and complete separat1on and independence are maintained between equipment and cixcuits, including raceway. A single failure at any po1nt in eithex system will not disable both systems.

The station batteries which are being replaced perform a safety related function within the 125V DC distribution system and are designed for operation under conditions that could be imposed by a Design Basis Accident (DBA). This Engineering Package has been classified as Safety Related.

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

(1) The probability of occurrence or the consequence of an accident ox'alfunction of equipment important to safety previously evaluated in the safety analysis report is not increased, since the replacement of the station batteries enhances the opera-bility of the equipment The addition of new batteries ensures that the batter'ies will supply the minimum DC power requirements to safely shutdown the plant and/or mitigate the consequences of a DBA.

(ii) As a result of this modification, there is no poss1bility for an accident or malfunction of a different type than any pxeviously evaluated. This modification affects accident m1tigating equip-ment to enhance their operation. The DC system voltage remains the same but the new batteries provide an increased spare design margin (capacity) for future load growth. There is . no introduction of any new failure mode for the equipment.

(iii) This modification does not reduce the margin of safety as for any Technical Specification. The defined in the bases safety function that is controlled by the various applicable Technical Specifications is maintained by this change. The px'oposed design ensures that the batteries will funct1on as assumed dur1ng an accident.. Thus the margin of safety provided by the Technical Specifications is preserved.

The implementation of this PCM 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.

ST LUCIE UNIT 1 PCM 119-187 GROUTING OP MASONRY BLOCK WALLS REA SLN 87-061 In the course of preparing the Pire Protection Appendix of the Unit 1 FSAR, e concern was raised as to whether certain masonry block walls assumed to be 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire barriers are actually grout filled. A safety evaluation was performed (Reference 6.5) vhich established that, if these walls are i.n fact not filled with grout aad therefore not providing the full 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of fire protection, the plant bti11 maintains its ability to achieve safe shutdown.

Thi s fety evaluation recommended that an inspection of these veils be performed to establish their as-built condition. Such an inspection was performed and concluded that the walls are not fully grouted.

This Engineeriag Package (EP) provides the details/requirements for grouting the voids ia block walls 79, 84, 84A, 85, 92A, 114, 115, and 115A. This ti grou ng will be performed in two phases. 'A HOLD POINT is place construction activiti.es at the completioa of Phase I work. se on construction activities will resume following engineering approval of the Phase II groutiag material.

This modificatioa does not involve aa uarevieved safety question, has ao effect on plant safety and operation, and does aot involve a change to any plant Technical Specification. Upon completion of this modification, the action ia Technical Speci.fication 3/4.7.12 vill no longer be required for the wells modified. This EP is classified Quality Related since all of the veils involved are required per 10 CFR 50 Appendix R to be fire barriers.

SAFETY EVALUATION Safet Anal sis With respect to Title 10 of the Code of Federal. Regu1ations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question; (i) if the probability of occurrence consequeaces of an accident or malfunction of equipmeat important to or safety previously evaluated in the safety analysis report may be increased; if (ii) a possibility for an accident or malfuncti.on 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.

When a concern vas raised that the veils modified by this EP might not be fully grouted, a report (Reference 6.5) vas wri.tten to evaluate the safety implications if the walls vere found to be not fully grouted. This report demonstrated that, if an ungrouted condition was confirmed, no unreviewed safety questions exists end continued operation of the plant is )ustified. This EP provides the details/requirements for grouting the walls so that they are in conformance vith the design bases established in Subsection 3.11.2 of the St Lucie Unit 1 FSAR Appendix 9.5A; consequently, this modification cannot give rise to an unreviewed safety question.

Although the walls do not perform a safety-related function, this EP is classified Quality Related, since ell of the walls are required per 10 CFR 50 Appendix R to be fire barriers.

PCM 119-187 Based on the above, the following provides the Justification that an unreviewed safety question does not exist:

(i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased. - Since the ~alla located in the vicinity of safety"related equipment maintain their seismic qualification, no accidents due to structural failure are postulated. The only other type of accident potentially associated with the walls affected by this modification involves damage that could occur if the walls fail to provide three hours of fire protection. The JCO discussed above, however, demonstrated that no single fire event could impair the plant's ability to achieve safe shutdown. Consequently, there are no accidents or malfunctions of equipment important to safety previously evaluated whose probability of occurrence or consequences are increased by this modification.

Construction activities will stop when Phase I is completed.

Phase II construction will continue after Phase II materials have been reviewed and approved with respect to their density and their structural, radiation resistance, and thermal resistance properties, and the use of these materials has been shown not to degrade the seismic qualification of Walls 85 and 114. This item is identified as a HOLD POINT and must be resolved prior to the implementation of Phase II. The safety evaluation will be r'evised upon resolution of this item.

(ii) There is no possibility for an accident or malfunction of a different type than any evaluated previously since the modification provides the walls with a three hour fire rating while the design ensures that the structural integrity of the seismically designed walls is maintained.

(iii) This modification does not change the margin of safety as defined in the bases for any Technical Specification. The basis for Technical Specification 3/4.7.12 indicates that fire barriers ensure that fire damage will be limited and the possibility of a single fire event involving more than one fire area prior to detection and extinguishment will be minimized. The referenced JCO indicated that the current situation, in combination with compensatory measures, does not violate this basis. When the walls are fully grouted, the barriers will be fully operational, eliminating the need for the said compensatory measures.

The implementation of this PCM does not require a change to 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 Commission approval for the implementation of this PCM is not required.

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PCM 1 28-1 8 7 ST LUCIE PLANT - UNIT NO 1 SAFETY INJECZION TANK AND CONTAINMENT FAN COOLER INSTRUMENTATION UPGRADE REA-SLN-86&76-11, -13, -21, -23 ABSTRACT This Engineering Package addresses level, temperatux'e, and flow instrumentation upgrade for the Safety Infection Tank (SIT), Component Cooling Water and Contaiameat Fan Coolers.

The Safety In)ection Tanks are part of the Safety In)ection System which automatically discharges borated water into the Reactor Coolant System on depressurization of RCS as a result of a Loss of Coolant Acciden (

level instrumentation being upgraded measures the Safety In)ection Tank water level and provides indication at the RTGB.

The Containment Fen Coolers are part of the Containment Cooling System which provides the means of Containment heat removal during normal operations and in the eveat of a LOCA. The flow iastrumentatioa being upgraded detects low Component Cooling Water flow through the Containment Fan Coolers, indication and remote annunciation. The temperature detecting elements providiag'ocal (thermocouples) at the inlet and outlet'of the Containmeat Fan Coolers used to measure the duct air tempexature are also being upgraded.

These instruments currently are designated as Non"-Nuclear Safety Related.

This effort vill upgrade selected instxumentatioa, associated electrical circuit loops and structural support to Nuclear Safety Related meeting the requirements of USNRC Regulatory Guide 1.97, Rev 3, Category 2, Type D Variable.

This U S Nuclear Regulatory Commission requirement is defined as those instruments that x'emaia functional during all accident conditions and provide indication and records for many variables required to follow the course of the acc id en t . S pec ecifically Type D variables are defined as those variables that provide information to indicate the opexation of individual safety sy s stems and other systems important to safety. Category 2 provides for equipment qualification which is less stringent ia that it does not include seismic qualificatioa, redundancy or continuous displays aad requires only a high-reliability power source.

Based on the usage of these instruments to monitor safety related equipment, this EP is classified as Nuclear Safety Related.

The safety evaluation of this package has shown that the implementation of this PCM does not constitute an unreviewed safety question and prior commission appzoval for its implementation is aot required This EP has no impact on pleat safety end operation or Plant Technical Specifications.

SUPPLEMENT 1 This engineering package revision revises control wiring diagrams and cable splice details dealing with the revised Conax thermocouple electx'ic conductor seal assembly. Environmental Qualification Documentation Package 8770-A-451-6.0, Continental Wire and Cable, hae been updated to include references to model CC-2200 (XLPE) B/M D5-1.

The original safety evaluation hae not been affected as a result of .this supplement.

II SAFETY EVALUATION P~ 128-ZS7 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: (1) if the probability of occurrence or the consequences of aa accident or malfunction of equipment importaat to safety previously evaluated in the Safety Analysis Report may be increased, or (ii) if a possibility for an accident or malfuaction of a different type than any evaluated previously 1n the Safety Analysis Report may be created, or (iii) if the margin of safety as defined in the bases for any technical specification is reduced.

The modifications included in this Engineeriag Package do not involve an unreviewed safety question because of the following reasons:

(1) The probability of occurrence and the consequences of an accident or malfunction of equipmeat important to safety previously evaluated in the Safety Analysis Report vill not be increased by this modification because existing equipment availability, redundance, capacity, or function requ1red to mitigate the effects'of an accident are aot affected.

(ii) The poseibili.ty for an accident or malfunction of a different type than any evaluated previ.ously in the Safety Analysis Report will not be created by this modification because replacing moaitoring instrumentation with similar replacements having better environmental qualifications does not create changes which could postulate a aew accident or malfunction.

(iii) The margia of safety as defined in the bases for any technical speci,fication is not reduced since thi.s modi.fication installs qualified thermocouples and flow switches which will enhance the monitoring of the Containment Heat Removal System.

Furthermore~ this new equipment is sei.smically and environmentally qua11fied to withstand the normal and accident conditions anticipated ia the areas that they are installed.

This modificati.on is for the upgrade of the Safety Injection System, Component Cooling Water System aad Containment Cool1ng System instrumentation in order to meet the requirements of USNRC Regulatory Guide 1.97, Ryv 3, Category 2, Type D Variable. This modification upgrade will provide a more reliable and qualified instrumeatatioa loop to detect and monitor Containment Heat Removal System operation. Hence, th1e EP is considered Nuclear Safety Related.

Since this modif1catioa replaces existing monitoring instrumentation with quali.fied devices and involves no other modificatioas to safety related equipment, the degree of protect1oa provided to nuclear safety related equipment ie unchanged. The probability of malfunction of equipmhat important to safety previously evaluated in the FSAR remains unchanged. The consequences of malfunction of equ1pment important to safety previously evaluated in the FSAR are unchanged. The pose1bility of malfunctions of a different type than those analyzed in the FSAR ie not created.

The implementation of this EP does not require a change to the Plant Technical Specifications, nor does question.

it create an unreviewed safety The foregoing coastitutes, per 10CFR50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question aad prior Commi.ssion approval for the implementation of this PCH is not required.

PCM 1 41-1 8 7 ST LUCIE UNIT 1 480V SWITCHGEAR 1A2 & 1B2 TRANSFORMER REPLACEMENT (REA SLN"86-007-10)

ABSTRACT Due to environmental concerns attendant to polychlorinated biphenyl (PCB) cooliug/insulating liquids, all transformers filled with PCB are being eliminated from FP&L's system. The station service transformers for 480 volt switchgear 1A2 and lB2 are filled with PCB cooling/insulating oil. Each transformer contains 370 gallons and 254 gallons respectively of PCB liquid.

This Engineering Package provides for the replacement of the existing PCB filled station service transformers with equivalent transformers of dry type construction and for the removal of the concrete curbs surrounding the transformers. The curbs are no longer requi.red since their function was to retain leakage of cooling/insulating liquid which is no longer present in the replacement transformers.

Station service transformers 1A2 and 1B2 perform nuclear safety related functions. Because of their importance in Class lE service applications the replacement transformers are classified as safety related in this Engineering Package.

Transformers (1A2 & 1B2) are located in the Switchgear Room at Elevation 43'0" of the Reactor Auxiliary Building.

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

The implementation of this PC/M will not have an adverse impact on plant safety or operations-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 3m created; or (iii) if the margin df safety as defined in the bases for any technical specification is reduced.

This Engineering Package addressed the replacement of PCB liquid filled 480V station service transformers lA2 & 1B2 located on elevation 43'n the Reactor Auxiliary Building of Unit 1. The replacement transformers will be furnished dimensionally compatible and equivalent in electrical characteristics with the existing transformers.

The physical characteristics of the replacement transformers are different because they are dry type.

PCM 141-187 The new transformers are safety related because of their importance to essential plant operations. These new transformers perform the same function as the existing transformers 1A2 and 1B2. The replacement transformers have been seismically and environmentally qualified (References 6.18 and 6.19) and will be seismically mounted. The existing seismic qualification of switchgear lineups, 1A2 and 1B2 will not be affected by the replacement of the lA2 and 1B2 PCB filled transformers with the new dry type transformers.

The curbs do not perform any safety function. They were designed to contain cooling/insulating liquid which will no longer be used; therefore these curbs are no longer required.

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

(i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR will not be increased because the existing transformers are being replaced on a one-for-one basis by transformers that are essentially equivalent in function, capacity and qualifications. The curbs did not perform a safety related function. Their removal vill not have any safety related implications.

(ii) This modification does not change the operation of the 480V safety related station service transformers and switchgear.

Therefore, there is no possibility that an accident or malfunction of a different type than any evaluated in the FSAR may be created.

replacement station service transformers are essentially (iii) equivalent The in purpose and capability to the existing transformers. Therefore, this modI.fication does not reduce the margin of safety as defined in the bases for any technical specification.

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

PCM 142-187 ST LUCIE UNIT 1 480V SMITCHGEAR 1A3 6 1B3 TRANSPORMER REPLACEMENT (REA SLN-8607-10)

Due to environmental concerns attendant to polychlorinated biphenyl (PCB) cooling/insulating liquids, all transformers filled with PCB axe being eliminated from PP&L's system. The pressurizer heater transformers for 480 volt switchgeax 1A3 and 1B3 are filled with PCB cooling/insulating oil. Each transformer contains 208 gallons of PCB liquid. This Engineering Package provides fox the replacement of the existing PCB filled presaurizer heater transformers with equivalent transformers of dry type construction, and for the removal of the concrete curbs surrounding the transformers. The curbs are no longer required since theix function was to retain leakage of cooling/-

insulating liquid, which is no longer present.

Pressurizer heaters transformers IA3 and 1B3 perform non-nuclear safety related functions. Because of their importance in pLant opexations an and because they are fed from Safety Related buses, 4160V 1A3 and 1B3, the replacement transformers are classified as Quality Related in this Engineering Package.

Transformers (lA3 6 1B3) are located on Elevation 43'0" of the Reactor Auziliary Building.

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

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

SAFETV EVALUATION

~ith respect to Title 10 of the Code of Federal Regulations, Part S0.59> 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 ox malfunction of a different type than any evaluated pxeviously 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 addressed the xeplacement of PCB liquid filled 480V pressurizer heater transformers 1A3 6 1B3 located on elevation 43'n the Reactor Auxiliary Building of Unit 1. The transformers supply power to the pressurizer heaters and are located in an area of the plant containing safety-related equipment. The 480V Pressurizer Heater Transformers lA3 and 133 do not perform any nuclear safety related functions, however, because of their importance to normal plant operations and because transformers 1A3 and 1B3 are fed by safety related 4160V Buses 1A3 and 1B3, the replacement transformers are classified as Quality Related in this Engineering Package. The 'dry type'eplacement transformexs will be furnished dimensionally compatible and. equivalent in electrical characteristics with the existing transformers.

PCM 142-187 The physical characteristics of the replacement transform<

therefore these cuxbs are no longer required.

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

(1) The probability of occuxrence or the consequences of an accident ox'alfunction of equipment important to safety'reviously evaluated in the FSAR will not be increased because the existing txansformexs are being replaced on a one-for one basis by transformexs that are equivalent in function, capacity and electxical characteristics. The curbs did not perform a safety x'elated function, their removal will not have any safety related implication.

(ii) This modification does not change the operation of the 480V non-safety related pressurizer heater transformers and switchgear. Therefore, there is no possibility that an accident or malfunction of a different type than any evaluated in the RSAR may be czeated.

(111) The replacement pressurizer heater transformers aze equivalent in purpose and function to the existing txansformezs and perform no safety related functions. Therefore, this modification does not reduce the margin of safety as defined in the bases for any technical specification.

The implementation of th1s PC/M does not xequ1re a change to the plant Technical Specifications. I The foregoing constitutes per 10CRR50 59(b) the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question and prior NRC approval for the implementation of this PC/M is not requix'ed.

PCM 143-187 TrTLE 480V PCB FILLED TRANSFORMER REPLACE DESCRlPTlONOF CHAN@K/ABSTRACTt EXISTING PCB FILLED 1500 KVA STATION SERVICE TRANSFORMERS lAl and 1Bl ARE BEING REPLACED WITH NON-PCB FILLED SILICONE IMPREGNATED DRY TYPE TRANSFORMERS TO SATISFY ENVIRONMENTAL CONCERNS REGARDING PCB'S.

Ifftplefftentetion af this DEEP does not constitute en unrevieared sefety question nor effect Plent Technicel Specifica-tions. NRC epprovel is not required prior ta istplementetion. This DEEP hes no irdpect on plent sefety or operetion.

NUCLEAR SAFETY EYALUATLON CHECKLlST The ~ritten evaluation of the proposed design change to demonstrate that the change does not alter the plants design basis and is bounded by the design analyses is attached to the Design Equivalegt Engineering Package. The answers below are supported by this evaluation.

TYPE OF CHANGE Yes X No A change to the plant as described in the FSAR?

Yes No x A charge to procedures as described in the FSAR?

Yes No X . A test or experiment not described in the FSAR?.

Yes No x A change to the plant technical specifications?

EFFECT OF CHANCE Yes No x Wftl the probability of an accident previously evaluated in the FSAR be increased?

Yes No x %ill the consequences of an accident previously evaluated in the FSAR be increased?

Yes No May the possibility of an accident which is different than any already evaluated in the FSAR be created?

Yes No 1Vill the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

Yes No x  %'ill the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

Yes No x May the possibility of a malhaction of equipment important to safety different than any already evaluated in the FSAR be created?

'es No~ .

%ill the margin of safety as defined in the bases to any technical specification be reduced?

PCM 152-187 SIT SAMPLE VALVE AS BUILDING MODIFICATION DESCRIPTION OP CHANGE/ABSTRACT: Revise CWD 8770-B-327 Sh 322 to show as-building state of SIT Sampling Isolation Valve, I-FCV&3-1P wiring as follows: W&B conductors of H-SB to be shown connected to TB639: 9 & 10 respectively instead of TB635: 7 & 8 as per attached marked ordrawing. This.is a drawing change only. It does not affect system function qualification.

it It does not require a Tech. Spec. change and does not involve an unreviewe safety question.

NUCLEAR SAFETY EVALUATiON CHECKLlST The written evaluation of the proposed design change to demonstrate that the change does not alter the plants design basis and is bounded by the design analyses is attached to the Design Equivalent Engineering Package. The answers below are supported by this evaluation.

TYPE OF CHANCE Yes .

No x A change to the plant as described in the FSAR?

Yes No x A change to procedures as described in the FSAR?

Yes I'lo x A test or experiment not described in the FSAR?

Yes No x A change to the plant technical specifications?

EFFECT OF CHANCE Yes No x Will the probability of an accident previously evaluated in the FSAR be increased?

Yes No x Will the consequences of an accident previously evaluated in the FSAR be increased?

Yes No X May the possibility of an accident which is different than any already evaluated in the FSAR be created?

Yes No X Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

Yes No~ Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

Yes No x May the possibility of a mal function of equipfnent important to safety different than any already evaluated in the FSAR be created?

Yes No x Will the margin of safety as defined in the bases to any technical specification be reduced?

PCM 157-187 ST LUCIE UNIT 1 CONTROL ELEMBC DRIVE SYSTEM 6 COIL POWER PROGRAMMER PART LENGTH REMOVAL REA-SLN-86&5 This Engineering Package (EP) provides for the removal of unused equipment in the Control Element Drive System (CEDS). The unused equipment wss previously employed for power shaping with part-length control elements. The part-length control elements have been xemoved from the reactor. The electronic components associated with these elements (power supplies, coil power programmer modules, power shaping group modules, displays, etc.) will be removed and maintained as spares.

The Control Element Drive System is not a Nuclear Safety Related System (see FSAR Section 7.1). Howevex, since the CEDS is used to contx'ol reactor operation, and since modifications to the RTG Board must be reviewed for their effect on RTGB seismic qualification, this Engineering Package is classified as Quality Related Implementation of this PCM does not involve an unreviewed safety question or a change to the Plant Technical Specifications. Therefore, pxior commission approval for the implementation of this FCM is not required.

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

pCN 157-187 SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulatione, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety questioat (1) if the probability of occurrence or the consequences of an accident or malfuaction of equipment importaat to safety previously evaluated in the Safety Analysis Report may be 1ncreased: or (11) if the possibility for an accident or malfunctioa of a differeat type thea any evaluated previously in the Safety Analysis Report may be created; or (iii) if the margin of safety as defined in the basis for aay technical specificatioa 1s reduced.

The modifications iacluded in this Engiaeering Package consist of thc removal of non-functioaing equipment, aot classified as safctymelated, which has no effect on operating plant systems. The modificatioae do not involve an unreviewed safety question because:

1) The probability of occurreace or the consequences of an acc1dent or malfunction of equipment important to safety previously evaluated are aot increased since ao modification is made to any safety related component, system, or function.

ii) There is no possibility for an accideat or malfuaction of a different type than any previously evaluated s1nce ao new safetymelated functioae or. interfaces with safetymelated systems are created by this EP.

iii) This modification does not change the margin of safety as defined ia the basis for any Techaical Specification, since ao equipmeat removed or modified by this EP affects any parameter refereaced in the Technical Specifications.

This EP does not mod1fy equipment which is nuclear eafetymelated.

However, since the Control Element Drive System ie used to control reactor operation aad siace modifications to the RTG Board must be rev1ewed for their effect on RTGB seismic qualification, this EP is classified as ~lity Related.

This EP has ao effect on cables or componeats necessary for safe shutdowa of the plant, or on equipment on the Essential Equipment List. Changes to equipment and structures involving 10CFR50 Appends "R" fire protection requirements have been addressed. Thus, the proposed des1gn is ia compliance with applicable requirements for fire protection.

The implementation of this change does not require a change to the Pleat Technical Specificatioas.

Thc foregoing constitutes, per 10CFR50.59(b), the writtea safety evaluation which provides the bases that this change does not 1nvolve an uareviewed safety questioa and prior Commission approval for the implementation of this PCM is not required.

PCM 001-188 ST. LUCIE UNIT I lSISTllRE SEPARATOR ZEHEATER SHELL REPAIR (REA-SLN-87-031)

This design package provides the necessary engineering for adding erosion protection features to the internal surfaces of the Moisture Separator Reheater {MSR) shells.

The effort involves the installation of chromium-molybdenum liner plates to the shells in the area(s) being affected by wet steam impingement/erosion.

Based on the failure modes analysis and 10CFR 50.59 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 and is correctly classified as non-nuclear safety related. Implementation of this modification, therefore, does not require prior NRC approval. There are no technical specifications affected and the modifications will not affect plant safety or operation.

~FU D This design package provides the necessary engineering for adding erosion protection features to the internal surfaces of the Hoisture Separator Reheater (HSR) shells.

The effort involves the installation of chromium-molybdenum liner plates to the shells in the area(s) being affected by wet steam impingement/erosion.

This modification is classified as non-nuclear safety related, since the HSRs perform no safety related function and do not interact with safety related equipment, components, or functions.

Based on the above 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.

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

The HSRs are not used in any safety analysis for accidents or malfunction of equipment and's such are non-safety related and have no effect on equipment vital to plant safety.

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

The components involved in this modification have no safety function and no changes have been made to the operational design to the system.

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

Since the components involve in this modification are not included in the bases of any technical specification, the margin of safety has not been reduced.

10 CFR 50.59 allows changes to a facility as described in the FSAR if an unreviewed safety question does nof, exist and if a change to the technical specifications is not required. As shown in the preceding sections, the change proposed by this design package does not involve an unreviewed safety question because each concern posed by 10 CFR 50.59 that pertains to an unreviewed safety question can be positively answered. Also, no change to the Technical Specifications is required based on the above evaluation. Therefore, prior NRC approval is not.required for implementation of this modification.

PCM 003-188D TITLE Condenaex'x anaion Joint Zm in emtht Plate Modifications DES CRIPTION OF CHANGE/ABS3RACTt The existing impingement plate design is inadequate for satisfactory long-term performance. Welded attachmenents on the plates have continuously failed, causing the plates to fall on and damage condenser tubes. The new plate design involves no welding and will prevent il any furth er fa urea. Thee Condenser is a Non-Nuclear Safety Related Quality t

Group D C omponen ~ Noo changes to Technical Specifications are require d , an "Dss and no unreviewed safety questions are involved. This PCM will not affec p lant t an safety or operation.

NUCLEAR SAFETY EVALUATION CHECKLIST The written evaluation of the proposed design change to demonstrate that the change does not alter the plants design basis and is bounded by the design analyses is attached to the Design Equivalent Engineering Package. The answers below 'are .supported by this evaluation.

Yes Yes No No

'A r "

TYPE OF CHANGE change to the plant as described in the FSARV A change to procedures as described in the FSAR'?

Yes No .~A test or experiment not described in the FSAR'?

Yes No t ~A change to the plant technical specifications'?

EFFECT OF CHANGE Yes Na I ~ 'Will the probability of an accident previously evaluated in the FSAR be increased'?

Yes No ~Wilt the consequences of an accident previously evaluated in the FSAR be increased'?

Yes Na ~May the possibility of an accident which is different than Yes No t'ill any already evaluated in the FSAR be created'?,

the probability of a malfunction of equipment important to safety previausly evaluated in the FSAR be increased?

Yes No L Will the cansequences of a malfunctian of equipment important ta safety previously evaluated in the FSAR be

~ increased?

Yes No May the possibility af a malfunction of equipment important to safety different than any already evaluated in the FSAR be created?

No uv Will the margin of safety as defined in the bases to any technical specification be reduced'?

PCM 005-188 METRAS COPE REPIACEMENT ST LUCIE UNIT NO 1 REA-SLN-87-56-10 ABSTRACT St Lucie Unit 1 currently utilizes a Metrascope System to monitor and display the Control Element Assembly (CEA) positions. This system will be replaced with a new one which has color graphics and a programmable computer for data processing and display creation. This will alleviate excessive calibration time, provide CEA displays more consistent with Unit 2, and modify Pre-Power and Power Dependent Insertion Limits (PPDIL/PDIL) which result in restricted CEA operation of several inches at full power. Additionally, the replacement will resolve eight open Human Engineering Discrepancies (HEDs) cited against the Metrascope System during the Detailed Control Room Design Review. The HEDs revolve around the existing system's display inadequacies and the lack of operator control over display generation.

The Control Element Assembly Position Display System (CEAPDS) is not a Safety Related system since it does not function to assure the integrity of the reactor coolant boundary, the capability for safe shutdown of the reactor, or the capability to prevent or mitigate the consequences of accidents which could result in off-site exposurea described in 10CFR100. However, the proposed components will be seismically mounted in RTGB-104. Therefore this EP is classified as Quality Related.

The safety evaluation concluded that the modifications implemented by this EP do not involve an unreviewed safety question and that prior NRC approval for the implementation of this EP is not required. Since the monitoring function of the system will not be changed by the upgrade, there will be no effect on plant safety and operation. There is no change to the plant Technical Specifications.

SUPPLEMENT 1 This supplement to the Engineering Package adds a cable retractor for the CEAPDS CRT which will help protect its cables and a noise isolator to the 9-power input, which will prevent a potential ground fault from being transferred from the new CEAPDS to the RPS.

Similar to the original issue, this supplement is Quality Related. The modifications implemented by this supplement do not involve an unreviewed safety question, therefore prior NRC approval for its implementation is not required. There will be no effect on plant safety and operation or to the Plant Technical Specifications.

PCM 005-188 SAFETY EVALUATION Qfth respect to Title 10 of the Code of Federal Regulatione, Part 50 59,' proposed change shall bc deemed to involve aa uareviewed safety question: (i) if the probability of occurrence or the consequences of an accident or malfunction of equipmeat important to safety previously evaluated ia 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 ae defined in the bases for any Technical Specification is reduced.

(i) The probabilfty of occurrence and the consequences of an accideat or malfunction of equipment important to safety previously evaluated in the Safety Aaalysis Report vill not be increased by this modification because it does not modify or affect any Safety Related equfpment and the new components are eefsmically mounted. Therefore it hae ao effect oa the function of any equipmeat required to preveat or to mitigate the effects of an accident.

(ii) The possibility for an accident or malfunctfon of a different type than any evaluated previously in the Safety Analysis Report vill not be created since no new failure modes are introduced which could change the function of any Safety Related equipment (iii) The margin of safety as defined in the bases for any Technical Specification ie not reduced since this modification does not reduce the operability of the rod block cfrcuit or the CEA poeitfoa indication systems. Instead, the modffications implemented by this EP will improve the operator's ability to determine the position of the CEAs aad to identify limiting conditione.

The Coatrol Element Assembly Position Display System (CEAPDS) is not a Safety Related system since it does not function to assure the integrity of the reactor coolant boundary, thi capability for safe shutdown of the reactor, or the capability to prevent or mitigate the consequences of accidents which could result in off-site exposures described fn 10CFR100. However, the proposed components will be seismically mounted in RTGB-104 and qualification of the board has been reviewed to ensure its seismic integrity. Therefore this EP is classified as Quality Related.

The implemeatatfon of this EP does aot require a change to the Plant Technical Specifications.

The foregoing constitutes, per 10CFR50.59(b), the mitten safety evaluation which provides the bases that this change does not involve an unrevfewed safety qucstfon and prior NRC approval for the implementation of thfe PCM fs not required.

i Pm 006-188 ST LUCIE PLANT UNIT NO 1 RCP SEAL COOLER HEAT EXCHANGER TUBE LEAK DETECTION ABSTAACT This Engineering Package addresses the replacement of existing limit switches for Component Cooling Water (CCW) outlet valves HCV-14-11-A1, 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 o<<e<<<<ng existing time delay relays to introduce a 60 second time Thi<<ime delay vill allow 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.

CCW to the RCp is classified as Non-Nuclear Safety Related and non-seismic according to St Lucie Plant - Unit 1 (PSL-1) FSAR Section 9.2.2.3. Also, the valve position indication circuits are Non-Nuclear Safety Related. Howevers since the function of the seal cooler isolation valves is to isolate reactor coolant leakage into the component cooling system, this EP is classified Quality Related.

The safety evaluation of this package indicates that neither the replacement of the limit switches no>> the valve control circuit wiring modifications constitute an unreviewad 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.

PCM 006-188 SAFETY EVhLUATION 1th respect to Title 10 of the Cgde of Federal Regulati.ons> Part 0.59, a proposed change shall be deemed to involve an unreviewed if

~ ~

safety question.'1) 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 acc1dent 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 modif1cations included in this Engineering Package do not involve an unreviewed safety question because of the following reasons:

(1) The probability of occurrence and the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analys1s Report will not be increased by this modification. Electrical separation is maintained between safety related wiring and components. The modifications provided by this 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 malfunction of equipment important to safety.

(ii) The possibility for an accident or malfunction of a different type than any eviluated previously 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.

(iii) The margin of safety as defined in the bases for any technical specification is not reduced since this modificat1on does not degrade the CCV system, and the CCW Seal Coolers do not form the bases of any Technical Specification.

As described 1n PSAR section 9.2.2 the Component Cooling System is a closed loop cooling water system that utilices demineralised water to cool various components. The modifications described in this PCM involve replacing existing limit switches and rewiring the associated CCM outlet valve circuits. These changes do not interrupt the closed loop Component Cooling Mater System and are to a Non-Nuclear Safety Related valve indication function which discriminates between a fully closed and not fully closed valve position However, since the function of the seal cooler isolation valves is to isolate reactor coolant leakage into the component cooling system, this EP has been determined to be Quality Related.

The implementation of this EP does not require a change to the Plant Technical Specifications, nor does question.

it create an unreviewed safety The foregoing constitutes, per LOCFR50.59(b), the written safety evaluation which provides the bases that these changes do not involve an unreviewed safety question and prior NRC approval for the implementation of this PCM is not required.

PCM 007-188 RCP VIBRATION MONITORING E(}UIPMENT UPGRADE

( ST LUCIE UNIT NO 1 REA-SLN-86-018 ABSTRACT The Reactor Coolant Pump (RCP) Vibration Monitoring System 1e utilized to moaitox the v1bration of the RCP shafts on all four RCPs. It 1s made up of two radial probes (X 6 Y) located 90o out of. phase with each other gust above the mechanical seal of each RCP. Vibrations sensed by the probes are monitored by four electronic modules mounted on the rear face of RTGB-104 in the Contx'ol Room.

This Engineering Package will implement an upgrade to the RCP V1bration Monitoring System which will include the replacement of the X and Y probes, their relocation to the lower motor shaft area of each pump, and the addition of a third probe (Keyphasor) in the same area of each pump to provide rotational phase position information. The four electronic modules of the existing system vill be replaced by two modular instrument racks containing probe monitors for all twelve new probes, two pumps per rack.

The RCP Vibration Monitoring System is not a Safety Related system since it does not function to assure the integrity of the reactox coolant boundary, the capability fox safe shutdown of the xeactor, or the capability to prevent or mitigate the consequences of accidents. However, the proposed components will be seismically mounted in RTGB-104. Therefore this EP ie classified as Quality Related.

.The safety evaluation concluded that the modifications in the RTG board as implemented by this EP do not involve an unreviewed safety question and that pxior NRC approval for the implementation of this EP is not required. Since the monitoring function of the system will not be changed by the upgrade, there will be no effect to plant safety and operation. There is no change to the Plant Technical Specifications-SUPPLEMENT 1 Revision 1 of this Engineering Package has been issued to provide the installation of cable, conduit, and probes associated vith Reactor Coolant Pumps (RCPs) iBl and lA2 only. The safety evaluation xemaias valid with the implementation of this supplement; this EP does not involve an unreviewed safety question and prior NRC approval is not xequired for ite implementation. This revision to the EP has no effect on plant safety or operatioa and does not involve any change to the Plant Technical Specificatioas.

SUPPL1RENT 2 Revision 2 of this Eng1neering Package has been 1ssued to 11ft all remaining hold points to allow the iastallation of cable, conduit, and probes associated with Reactor Coolant Pumps (RCPs) 1A1 and 1B2 in order to complete the implementation of this PCM. In additioa, the proximitoxe fox the tvo pumps vill be relocated in nev electrical boxes. The safety evaluation remains valid vith the implementation of this supplement; th1s EP does not involve an unreviewed safety question and prior NRC approval is not required for its implementation.

This revision to'he EP has no effect on plant safety or operation and does not involve any change to the Plant Technical Specifications.

PCM 007-188 SAFETY EVALUATION With respect to Title 10 of the Code of Fedexal Regulations, Part 50.59, a px'oposed change shall be deemed to 1nvolve 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 if (iii) the margin of safety as defined in the bases fox any Technical Specification 1s reduced.

(1) The probability of occurrence and the consequences of an accident or malfunction of equipment important to safety pxeviously evaluated in the Safety Analysis Repoxt will not be increased by the modif1cations in the RTG board as implemented in this Engineering Package because it does not mod1fy any Safety Related equipment and involves the seismic installation of all RTG board components. Therefore, it has no effect on the function of any equipment required to prevent or to mitigate the effects of an accident.

(11) The possibility for an accident or malfunction of a different type than any evaluated previously 1n the Safety Analysis Report will not be created since no new failure modes are introduced which could change the function of any Safety Related equipment.

(iii) The margin of safety as defined in the bases for any Technical Specification is not reduced since this modification does not interface with equipment listed in the Technical Specifications.

The RCP Vibration Monitoring System is not a Safety Related system since it does not function to assure the integrity of the reactor coolant boundary, the capability for safe shutdown of the reactor, or the capability to prevent or mitigate the consequences of accidents.

However, the proposed components will be mounted in RTGB-104 such that a seismic event will not cause them to damage adjacent Safety Related equipment. Therefore this EP is classified as Quality Related.

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

The foregoing cqnsti.tutes, per 10CFR50/59(b), the written safety evaluation which provides the bases that this change does not involve an unx'eviewed safety question and that prior Commission approval for the implementation of this PCM is not requird for work to be performed in the RTG board.

ST LUCIE PLANT - UNIT NO 1 PCM 009-188 UPDATED OP LIMITORQUE EQ DOCUMENT PACKAGE AND DISCONNECT SPACE HEATERS (REA-SLN-87-007)

ABSTRACT Limitorque valve operators whose limit switch compartments have been furnished with space heaters have been recognized by the NRC (IE Notice 86"71, "Recent Identified Problems With Limitorque Motor Operators" ) to pose a potential hazard to the internal wiring of the Limitorque operator. The hazard arises from internal limit switch compartment wiring potentially making contact with the energized space heater or the heater bracket. The resultant insulation damage could conceivably result in these wires becom1ng grounded to the limit switch housing This Engineering Package will facilitate the removal of power to the space heaters thereby eliminat1ng the problem.

The Limitorque valve operators of this Engineering Package are aU.

Safety Related in as much as the valves they control perform nuclear safety related functions. Information relating to disconnecting power to the limit switch compartment space heaters will be included in the EQ Documentation Package for Limitorque motor operators. It is also the intent of this Engineering Package to remove from the Limitorque EQ Documentation Package Marathon . 1600 terminal blocks which are not considered, at this time, to be suitable for use in Environmental Qualification (EQ) applications. Additionally the use of 3M taped splices is prohibited in Limitorque operators 1nside containment and this also will be reflected by revis1on to the Lim1torque EQ Doc Pac.

Results of the safety evaluation conclude that mod1fications presented by this Engineering Package do not constitute an unreviewed safety question, do not require any changes to the Plant Technical Specifications and do not require prior Commission approval for the implementation of this PC/M.

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

PCM 009-188 SAFETY EVALUATION Qith 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 th'e safety analysis report may be created; or (iii) if the margin of safety es defined in the bases for any technical specification is reduced.

To determine the effect of removal of the space heater with regard to the criteria outlined in 10CFR50.59(a)(1) which allows plant changes without prior Commission approval, providing that the changes do not involve a change to plant Technical Specifications or an unreviewed safety question, the following criteria were addressed as required by 10CFR50.59(a)(2):

1) The probability of occurrence or the consequences of a design basis accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.

Disconnecting the power to the .space heaters does not increase the probability of occurrence or the consequences of a design basis accident or malfunction of equipment important to safety previously evaluated in the FSAR since operability of the Limitorque motor operator is not dependent upon having the space heater energized.

Limitorque does not recommend the use of energized space heaters for normal operation and has reported that their qualification testing was conducted with the space heaters de-energized. Limitorque furnishes space heaters for use during long term valve storage in an uncontrolled atmosphere.

Problems relating to potential motor operator inoperabilit'y were reported via ZZ Notice 86-71. This notice was in regard to the possiblity of damage to the control wiring of the motor operated valve due to contact with the apace heater/heater bracket. Power to the limit switch compartment space heaters in all valves in this package except for five auxiliary feedwater valves is removed from terminals in the Motor Control Centers. The determinated conductors are taped and otherwise left fn place. The five auxiliary feedwater motor operated valves have motor space heaters as well as limit switch compartment apace heaters. The limit switch compartment space heaters were paralleled with the motor space heaters for the five auxiliary feedwater valves. Therefore, the determination of the valve operator limit switch compartment space heaters for the auxiliary feedwater valves will be made at the terminals in the valve limit switch compartment in order to maintain the motor space heaters energized.

The 120 volt feeder to the motor compartment space heaters will be reconnected to the terminals in the limit switch compartment as wiU.

the leads to the valve motor enclosure space heater thereby keeping the motor space heater energized. The leads to the limit switch compartment space heater wi11 be taped and left in place.

r<

PCM 009-188 SAFETY EVALUATION (Continued)

' The safety related valves have not been physically modified and their operation is to remain the same. Therefore, there. is no change to their seismic or environmental qualification.

Prohibiting the use of Marathon 1600 terminal blocks for use with safety related valves and prohibiting the use of 3M taped splices inside of containment resolves concerns regarding the use of this equipment in view of NRC question regarding their suitablity.

2) The possibility of an accident or malfunction of a different type than any evaluated previously in the PSAR will not be created.

Removing the power to the space heaters wi11 have no affect on the operability of the plant motor operated valves. As stated above, Limitorque does not recommend the use of energized space heaters except for valves in storage and their qualification testing was not conducted with the apace heaters energized- Currently there are no Marathon 1600 terminal blocks in use with EQ related valves or 3M splices in use in containment. This Limitorque EQ Doc Pac update will prevent the possiblity of future accidents occurring due to the use of this material.

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

Since operability of the plant safety related motor operated valves is not affected by disconnecting the space heaters, 'since no Marathon 1600 terminal blocks are in use in the plant in association with EQ related valves and since no 3M splices are in use in containment, the basis for any'lant Technical. Specification is unchanged. Therefore, Plant Technical Specifications are unchanged by disconnecting the space heaters to plant safety related motor operated valves or prohibiting the use of Marathon 1600 terminal blocks or 3M splice in containment.

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 .change to plant Technical Specifications.

CONCLUSION:

It is therefore recommended that space heaters be de-energized for all safety related Limitorque motor operators at the St Lucie Plant in response to IE Information Notice 86-71. This recommendation is based upon the manufacturer's recommendation, upon favorable industry and PPL experience with motor operators having their space heaters de-energized and upon reported problems relating to potential damage to the internal wiring of the motor operators as described in IE Notice 86-71. Unit 1 motor operators have their heaters fed from circuits which are common to other heaters for fans and motors. The Unit 1 apace heaters aust have their power leads lifted at the respective MCCs or as described above.

/q PCM 010-188 ST LUCIE PLANT - UNIT 1 STATION AIR/INSXRM9iT AIR PRESSURE INDICATOR REPLACEMENT REA-SIR-87-13-10 This Engineering Package xeplaces the ezisting voltage driven dual indicating meter (PI-18-9/PI-18-16) with a curxent driven device and modifies the 4-20mADC curx'ent loop so that the new indicator will be in series in the loop. Replacement of the ezisting dual indicator is required due to its failux'e; revision to the loop configuration will allow for the replacement of the indicator with parts maintained in stores inventory.

This meter provides control room indication of station air and instrument air pressuxe, neither of which are classified as nuclear safety related systems This EP has no affect on any equipment required for safe reactox shutdown, used to mitigate the consequences of a design bases event (DBE), or control radioactive releases to the atmosphere in the event of a DBE. Since this EP involves the seismic analysis of mounting details for equipment mounted in the Reactor Turbine Genexator Board (RTGB), this package is classified as polity Related.

The implementation of this EP does not constitute an unreviewed safety question nor would its implementation affect the Plant Technical Specifications. Thus, Commission approval is not required prior to implementation This EP has no impact on plant hafety or operation.

1 SAPETY EVALUATION arith respect to Title 10 of the Code of Pederal Regulations, Part 50.59, a pxoposed change shall be deemed to involve an unreviewed safety questionx (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety hnslysis Report may be increased, or (ii) if the possibility fox an accident or malfunction of a different type than any evaluated previously in the safety as analysis repoxt may be created, or (iii) if the margin of safety is reduced.

defined in the bases for any technical specification

010-188 SAFETY EVALUATION (Coatiaued)

The modificatioas included in this EagineerIng Package do not Involve an unrevIewed safety question becauses (1) The probability of occurrence or thc coasequeaces of aa accident or malfunction of equipment important to safety previously evaluated is not increased. This ie confirmed by the followiag:

The Compressed Air System (statioa air aad Instrument air) serves no'afety function pex St Lucie Unit J FSAR Section 9.3.1.1.

The replacement of the dual pressure indicator PI-18-9/Pl-18-16 has no effect on any nucleax safety related equipment and its failure will not increase the probability of occurrence or the consequences of aa accident as. Indicated in Section 2.1.8 of this EP (11) There is no possibility for an accideat or malfunction of a different type than any previously evaluated ae confirmed by the following:

The modifIcation of the compressed aIr instrumentation loop uses the same cixcuit design used throughout the plant and has been previously evaluated for both safety and non-safety loops.

Replacement of the dual pressure Indicatox'I-18-9/PI-18-16 provides control room indicatioa of station air and iastrument air pressure by utilieing a current drivea device.

This configuration does not introduce any possibility of accident or malfunction not previously evaluated. Sec sect1on 2.1.8 of this EP.

(iii) This modification does not xeducc the margia of safety as defined in the bases for any technical specification since it hae ao negative effect on safety related components or systems as defined in any Technical Specifications and provides for statioa air and aix indication as originally specified in the St Lucie

-instrument Uait 1 Final Safety Analysis Rcport.

Since this package does aot affect any equipment that is identified as nuclear safety related, this package need not be considered related. However, since the implementation of this PCM nucleax'afety xequires work to be done inside the reactor turbine generator board (RTGB)> this package is classified Quality Related as the RTGS is a seismically desigaed control panel.

This EP does not involve any equipment on the Essential Equipment List and has no effect on safe reactor shutdown or alternate shutdown.

There are ao other changes to equipment which involve 10CFR50 Appendix "R" fire protectIoa {see Attachment 7.1).

Implemeatation of ~ity Related PCM 010-188 does not require any change to the Plaat Technical Specifications.

The foregoing constitutes, per 10CFR50.59 (b), thc wr1tten which provides the bases that t+s change does not involve safety'valuation an unreviewed safety question nor a change to any Technical Specificatione aad prior Commission approval for the implementation of this PCM is aot required.

PCM 011-188 ST. LUCIE UNIT I W KWAY NC OSUR REA-SLN-87-037 Ql~SRACT This engineering package covers the modification of the enclosed walkway which connects the Reactor Auxiliary Building and the personnel hatch enclosure at the Reactor Containment Building. The existing fiberglass panels which cover the walkway are being replaced with non-combustible materials. Also, the existing opening 'located at the south end of the personnel hatch enclosure will be sealed to prevent the entry of stormwater into the RAB RCB walkway and personnel hatch enclosures.

The existing RAB RCB walkway and personnel hatch enclosures do not perform any nuclear safety-related functions so this modification will not be classified as nuclear safety-related. However, this modification does require the installation of concrete expansion anchors in Seismic Class I structures including the Reactor Auxiliary Building. Since reinforcement steel in Seismic Cl.ass I structures could potentially be damaged during installation of this modification, quality-related requirements are applied to this design.

A safety evaluation of this modification has been performed in accordance with: $ 0CFR50.59. This evaluation indicates that implementation of this Engineering Package does not involve an unreviewed safety question.

Furthermore, the implementation of this modification does not require a ch'ange to the plant Technical Specifications and has no detrimental effect on plant safety and operation. Therefore, prior NRC approval for implementation of this modification is not required.

F Y VA U T ON fifth 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 equipment important to safety previously evaluated in the Safety Analysis Report may be increased, or (ff) if a possibility for an accident or malfunction of a different type than any evaluated previou'sly in the Safety Analysis Report may be created, or (iif) ff the margin of safety as defined fn the bases for any technical specification is reduced.

The modifications included in this engineering package do not involve an unreviewed safety question because of the following reasons:

(i) The probabflfty of occurrence and the consequences of an accident or . malfunction of equipment important to safety previously evaluated fn the Final Updated Safety Analysis Report are not increased by this modification because it does not affect the availability, redundancy, capacity, or function of any equipment required to mitigate the effects of an accident.

(ff) The possibility for an accident or malfunction of a different type than any evaluated previously fn the Final Updated Safety Analysis Report will not be created by this modification because the modfficat4on fnvolves non-nuclear safety-related structures and failure of any items added . by this modification will not impact any nuclear safety-related functions.

(iii) The margin of safety as defined in the bases for any technical specification fs not affected by this modification since the components fnvo1ved in this modification are not included fn the bases of any Technical Specifications.

The RAB RCB walkway and personnel hatch enclosures are classified as non-nuclear safety-related. Oue the location of the modificatfon, failure of the RAB RCB walkway enclosure or the modified portion of the personnel hatch enclosur'e will not affect any nuclear safety-related equipment. However, this modification does involve the installation of concrete expansion anchors in Seismic Class I structures. Since steel reinforcement in the structures could potentially be damaged during installation of the anchors, qualfty-related requirements have been applied to this design.

The implementation of this EP does not require a change to the Plant Technical Specifications, nor does question.

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

PCM 012-188 TITLE: IB & ID Instrument Inverter Drawing Change 0 DESCRIPTION OP CHANGE/ABSTRACT: This change modifies a drawing (See drawing list) to show the correct circuit numbers for the 125VDC feeds to the IB &

ID instrument inverters. No physical modifications are required, only a correction to a drawing. No unreviewed safety question or change to technical specifiction is required.

NUCLEAR SAFETY EVALUATION CHECKLIST The written evaluation of the proposed design change to demonstrate that the change does not alter the plants design basis and is bounded by the design analyses is attached to the Design Equivalent Engineering Package. The answers below pre supported by this e valua tion.

Yes No ~

~

TYPE OF CHANCE A change to the plant as described in the FSAR?

Yes Yes Yes No No No

~

~

A change to procedures as described in the FSAR'?

A test or experiment not described in the FSAR?

A change.to:the plant technical specifications' s

EFFECT OF CHANCE Yes No Will the probability of an accident 'previously evaluated'in No ~

~'es the FSAR be increased?

Will the consequences of an accident prev'iously evaluated in the FSAR be increased?

Yes No ~f May the possibility of an accident which is different than any already evaluated in the FSAR be created'?

Yes No Q~ Will the probability of a mal function of equipment important to safety previously evaluated in the FSAR be Yes No ~ increased?

Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

Yes No ~(. May the possibility of a malfunction of equipment important to safety different than any already evaluated Yes No ~ in the FSAR be created?

Will the margin of safety as defined in the bases to any technical specification be reduced?

0191L

Pm 013-188 Il SECURITY LIGHTING PANELS RELAYS AND CONTACTORS DESCRIPTION OF CHANCE/hBSTRACTs These cha es ovlde for the ro tacernent of a and for the documentation of failed ASCO re'la utilized ln Security Lighting Panel 2S, and there ls no change to condltlons An unrevlewed safety question does not exist 'Hwllt technical s clficatlons.involved s

~~1emcnt No. 1- This supplement is to add page 3a (Design Interface ecor an quahty eve on a ac men . o t~o'dicate the purchasingemst there 4 nochange to technical specificati~ involved.

suety question does not and NUCLEAR SAFETY EVALUATION CHECKLIST Thc written evaluation of the proposed design change to demonstrate that the change does not alter the plants design basis and is bounded by the design analyses is attached to the Design Equivalent Engineering Package. The answers below are supported by this evaluation.

TYPE OF CHANCE Yes No X A change to the plant as described in the FSAR?

Yes No X A change to procedures as described in the FSAR?

Yes No X A test or experiment not described in the FSAR?

Yes No . A change to.the plant.technical specif ications?

EFFECT OF CHANCE Ycs No Will the probability of an accident previously evaluated in the FSAR be increased?

Yes No X WIII the consequences of an accident previously evaluated in the FSAR be increased?

Yes No May the possibility of an accident which is different than any already evaluated in the FSAR be created?

Yes No X Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

Yes No X Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

Yes No X May the possibility of a malfunction of equipment important to safety different than any already evaluated in the FSAR be created?

Yes No X Will the margin of safety as defined in the bases to any technical specification be reduced?

PCM 015-188 S 'UCIE PLANT UNIT NO 1 ICM LUBE WA ER PIPE RES~~T MODIFICATIONS ABS'iiACT The tornado m'ss'le barriers around the intake cooling water pumps are exposed to salt water spray from the pump packing. The barriers are constructed of coated ca-bon steel and have suffered corrosion from the salt water exposure.

In particular, the structural members near the bottom of the enclosure on the

.east and west faces have experienced severe deterioration. Several of these members furnish support for ICW lube wate" system pipe restraints.

An evaluation of the as-found condition is being performed as part of the overall effort associated with the disposition of NCR 1-133. Fending a long oblem this term solution to correct the root causes of the corrosion probl Engineering Package is being issued to modify those corroded structural elements which are integral parts of the pipe restraints.

This Engineering Package does not involve an unreviewed safety question and has no effect on plant safety or operation, nor does it require a change to the plant Technical Specifications The system involved is classified as Safety Related, consequently this Engineering Package is also classified as Safety Related.

PCM 015-188 SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to 1nvolve an unrevtewed 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 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 1s reduced.

This Engineering Package provides modificat1ons to pipe restraints for the ICW lube water system, which is a safety related system.

Accordingly, this Engineering Package has been classified as Safety Related. It does not involve an unreviewed safety question. The following are the bases for this conclusion:

(1) 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 restores and enhances the original design margin of the affected restraints and w111 be performed in accordance with Safety Related requirements, hence there can be no impact on any Safety Related structures, systems, or equipment.

(11) There is no possibility for an accident or malfunction of a different type than any evaluated previously since there is no potential for the interaction of these modifications with any Safety Related equ1pment or systems other than the ICW lube water system itself and the ICW pump missile barrier to which the affected restraints are attached; the restoration of these components to their original design margin will not affect any safety related systems or equipment.

(iii) This modification does not change the margin of safety as defined in the basis for any technical specification as the mod1fications have been designed to the same criteria as the restraints of which they are a part.

~

implementation of this Engineering Package does not require a

'he change to plant technical specif1cations.

The foregoing constitutes, per 10CFR50.59(b), the written safety evaluation which provides the bases that this change does not involve an unrevtewed safety question or a Technical Specification change and thus prior Commission approval for the implementation of this Engineer1ng Package is not required.

PCM 01 8-188 FLORIDA PSKR & LIGHT COMPANY ST LUCIE PLANT - UNIT NO 1 CONDENSATE PUMP DISCHARGE SAMPLING LINES REA-SLN-86%61-92 ABSTRACT This Engineering Package provides details for the addition of condensate sampling points downstream of each condensate pump (lA, 1B, 1C) and in the common discharge line for all the pumps. It also provides for the connection of these sample points, through a valve manifold, to an existing sample line to the Chemical Analyzer in the Cold Chemistry Laboratory.

~s EP is classified as non-safety related since it provides for a-modification to a non-safety related system. The safety evaluation has shown that this EP does not constitute any unreviewed safety question.

The implementation of this EP does not require a change to the Plant Technical Specification; therefore prior NRC notification for implementation of the EP is not required.

This sampling system is non-safety related and will have no effect on equipment vital to plant safety, nor will it effect plant operation.

PCN 018-188 SAFETY EVALUATEON arith respect to Title 10 of the Code of Federal Regulat1ons, 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 malfunct1on 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 (111) if the margin of safety as defined in the bases for any technical specification is reduced.

This EP provides for the addition of sampling points in the condensate lines, connection of these points to a valve manifold station and then to the existing common line to the Cold Chemistry Lab. Stainless steel tubing with compression fittings will be used for the sample lines.

The EP has been classified as non-safety related and does not involve an unreviewed safety question because-(1) The probability of occurxence ox'he consequences of an accident or malfunction of equipment impox'tant to safety previously evaluated in the safety analysis report is not increased. The St Lucie Unit No l FSAR, Section 10.4, "Steam and Power Conversion Syst: em", states that the features and components of this system, which includes the condensate system, serve no safety function since they axe not required for safe shutdown or to mitigate the effects of a LOCA. This modification is on a non-safety related system and will have no effect on equipment vital to plant safety.

(11) The possibility for an accident or malfunction of a different type than any evaluated previously 1n the safety analysis report is not cxeated. The components involved in this modification have no safety related function and no changes have been made to the operational design of the system.

(iii) The marg1n of safety as defined 1n the bases for any Technical Specification is not affected by this P(X, since the components involved in this modification are not included in the bases of any Techn1cal Specification.

The implementation of this PQ4 does not require a change to the plant Technical Specifications.

The foregoing constitutes, per 10CFR 50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question and prior Commiss1on approval for the implementation of this P(H is not required.